ML20202J305

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

ML20202J305
Person / Time
Issue date:	11/30/1997
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-0040, NUREG-0040-V21-N03, NUREG-40, NUREG-40-V21-N3, NUDOCS 9712110121
Download: ML20202J305 (167)

Text

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NUREG-0040 Vol. 21, No. 3 l

Licensee Contractor and Vendor Inspection Status Report Quarterly Report July - September 1997 U.S. Nuclear Regulatory Commission OITice of Nuclear Reactor Regulation i l

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

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

2. The Superintendent of Documents, U.S. Government Printi. 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 documents available for inspection and copying for a fee from the NRC Public Document Room include NRC correspondence and internal NRC 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 purchese f rom 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 Co;;s of Federal Regula-tions, and Nuclear Regulatory Commission Issuances.

Documents evallable from the Nationt.1 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, Documst' available from public and special technical libraries include all open literature items, such as tooks, journal articles, and transactions, Federal Register notices, Federal and State legislation, and congressional reports can usually be obtained f rom these libraries, Documents such as theses, dissertations, foreign reports and translations, and non-NRC con-forence proceedings are available for purchase from the organization sponsoring the publica-tion cited.

Single (,opies 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. Nucl ear  :

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-ville, MD 20852-2738, for use by the public. Codes and standards are usually copyrighted and may be purchased from the originating organization or, if they are American National Standards, from the Ame ican National Standards institute,1430 Broadway, New York, NY 10018-3308.

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

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! NUREG-0040 Vol. 21, No. 3

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l Licensee Contractor and Vendor Inspection Status Report Quarterly Report July - September 1997 Manuscript Canpleted November 1997 Date Publisha!. November 1997 Division of Reactor Controls and Iluman Factors Omce of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Wcshington, DC 20555 0001 f"' %,,,

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l NUREG-0040, Vol. 21, No. 3 has been reprortuced from the best available copy.

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ABSTRACT This periodical covers the results of inspections performed between July 1997 and September 1997 by the NRC's Special Inspection Branch, Vendor inspection Section, that have been distributed to the inspected organizations.

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l CONTENTS PAGE Abstraet............................................................................................................... ill I n t r od u ctio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

I n s pe cti o n R e port s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . 1. . . . . . . . . . . . . . . . . . . . . . . .

ABB Services, Inc. (99901281/97 01) .. ............... ... ........ 2 Columbia, MD Hartford Steam Boiler (99900601/97 01) ..... .......... .. .......... 19 i inspection and Insurance Company Hartford, CT NUS Instruments, Inc. (99901320/97 01) ............ .... ............. 40 Idaho Falls, ID Pacific Gas & Electric Company (50 275/97-201, 50-323/97 201).......... 52 Diablo Canyon Power Plant San Francisco, CA PECO Nuclear (50-352/96 201, 50-353/98-201).......... 68 Wayne, PA Tritium Target Qualification Program (99900541/97-01) ....... .................. . . 94 Pacific Northwest National Laboratorf Richland, WA SOR, Inc.

(99900824/97-01).......... .. .............116 Lenexa,KS Yuasa Exide, Inc. (99900358/97-01, 99900359/97-01).. 132 Reading, PA Select Generic Correspondence on the Adequacy of Vendor .. ..........,..........155 Audits and the Quality of Vendor Products v

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INTRODUCTION 1

A fundamental premise of the U. S. Nuclear Regulatory Commission (NRC) licensing and I inspection program is that licensees are responsible for the proper construction and safe and efficient operation of their nuclear power plants. The Federal government and nuclear industry have established a system for the inspection of commercial nuclear facilities to provide for I

multiple levels of inspect:on and verification. Each licensee, contractor, and vendor participates

] in a quality verification process in compliance with requirements prescribed by the NRC's rules

and regulations (Title 10 of the Code of Federa/ Regulations). The NRC does inspections to l oversee the commercial nuclear industry to determine whether its requirements are being met

] by licensees and their contractors, while the major inspection effort is performed by the industry within the framework of quality verification programs.

The licensee is 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, contracton, and vendors also have suitable and appropriate quality programs that meet NRC requirements, guides, codes, and standards.

The NRC 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 veridors to verify conformance with applicable NRC and industry quality requirements, to verify oversight of their vendors, and coordination between licensees and vendors.

The NRC does inspections to verify the quality and suitability of vendor products, licensee-vendor interface, environmental 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 4

appropriate corrective action and to institute preventive measures to preclude, recurrence.

j When generic implications are found, NRC ensures that affected lic6nsees 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 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 require special attention. This report lists selected bulletins, generic letters, and information notices, and include copies of other pertinent correspondence involving vendor

. Issues.

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k INSPECTION REPORTS

n a tecg l y-  % UNITED STATES 3

4 j NUCLEAR REGULATORY COMMISSION o t WA6HINoTON. D.C. 205S4001 *

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July 23, 1997 '

Mr. John J. Connolly, Vice President Business Development ABB Service incorporated 4

9050-A Red Branch Road Columbia Maryland 21045

SUBJECT:

NRC INSPECTION REPORT NO. 99901281/97-01

Dear rir. Connolly:

, On May 14, 1997, the U.S. Nuclear Regulatory Commission (NRC) conducted an inspection at one of the ABB Service, incorporated (ABB Service), repair and refurbishment facilities. The enclosed report presents the results of the inspection that was conducted at your Cleveland, Ohio facility, the discussior ,

onducted with you during the exit meeting at the ABB Service Corporate office '

in Columbia, Maryland, on June 5, 1997, and subsequent discussions between July 14 through July 22, 1997, regarding the enclosed Notice of Violation (NOV).

During this inspection, the NRC inspectors found certain activities to be in violation of NRC requirements. Specifically, ABB Service failed to adequately evaluate seteral examples of potential defects regarding miswiring errors concerning K-Line circuit breaker solid state trip devices that had been shipped to the Perry Nuclear Power Plant (PNPP). The team found that ABB .

Service did not-evaluate the errors or transmit the information to other ABB Service customers even though the concern was potentially generic.

This violation is cited in the enclosed 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 follow the instructions specified in the enclosed NOV when preparing your response. The NRC will use your response, in part, to determine whether further enforcement at. tion is necessary to ensure compliance with regulatory requirements.

The NRC considers this violation significant because, as early as January 1993, ABB-Service was aware of potential premature tripping on new and refurbished K-Line breakers as a result of miswiring errors and reversal of current sensor polarity into the solid state trip device of K-Line low voltage circuit breakers. However, ABB Service did not assure that, as a minimum, all of its circuit breaker refurbishment facilities were aware of the potential deficiency. ABB Service was also aware in Januar.v 1993, that single phase testing methods of licensees and ABB Service would not identify a reversed polarity problem which could affect both safety and non-safety-related K-Line breakers. Similarly, ABB Service did not apprise their refurbishment facilities or licensees of the weakness in testing. If ABB Service had apprised licensees of the testing weakness, a report to the NRC would likely have been generated.

I Mr. John J. Connolly l l

Although ABB Service's failure to inform its customers of a potential defect in January 1993 was not a violation of NRC requirements, the staff believes that ABB Service did not act appropriately given the potential problems that could have ensued at operating nuclear power plants. The staff is also concerned that the January 1993 ABB Service evaluation report.that delineated this problem may not be the only ABB Service example where appropriate action was not taken.

In addition, the NRC inspectors found that the implementation of your quality assurance program failed to meet certain NRC requirements imposed on you by your customers. Specifically, ABB Service inspections performed on K-Line breakers failed to detect that the current transformers on two circuit breakers were assembl6d incorrectly. This nonconformance is cited-in the enclosed Notice of Nonconformance (NON)..and the circumstances surrounding it 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 (PDR).

Sincerely,

' l

( d M O., l Stuart A. Richards, Chief Special Inspection Branch l Division of Inspection and Support Programs  ;

Office of Nuclear Reactor Regulation I Docket 99901281/97-01

Enclosures:

1. Notice of Violation 2.- -Notice of Nonconformance

3. Inspection Report 99901281/97-01 cc: Mr. Joseph M. Tate General Manager ABB Service Inc.

Regional Service' Center 5311 Commerce Street Cleveland, OH 44130

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Enclosure 1 N0i!CE OF VIOLATION ABB Service Docket No.: 99901281 Cleveland Ohio During an NRC inspection conducted on May 14, 1997, and discussions corducted between July 14-22, 1997, a violation of NRC requirements was identified. In accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions," NVREG-1600, the violation is listed below:

10 CFR Part 21.21, " Notification of failure to comply or existence of a defect and its evaluation," requires in part that. (a) Each individual, corporation, partnership, dedicating entity, or other entity subject to the regulations in this part shall adopt appropriate procedures to: (1) Evaluate deviations and failures to comply to identify defects and failures to comply associated with substantial safety hazards as soon as practicable, and, except as provided in paragraph (a)(2) of 6 21.21, in all cases within 60 days of discovery.

ABB Service Quality Assurance Procedure (QAP) 15.1, " Control of Nonconforming items," Revision 4 dated December 15, 1996, implements, in part, requirements contained in 10 CFR Part 21.21. Section 3.5 of QAP No. 15.1 requires that any significant nonconformance in nuclear-safety-related equipment shall be evaluated to determine if a 10 CFR Part 21 report needs to be filed.

ABB Service QAP 15.2, " Reporting of Defects and Noncompliance in Accordance with 10 CFR 21," Revision 2, app-oved March 24, 1995, states in part, the evaluation of potential defects or potential failures to comply must be completed as soon as practicable, and in all cases, within 60 days of discovery.

Contrary to the above, even though ABB Service was aware of examples of potential defects regarding miswiring errors concerning K-Line circuit breaker solid state trip devices that they had shipped to Perry Nuclear Power Plant (PNPP), ABB Service did not adequately evaluate the errors or transmit information regarding the deviations to other ABB Service customers even though the concern was potentially generic. PNPP identified three refurbished K-Line breakers that had incorrectly wired or installed sensors (Serial Numbers 51817A-107073, 51817C-264135'and 518170-211135). All three had been incorrectly assembled by ABB Service during refurbishment, tested, and sent back to PNPP by ABB Service. Additionally, PNPP found the "C" phase Power Sensor current transformer inverted on a K-6005 low voltage breaker (Serial Number 518170-211135), and identified the problem to ABB Service on April 28, 1997. (99901281/97-01-01)

This is a Severity Level IV violation (Supplement Vil).

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Notice of Violation  !

Pursuant to the provisions of 10 CFR 2.201, ABB Services Inc., is hereby required to submit a written statement or explanation to the U.S. Nuclear Regulatory Comission, 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 clearly marked as a " Reply-to a Notice of Violation" and should include for each violation: (1) the reason for the violation, or, if contested, the L Pis 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. Your response may reference or include previous docketed correspondence, if the correspondence adequately addresses the required response. Where good cause is shown, consideration will be given to extending the response time.

Dated at Rockville, Maryland this ; '

• day of July 1997

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l Enclosure 2 NOTICE OF NONCONFORMANCE ABB Service Docket No.: 99901281 Cleveland, Ohio Based on the results of an inspection conducted on May 14, Ib7, it appears 1

that certain of your activities were not conducted in accordance with NRC requirements.

Criterion X of Appendix 8 to 10 CFR Part 50, " Inspection," requires, in part, that a program for inspection of activities affecting quality be established and executed by or for the organization performing the activity to verify conformance with documented instructions, procedures, and drawings for accomplishing the activity.

Contrary to the above, two safety-related K-Line breakers that had been returned by ABB Service, Cleveland, to a licensee after being refurbished were not properly inspected to verify their conformance to documented drawings, specifically: (99901281/97-01-02)

• On November 8, 1996, maintenance personnel at the licensee discovered that the control wires on terminals 12 and 13 of the Power Shield trip unit were reversed on breaker S/N 51817A-107073.

• On April 28, 1997, maintenance personnel at the licensee discovered that the power sensor on phase "C" was inverted (i.e., it was installed upside down) on breaker S/N 518170-211135.

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 Nc'conformance.- This reply should be clearly marked as a " Reply to a Notice of honconformance" 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 recerrence; and (3) the dates your corrective actions and preventive measures were or ' sill be completed.

Dated at Rockville, Maryland this _h ~

• day of July 1997

Enclosure 3 INSPECTION REPORT U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION DIVISION OF INSPECTION AND SUPPORT PROGRAMS ORGANIZATION: ABB Service incorporated Cleveland Service Ctnter 5311 Commerce Parkway West Cleveland, Ohio 44130 REPORT NO.: 99901281/97-01 ORGANIZATIONAL Mr. D.E. Leckey CONTACT: Quality Assurance (QA) Manager (216) 267 2882 NUCLEAR INDUSTRY Can provide new ABB switchgear, procurement of ACTIVITY: replacement parts, maintenance, refurbishment and on-site switchgear services.

INSPECTION CONDUCTED: May 14, 1997 SUBSEQUENT OlSCUSSIONS: June 5 and July 14-22, 1997 INSPECTORS: Kamalakar R. Naidu, NRR l Virgil L. Beaston, NRR '

Joseph J. Petrosino, NRR APPROVE 0 BY: Gregory C. Cwalina, Chief l Vendor inspection Section  !

Special Inspection Branch  !

Division of Inspection and Support Programs Office of Nuclear Reactor Regulation

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N 1 INSPECTION $UMARY ABW Service incorporated (ABB Service), is headquartered at-its Columbia, Maryland, Service Center. Prior to 1996 ABB Service had approximately 17 service center facilities throu  !

potentially performed engineering,ghout consulting the United services, States thatwork or other couldon, have or  !

related to, safety-related circuit breakers for NRC licensees. However, due i to quality concerns, ABB Service modified that policy and currently allows  !

only four of its service centers to process safety-related circuit breaker 1 work. They are: Cleveland, Ohio Service Center; Columbia, Maryland Service  :

Center; Houston, Texas Service Center; and Charlot9, North Carolina Service Center.

The ABB Service Center located in Cleveland, Ohio, has been performing refurbishment services on metal-cli.d. low and medium voltage, K-Line circuit breakers installed at Centerior Energy's Perry Nuclear Power Plant (PNPP) for several years. The K-Line breakers were originally designed and manufactured by I.T.E. Imperial Company (ITE), which changed ownership and became known as I.T.E.- Gould, Gould-Brown Boveri, Brown Boveri Electric and finally ASEA Brown Boveri. ABB Power Transmission & Distribution Company, incorporated (ABB Power) manufactures metal-clad low-voltage and medium-voltage K-Line breakers at Florence, South Carolina. ABB Power at Sanford, Florida assembles complete switchgear installations. The Cleveland Service Center currently services safety-related K-Line breakers.only for PNPP.

On May 14, 1997, the inspectors reviewed records documenting the refurbishment work performed on PNPP breakers, reviewed the actions taken to correct .

nonconformances identified in report 99901281/94-01, and performed subsequent  ;

review of records and documents associated with the reversed polarity issue in K-Line circuit breakers. Additionally, the inspectors reviewed an ABB Service Report performed by ABB Service for Public Service Electric and Gas Company's (PSE&G) Salem nuclear generating station (Salem) as discussed in Section 3.2.b.3. The inspection bases were:

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

• 10 CFR Part 21, " Reporting of Defects and Noncompliance" During this inspection, the inspectors identified one nonconformance (Section 3.1.b.3 and 3.1.b.5) and one violation (Section 3.2.)

2 STATUS OF PREVIOUS INSPECTION FINDINGS Unresolved item 94-01-01 (Closed). The previous inspection identified that a current organization chart was not available. The team determined that ABB Service has a current organization chart which depicts the organization.

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l Unresolved item 94-01-02 (Closed). The previous inspection identified that job descriptions and quallfications of persons performing safety-relhted

, activities were not available. 'The team identified that ABB Service has identified and delineated job descriptions and qualifications-of persons performing safety-related activities.

Nonconformance 94-01-03A (Closed). The previous inspection identified that potential deviations were not-appropriately dispositioned. Although the team determined that adequate corrective action was not performed for this issue, similar corrective action will be required as a result of Violation 99901281/97-01-01.

Nonconformance 94-01-03B (Closed). The previous inspection determined that not all contact resistances were delineated for purchaser use. The team determined that maximum acceptable contact resistance values are now furnished

in the ABB Service test instructions.

3 INSPECTION DETAILS l 3.1 K-Line_. Breaker Refurbishment

a. Es. ggt The inspectors reviewed circumstances that led to incorrectly assembled ABB Sairvice K-Line breakers being delivered to PNPP. During and subsequent to the inspection at ABB Service's Cleveland facility, the inspectors:

• visited PNPP to examine the suspect K-Line circuit breaker,

• reviewed the purchase order (P.O.) PNPP issued to ABB Power-Florence, for the supply of new ABB K-Line low-voltage metal-clad K-Line breakers,

• examined the documentation related to the work performed by ABB Service on PNPP breakers to determine if it met tne P.O.,

• observed ABB Service technicians perform work on selected K-Line breakers at the Cleveland Service Center,

• discussed K-Line breaker wiring problems identified by PNPP and other licensees with representatives of ABB Service, and

• ' reviewed associated documentation and conducted discussions regarding identified K-Line polarity problems including PNPP and '

Salem.

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) b. Observations andfidiDai b.1 K-Line Breaker Fat re at PNPP On October 4, 1996, while PNPP was operating at 100 percent power, a new ABB K6005 K-Line Breaker (Serial Number 934277-031295) supplying 4BO-volt power to a motor control center prematurely tripped. PNPP personnel investigated the cause of this incident and determined that ABB Power-florence, had shipped the K-Line Breaker with tro lead wires from one of the areaker's three phase sensors (current transformers) reversed. PNPP shipped the breaker to ABB Power for analysis. On March 26, 1997, ABB Power-Sanford, informed the Nuclear Regulatory Commission in accordance with 10 CFR Part 21 requirements that the two wires on the phase-C phase sensor were crossed on the breaker that was shipped to PNPP. Crossing the phase sensor leads changed the polarity of the phase sensor and caused the K-Line Breaker's Power Shield trip unit to trip at 350 amperes of primary current instead of the Power Shield trip setpoint setting of 660 amperes.

Six doughnut-shaped current transformers (cts) are mounted on the lower molding current transformer assembly of the K6005 K-Line breakers equipped with trip units. Three of the six cts are referred to as phase sensors, and they are used to detect fault currents. The other three are referred to as i power sensors, and they are used to develop a reference signal within the Power Shield trip unit. One phase sensor and one power sensor are installed on each phase of the K-Line Breaker. According to Revision 16 of ABB Power Distribution Inc. Drawing 709551, " Aux. Physical Wiring Drawing," the phase sensors are mounted on top of the power sensors and the leads of both the sensors are terminated on a terminal block attached to the lower molding. Two leads emerge from the top of each sensor, and a red dot (a polarity mark) on the top of each sensor distinguishes the polarity of the sensor. The other ends of the lower terminal blocks are connected to the appropriate terminals of the Power Shield trip unit using a multiconductor cable.

In the new breaker that tripped prematurely at PNPP, the blue wire from the phase-C phase sensor, which should have been landed on terminal 2 of the lower terminal block, was found fastened to terminal 1; the yellow wire, which should have been fastened on terminal 1, was founded-to be fastened to terminal 2. This error caused an errant, phase-shifted signal to be sent from the phase-C phase sensor to the K-Line Breaker's Power Shield solid state trip unit. The errant signal caused the trip unit to sense an abnormally high current value, tripping the breaker.

After learning that an incorrectly wired phase sensor reduces the amount of primary current needed to actuate ABB's Power Shield trip unit, PNPP began conducting secondary wiring checks of ABB K-Line circuit breakers. The wiring checks performed by PNPP identified three refurbished K-Line breakers that had incorrectly wired or installed sensors (Serial Numbers 51817A-107073, 5)8170-264135 and 518170-211135). All three of these incorrectly assembled K-Line breakers had been refurbished and tested by ABB Service.

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The inspectors determined that current and phase sensors (sensors) wiring and  !

assembly errors may occur in any of the following manners: l

• Crossing wires from the sensors to the terminal block attached to the lower molding current transformer assembly;

• Crossing wires from the terminal block on the lower molding to the terminal-block on Power Shield Static Trip Unit.

• Incorrectly re-connecting selected wires on the Power Shield Static Trip Unit after tie wires have been lifted to conduct K-Line Braaker tests.

• Installing a sensor upside down or incorrectly wiring i sensor when the i lower molding current transformer assembly of the K-L4ne Breaker is disassembled (to replace a sensor).

1 Additionally, as discussed in Section 3.2.b.2 below, the reversed polarity

. problem in K-Line circuit breakers appears to have been initially identified

by ABB Service in the 1992-1993 time period. ABB Service was aware that the polarity problem could have the potential to exist in new or refurbished K-Line breakers and that single phase testing methods typically used by the licensees and ABB Service organization would not identify the problem. The >

problem could affert safety and non-safety-related K-Line brtikers. However, ABB did not take action to assure the circumstances of the problem were i affectively disseminated.

b.2 Procurement Documents j Purchase Order (P.O.) S137920, Rev. 3, dated June 14. 1996, controlled the work performed on safety related K-Line breakers including S/N 51817D-211135.

- Attachment 1 Rev. 002, Section C, " Quality Assurance Program Requirements,"

to P.O. S137920 required certain QA program requirements to apply to all rework, replacement part procurement, inspection, testing, handling, storage and shipping of K-Line breakers returned for refurbishment, including 10 CFR Part 50, Appendix B, and 10 CFR Part 21.

Attachment 1, Section E, " Quality Assurance Records," Item 2 to P.O. 5137920 required, in part, the following documentation to be submitted, as applicable:

• ABB Service Certificate of Compliance (CoC).

• Procedural Checklist for Safety-Related Nuclear Switchgear (ABB ,

Service Form QA2). l

• Final Inspection / Acceptance Criteria Checklist (ABB Service Form QA6). '

. Attachment lA, Rev. 000 (01/16/95), " Technical Requirements Safety Related 2

K-Line Breaker 10 Year Refurbishment," to P.O. S137920, Rev. 3, requires in '

part that:~

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A complete visual inspection of all areas of the K-Line Breaker shall be performed to verify configuration complies with factory approved physical assembly configuration... Any visual inconsistency... shall be further investigated. These activities shall be performed utilizing factory approved design documents...

Any observea discrepancies shall be noted and the K-Line Breaker reworked to comply with the specified configuration b.3 K-Line Breaker S/N 51817A-107073 K-Line Breaker S/N 51817A-107073 is a safety related type ABB K30005 breaker.

PNPP sent this breaker to ABB Service for an inspection, lubrication i evaluation, 10-year refurbishment, and repair in accordance with P.O. No. l S137920. ABB Service refurbished and tested the breaker and returned it to '

PNPP with a CoC, dated August 23, 1995. The CoC was signed by the Cleveland Service Center QA Hanager. On September 1, 1995, PNPP put the breaker in service.

On November 8, 1996, PNPP maintenance personnel discovered that control wires connecting the power sensors on the breaker lower molding to the Power Shield trip unit were incorrectly wired. Based on Perry's information, ABB Service l generated Nonconformance Report (NCR) ABB Service Job # 43-02714-26 on November 8, 1996.

The ABB Service NCR stated that during primary current injection testing, Power Shield control wires 11 through 14 are lifted and shorted together.

After testing, the control wires are re-connected to the Power Shield terminal block. The NCR stated that the wiring error discovered by PNPP may have occurred during the re-connection of the control wires following final testing of the breaker. The ABB Service NCR stated that the reversal of the control wires on terminals 12 and 13 of the Power Shield would not have affected proper operation oi the breaker, and therefore tiare was no impact to plant operability. The line item " Potential Part 21 Evaluation Required 7" on the NCR was checked "No" and this response was signed by both the Service Center Manager and the Service Center QA Hanager.

The actions taken by ABB Service to correct this nonconformance and prevent recurrence were to add line items to checklists ABB Service form QA2,

" Procedural Checklist for Safety Related Switchgear," and ABB Service Form QA6, "ABB Service QA final Inspection / Acceptance Criteria Checklist," to require point-to-point wiring checks of all controls including Power Shield, power and phase sensors, and on both ends of the Power Shield wiring harness to verify wiring configuration. The inspectors verified that these corrective actions were implemented by ABB Service with the issuance of Rev. 8 of ABB Service Form QA2 and Rev. 3 of ABB Service Form QA6. Both revisions were dated November 14, 1996.

The inspectors informed ABB Service personnel that failure to perform an adequate inspection of K-Line Breaker S/N 51817A-107073 is a nonconformance contrary to Perry's P.O., and Criterion X of Appendix B to 10 CFR Part 50.

l (This is one example for Nonconformance 99901281/97-01-02).

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b.4 K-Line Breaker S/N 51817C-264135 K-Line Breaker S/N 51817C-264135 is a non-safety-related type ABB K6005 K-Line Breaker, in November 1995, PNPP sent this breaker to ABB Service where it was refurbished and returned to PNPP on September 4, 1996. Upon receipt, the breaker was placed in a PNPP warehouse.

On March 20, 1997, while performing pre-installation checks on the K-Line Breaker, PNPP personnel discovered that the phase-B phase sensor was incorrectly wired. Based on Perry's information, ABB Servico generated NCR ABB Job # 43-02902-1. In this NCR also, the line item " Potential Part 2' Evaluation Required?" was checked "No" and this response was signed by the Service Center Man:ger. TL2 NCR stated that PNPP was assured that the revised ABB Service Forms QA2 and QA6 were being utilized to prevent recurrence of future wiring errors.

From the available documentation, the inspectors could not determine who caused the wiring error. The leads were not reversed at the Power Shield, as in the example above, but at the current sensor terminal block on the lower molding. ABB Service personnel informed the inspectors that these leads are not disconnected during routine refurbishing work unless the current sensors are replaced. The documentation reviewed by the inspector indicated that the current trinsformers were not replaced for this breaker.

ABB Service personnel informed the inspectors that this error was not identified because the breaker was refurbished and shipped before the issuance of Rev. 8 of ABB Service Form QA2 and Rev. 3 of ABB Service form QA6. Because this is a non-safety-related breaker, the inspectors did not identify this condition as a nonconformance. However, the team considered this as an example that should have been evaluated as affecting safety-related breakers since K-Line safety and non-safety-related breakers are identical in this area, b.5 K-Line Breaker S/N 518170-211135 K-Line Breaker S/N 518170-211135 is an ABB K6005 type safety-related breaker.

PNPP sent this breaker to ABB Service for an inspection, lubrication evaluation,10-year refurbishment, and repair in accordance with Purchase Order Number S137920, Rev. 3. On April 16, 1997, ABB Service tested the breaker and returned it to PNPP with a CoC, dated April 16, 1997, signed by the Cleveland Service Center QA Manager certifying that the breaker met the P.O. requirements. On April 28, 1997, PNPP personnel discovered that one of the breaker's power sensors was inverted (i.e., it was installed upside down).

PNPP reported this nonconforming condition to Corporate ABB Service and ABB Service documented this condition in NCR ABB Service Job # 43-02939, dated April 29, 1997. In the NCR, the line item " Potential Part 21 Evaluation Required?" was checked "No" and this response was signed by the Service Center QA Hanager.

7 l

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The inspectors determined from the evailable documentation that this j nonconforming condition was caused by ABB Service during replacement of the a lower molding current transformer assembly. The K-Line Breaker repair k worksheet (ABB Service Form CBRW, Rev. 4., Dated March 10,1995) for breaker h S/N 518170-211135 indicated that on April 2, 1997, ABB Service replaced the '

phase-C lower assembly which contained the upside-down power sensor. The ABB Service Form CBRW for breaker S/N 51817D-211135 was signed and approved by the J Service Center QA Manager on April 15, 1997.

On April 16, 1997, ABB Service issued a CoC for Breaker S/N 518170-211135, signed by the 0/ 1anager, which certified that the repairs and final tests on the subject apparatus, as described in Service and Test Reports, complied with the requirements of the subject purchase order, applicable industry standards, and the specifications of the original contract.

Based on the above, the inspectors informed ABB Service personnel that failure to perform an adequate inspection of K-Line Breaker S/N 518170-211135 is another example of a nonconformance contrary to Criterion X, " Inspections," of 10 CFR Part 50, Appendix B. PNPP's purchase order, and Step 9 of ABB Service Form QA2, Rev. 8, which required that a point to point wiring check be performed on all components, including the power and phase sensors (this is a secono example for Nonconformance 99901281/97-01-02).

c. Conclusions Based on the above, the inspectors determined that ABB Service had not properly developed and implemented an inspection program to assure that refurbished breakers conformed to their original design.

3.2 10 CFR Part 21 Procedure & Evaluations

a. Scope The team reviewed the adequacy of the procedure that ABB Service adopted to implement the provisions of 10 CFR Part 21, Quality Assurance Procedure (QAP) 15.2, " Reporting of Defects and Noncompliance with 10 CFR 21," Revision 2, approved March 24, 1995, and reviewed selected deviations to determine whether they were dispositioned in accordance with the provisions of 10 CFR Part 21.

b. Observations and Findinas b.1 Procedurg The team reviewed QAP 15.2, Revision 2, to assess whether it effectively implemented the provisions of 10 CFR Part 21 r.nd ensured that identified deviations or failures to comply were appropriately evaluated or transmitted to the purchasers. The team noted that revision 2 of QAP 15.2 is eight pages long with an additional seven pages of " supplemental guidance," and has a two page form, " Potential Noncompliance or Defect Report," to be completed by the ABB Service originator.

8

t

'l The inspectors reviewed the procedure as if a deviation had been identified, and determined that the procedure was comprehensive and contained useful notes and guidance to assist the evaluator. The team noted that several of the 4 '

Part 21 definitions were not in accordance with the latest revision of Part 21 and some definitions also contained ABB Service clarifications or interpretations that were integrated into the definition. Although the clarifications a2peared to be helpful, the integration of the ABB Service narrative could be misleading. Therefore, the inspectors recommended that ABB ,

Service assure that the clarifications are discernable from the Part 21

. definitions (such as, by the use of brackets). Additionally, Section 21.3,

" Interpretations," of Part 21 does not allow interpretations of the meaning of l Part 21 verbiage except for written interpretation by the NRC General Counsel.

1 The definitions that were contained in QAP 15.2 appeared to be from the 1995 i

revision. 10 CFR Part 21 was modified in 1996, but QAP 15.2 did not reflect the modifications. Consequently, some of the QAP 15.2 definition: are not in

accordance with the current revision of 10 CFR Part 21.

i The team also reviewed the ABB Service " Potential Noncompliance or Defect

!' Report" form that was an attachment to QAP 15.2 and noted that it contained instructions / guidance that are not appropriate for the audience that would be

! expected to be use the form. The requirements on the form state:

A. Identification of the basic component or activity which contains a " defect" or " failure to M oly."

B. Identification of the Company supplying the basic component or activity which contains the " defect" or " failure to comply."

C. Nature of the " defect" or " failure to comply" and the safety hazard which ir or could be created.

4

0. Number and location of all affected components: (include identification of all purchasers to whom r.omponent has been l supplied)

E. Corrective action, identification of party responsible or action, and schedule for action.

i l The team determined that these requirements may not be fully understood by the average employee / technician. Additionally, the requirements could tend to have somewhat of a chilling effect. The team believes that an employee would not typically have the expertise te identify whether a " defect" or " failure to comply" exists. Therefore, the team felt it possible that an employee would not complete the form since the employee would be unable to provide the required information. The employee's responsibility should be limited to i identifying problems or deviations to their supervision to ensure that all potential defects are identified and dispositioned in accordance with 10 CFR Part 21. This form, if not understood completely by the ABB Service personnel, may have prevented deviations from being reported in the past.

i-9 l

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After discussing the 10 CFR Part 21 regulation, intent nr.d QAP 15.2, with AB1 Service staff, the team informed ABB Service that it hai determined that it had not established an adequate procedure to appropriately implement 10 CFR Part 21. This failure is characterized as a violation of minor significance and will be treated as a Non-Cited Violation, consistent with Section IV of the NRC Enforcement Policy. The ABB Service Director of Quality committed to resolving the NRC concerns.

b.2 PNPP Evaluations The team reviewed licensee and ABB information regarding identified problems with the K-Line low voltage circuit breakers. On three separate occasions, t.s discussed above, PNPP reported that ABB Service refarbished 6nd tested K-line low voltage breakers were found by PNPP to contain wiring or assembly errors.

Although some of the errors could cause a breaker to trip prematurely, in each of the NCRs, the line item " Potential Part 21 Evaluation Re(suired7" was checked "No" and signed by an ABB Service Manager.

Premature tripping, or false trips are a concern especially if false trips occur during transient events when the breaker is required to carry its full current rating. Further, because single rnase calibration testing does not adequately verify the overcurrent trip sat points of some circuit breakers equipped with three phase solid state trip units, the potential exists for incorrectly wired or assembled circuit breakers to pass ABB Service and licensee calibration testing, but prematurely trip during a design basis accident, resulting in a loss of safety function.

The team discussed the disposition of these NCRs with the ABB Service Quality Director. The team asked the Quality Director why there were not any 10 CFR Part 21 evaluations performed. The Quality Director stated that in retrospect, they should have evaluated the issue. regarding breaker S/N 518170-211135, in accordance with 10 CFR Part 21. The ABB Service personnel also informed the inspectors that they did not evaluate the potential that they may have previously shipped safety related K-Line breakers with errors similar to those identified in the NCRs.

The inspectors informed ABB Service personnel that failure to identify potential defects and either evaluate or inform the purchasers was a violation of 10 CFR Part 21. Additionally, recurring current sensor errors in safety-related K-Line breakers in previously shipped safety-related K-Line breakers was a potentially generic problem that ABB service did not recognize, evaluate or inform ABB Service customers so its customers could determine if the condition existed at their facilities. Violation 99901281/97-01-01 was identified in this area, b.3 ABB Service Evaluation The team also reviewed an ABB Service report performed for Public Service Electric and Gas Company's Salem nuclear generating station (Salem) by the ABB Service Company's Mount Laurel, New Jersey facility. The report, " Harmonic Measurements and Circuit Breaker Tripping Evaluation," dated December 1992-January 1993, was requested by Salem due to unexplained K-Line circuit breaker 10

tripping during starting and normal operating conditions. The report stated the existing circuit breaker had been tested by single-phase fault simulation injection, and all tests indicated proper tripping functions; therefore, the

cause of the premature tripping was unide.ntifiable by the licensee.

Consequently, the licensee contracted ABB Service for further investigation.

The ABB Service report stated in part:

• The computer and laboratory simulation identify the effects of improper current sensor connections to the solid state trip unit.

The results clearly indicate the potential for premature tripping, if in roper current sensor polarity wiring exists.

• It appears that the error in polarity may have existed prior to shipment from the ABB manufacturing factory....

• NOTE: Single-phase overload testing is necessary, but will NOT identify improper polarity of the... current sensor connections.

• The (ABB Service) laboratory simulation confirmed that the mathematical results which clearly indicate that the reversal of one phase will result in an artificially high input into the logic box

, sensing circuitry. Two times normal. to be specific. This would cause premature tripping....

4

• The comparison of the field measurements and the circuit breaker limitations, with one phase reversed, clearly indicate that having 3

one phase reversed was a contributing factor to premature tripping during normal loading and inrush operation.

• In addition, if ABB type K-Line circuit breakers have been serviced, by in-house (licensee) personnel... these circuit breakers :hould be identified and have the polarities of the current sensors checked i during the next scheduled outage... If any of the identified circuit breakers supply loads whereby premature tripping is of an operating or safety concern, then provisions should be made to allow these circuit breakers to have the polarities checked as soon as possible.

The team determined that although ABB Service personnel were aware of this problem as early as January 1993, ABB Service failed to inform its customers of the potentially generic latent defect that could have existed on any new or refurbished ABB K-Line breakers,

c. Conclusions c.1 Procedure The team concluded from its review of QAP 15.2, Revision 2, and its attachments that the ABB Service procedure was cumbersome, contained outdated 10 CFR Part 21 definitions, and could have resulted in misleading ABB Stryice i employees into believing that they must perform a review of the circumstances 4

) 11 surrounding a deviation to determine the safety-significance and the safety hazard which could be created. Therefore, the team concluded that the current

revision of QAP 15.2 would not effectively implement all of the provisions of 10 CFR Part 21.

c.2 PNPP Evaluations a Tne team concluded that ABB Service failed to either perform an evaluation of the potential reversed polarity and miswiring or to inform all of its

[ customers so they could determine if a problem existed.

c.3 ABB Service Evaluation lhe team concluded that although ABB Service was aware of the potential for reversed polarity on new or refurbished K-Line low voltage circuit breakers as early as January 1993, ABB Service failed to act appropriately given the T' potential problems that could have ensued at operating nuclear power plants as j a result of the potentially generic K-Line series matter.

g 4. PERSONS CONTACTED ABB Service Company b

E J.J. Connolly, Vice President, Business Development D.E. Leckey, Quality Assurance Manager, Cleveland Service E. Link, Manager, Cleveland Service J.M. Tate, General Manager, North Central Region J.0, Webb, Director of Quality Centerior Eneroy Comoany M.R. Fournier, Quality Engineer, PNPP Nuclear Assurance l ITEMS OPENED, CLOSED, AND DISCUSSED Ooened 99901281/97-01-01 NOV inadequate evaluation / failure to inform 99901281/97-01-02 NON inadequate inspections Closed 99901281/94-01-01 URI unavailability of organization chart 99901281/94-01-02 URI unavailability of job descriptions 99901281/94-01-03A NON inadequate qualitative criteria 99901281/94-01-03B NON acceptable contact resistance not available 12

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

p2 8tc ,

p*  % UNITED STATES j,\

W j NUCLEAR REGULATORY COMMISSION .

WASHINGTON. D.C. 20555-0001 9,

• f August 29, 1997 Mr. Wilfred C. LaRochelle, Managst, Quality Assurance Hartford Steam Boiler inspection and Insurance Company One State Street P.O. Box 5024 Hartford CT 06103-3102

SUBJECT:

NRC INSPECTION REPORT 99900601/97-01 AND NOTICE OF NONCONFORMANCE

Dear Mr. LaRochelle:

On July 18,1997, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection of Hartford Steam Boiler inspection and insurance Company offices in Hartford, CT and Atlanta, GA. 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 with NRC requirements imposed on you by your customers and with American Society of Mechanical Engineers (ASME) requirements that are applicable to your activities under the scope of your ASME Certificate of Accreditation. Specifically, the NRC inspectors determined that internal audits of the Home Office and regional office Engineering Services activities had not been conducted at the required intervals, and that nonconformity reports were not issued to document, correct, and disposition the findings of those audits that had been performed. A contributing factor to the identified conditions appeared to be a lack of controlled procedures or instructions for the performance and documentation of audits or for the disposition of audit findings.

Additionally. the inspectors identiCad that the qualification files for two lead auditor eandidates did not contain adequate docuns.+ntation to support the point scores c' signed on the basis of their nuclear industry experience.

These nonconformances are cited in the enclosed Notice of Nonconformance (NON),

and the circumstances surrounding them are descr; bed in detail in tho enclosed report.

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

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

Sincerely,

. ORIGINAL SIGNED BY GREGORY CWALINA FOR:

Stuart A. Richards,-Chief SpecialInspection Branch Division of Inspection and Support Programs-Office of Nuclear Reactor Regulation-Docket No:. 99900601

Enclosures:

1. Notice of Nonconformance-

2. Inspection Report 99900601/97 01 1

NOT.0E OF NONCONFORMANCE Hartford Steam Boiler Inspection Docket No.: 99900601 and Insurance Company Hartford, CT Based on the results of an inspection conducted on July 14 through 18,1997, it appears that certain of your activities were not conducted in accordance with NRC requirements imposed on you by your customers, or with the requirements of The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (ASME Code) that are applicable to your activities under the scope of your Certificate of Accreditation.

A. Criterion XVill, " Audits," of Appendix B to 10 CFR Part 50 requires, in part, that a comprehensive system of planned and periodic audits shall be carried out to i verify all aspects of your quality assurance program and to determine the effectiveness of the program.

l Engineering Services Manual (ESM) Chapter 4400, " Audits," states, in part, in i Section 4421, that the Quality A*,surance Manager is responsible for annual audits of the Regional Manager, Engineering Services (RMES) activities and that the Internal Audit Department shall audit the activities of the Home Office (HO)

Engineering Services (ES) department ASME activities.

Contrary to the above,

1. Hartford Steam Boiler and Insurance Company (HSB) could only provide documented evidence that one audit (November 1995) of HO ES activities had been performed during the last five years. The HSB Quality Assurance Manager (QAM) stated that the Internal Audit Department no longer conducts audits of HO ES ASME activities and no other HSB organization or department has been assi0ned that responsibility.

2. HSB did not perform the annual audits of RMES activities for San Francisco (1996), Atlanta (1995), and Northeast / Philadelphia (1994 and 1995).

(Nonconformance 99900601/97-01-01)

B. Criterion XVI," Corrective Action," of Appendix B to 10 CFR Part 50 requires that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected.

, ESM Chapter 4300, ' Control of Nonconformities," states, in Section 4310, thet

'5is Chapter outlines the requirements for the identification, documentation, and du, position of nonconformances to the ESM, supporting procedures or instructions.

Enclosure 1

Section 4340, states, in part, that the RMES/QAM shall have a nonconformity report prepared whenever a nonconformance is identified.

Contrary to the above,

1. Nonconformity reports were not issued to document, correct, and disposition the six audit findings that were identified and documented as part of the November 1995 Internal Audit Department audit of the HO ES ASME activities.

, 2. Nonconformity reports were not issued to document, correct, and disposition all of the audit findings identified during the HO ES QAM audits of RMES.

The inspectors determined there was only one instance (1996 Atlanta regional office audit) where the RMES used a nonconformity report to

! document and disposition the audit findings that were identified during the OAM's annual audit of RMES activities. (Nonconformance 99900601/97-01-02) l l C. Section'.1-1, "The Autt.orized Inspection Agency," Subsection 1-1.2, " Duties,"

Paragr.ph 1-1.2.4 of ASME QAl-1-1995, requires, in part, that the agency shall establir h and implement an internal program which shall provide assurance that those of its employees holding the positions of supervisor or authorized nuclear inservice inspector (ANil) perform work in accordance with the requirements of Part 1 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 XI work, in accordance with these policies, procedures, or instructions.

Contrary to the above,

1. The ESM, which documents the requirements necessary to perform ASME Code and engineering service activities, did not include or reference the

" Engineering Service Audit Checklist" used by the HO ES QAM to perform annual audits of the RMES activities.

, 2. The " Engineering Services Audit Checklist," which is used by the HO ES QAM to perforrr % e annual audits of RMES activities, did not include any provisions for revis . ving the disposition and corrective actions implemented for findings identified during past audits of RMES activities.

2

3. No implementing procedure existed to control the internal audit process, and the ESM did not include guidance for conducting quality activities such as documenting internal audits, audit findings, and their closure.

4. National Board forms NB-71 and NB-178, " Audit Verification Record," are 3

referenced in Sections 4471 and 4475 of the ESM and used by HSB as they method to notify the National Board of completion of required audits, but are not included in Section H, " Forms," of the ESM. (Nonconformance 99900601/97-01-03)

D. Section 1-2, "The Authorized Nuclear Inspection Supervisor," Subsection 1-2.2,

" Duties," Paragraphs 12.2.6 & 1-2.2.7 of ASME QAl-1-1995, require, in part, that the anils shall audit the performance of each ANil under his supervision on a planned and periodic basis. Each ANll actively engaged in Section XI Code inspection shall be audited at least twice a year at the site to which he is assigned. The audit shall be recorded in writing and shall contain a written comment regarding the status of each item audited.

Contrary to the above, i

1. The HSB ESM did not include provisions that require the RMES to document and adhere to a schedule of two annual audits of each ANil.

2. The audits conducted by the Atlanta region RMES of the assigned ANil performance did not contain written comments regarding the status of each item audited. Documented objective evidence consisted of a check for either satisfactory, unacceptable, not observed, or not applicable.

(Nonconformance 99900601/97-01-04)

E. 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 "-th these instructions.

Section C-5," Experience," of Appendix C," Qualification of Lead Auditors," to the ESM, permits 9 points maximum " experience" to be credited towards lead auditor qualification and states that time spent in various activities will be awarded points on a reasonable basis in line with ANSI N45.2.23 and NOA-1, Appendix 2A-3. This section of the ESM also contains a provision to score one (1) point maximum for each full year's experience classified as " Industry" with other companies if it meets the requirements of Paragraph 2.3.1.2 of ANSI N45.2.23 and Paragraph 2.2 of Appendix 2A-3 of NQA-1.

3

l Section 2.3.1.2 of ANSI N45.2.23 states, " Experience (9 points maximum).

Technical experience in engineering, manufacturing, construction, operation, or l maintenance, score one (1) credit for each full year with a maximum of five (5) credits for this aspect of experience."_ Section 2.3.1.2 continues by providing guidance on scoring additional points for specific nuclear, quality assurance,. ,

and auditing experience. Similar provisions are contained in NQA-1.

Contrary to the above, two lead auditor candidates were credited the maximum i of points (5) for 5 years of work experience towards lead auditor qualification without any objective evidence that the experience provisions contained in Section 2.3,1.2 of ANSI N45.2.23 or Appendix 2A-3 of NQA-1 had been met.

(Nonconformance No. 99900601/97-01-05)

Please provide a w.itten statemeni 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 descr5 tion of steps that have been or will be taken to correct these items; (2) a description d steps that have been or will be taken to prevent recurrence; and (3) the dates your corrective actions and preventative measures were or wili be completted.

Dated at Rockyille Maryland thisN%ay ofWld,997 4

U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION 1 1

Report No: 99900601/97-01 Organization: Hartford Steam Boiler Inspection and Insurance Company One State Street Hartford, CT 06103-3102

Contact:

Wilfred C. LaRochelle, Manager Corporate Quality Assurance i

Nuclear Industry Activity: Authorized Nuclear inspection Agency Dates: July 14 - 18,1997 Inspectors: Uldis Potapovs, Senior Reactor Engineer Richard P. McIntyre, Senior Reactor Engineer Larry L. Campbell, Reactor Engineer Approved by: Gregory C. Cwalina, Chief Vendor Insoection Section Special inspection Branch-Division of Inspection and Support Programs Enclosure 2 25-

.1 INSPECTION

SUMMARY

l During this inspection, the NRC inspectors reviewed the implementation of selected portions of Hartford Steam Boiler inspection and Insurance Company's (HSB) quality l assurance (QA) program for providing third party inspection services to NRC licensees. ,

The first part of the inspection was conducted at HSB home offices (HO) in Hartford, ,

1 CT and included a review of the corporate organization structure,10 CFR Part 21 Implementation progra, and HO responsibilities for audits, training and qualification, and nonconformity control. The second pad of the inspection was conducted at the HSB regional office in Atlanta, GA and focused en the control and oversight of plant site activities related to the implementation of American Society of Mechanical Engineers (ASME) inspection responsibilities.

The inspection bases were:

. Appendix B, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Processing 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"

. Section lil, " Rules for Construction of Nuclear Power Plant Components" of the Asi1E Boiler and Pressure Vessel Code (Code)

. ASME QAl-1-1995, " Qualifications for Authorized Inspection" - .

During this inspection, two minor violations of NRC requirements were identified and are discussed in Section 3.2 of this report The inspection also identified 5 instances where HSB failed to conform to NRC requirements imposed upon them by NRC lic.ensees. These nonconformances are discussed in Sections 3.3.2 and 3.3.3 of this-repod.

l o

2 STATUS OF PREV'OUS INSPECTION FINDINGS This was the first NRC inspection of HSB activities performed under an ASME Certificate of Authorization.

3 INSPECTION FINDINGS AND OTHER COMMENTS 3.1 Descriotion of Facilities and Activities HSB is an Authorized inspection Agency (AIA) for performance of AIA activities controlled from One State Street, Hartford, CT for the following ASME Codes:

2 l

. ASME Section 'lli, Divisions 1 and 2, ASME Section XI, and ASME Sections I, IV, Vill, Divisions 1 and 2, and X. All inspection services provided to NRC licensees, including contract administration, are conducted through five regional .

offices, At the present time, HSB has services contracts with 27 nuclear utilities, l at 44 plant sites.

~

1 3.2 - 10 CFR Part 21 Frocram and its Imolementation HSB's procedures for implementing 10 CFR Pad 21 regulations are described in i

Appendix D, "NRC 10 CFR 50 Part 21," of the Engineering Services Manual (ESM). The current revision dates of the ESM and of Appendix D were December 20,1995, and October 27,1995, respectively.

Appendix D provided definitions of basic components, defects, and deviations, - ,

and required any person involved in nuclear inspection or supervision of such work, who notes a potentially reportable condition, to bring this condition to the attention of a responsible authority at the shop or plant site. -The individual observing the potentially reportable condition was also required to document it in his bound diary and provide details to the Regional Manager of Engineering

- Services (RMES) who, in turn, was required to forward the information to the 2nd Vice President, Engineering Services (2nd VPES). From that point on, HSB's reporting obligations were to be handled by the 2nd VPES, The employee '

originally reporting the condition was also required to send a written report, describing the corrective action to his supervisor, who was required to_ forward this information to the 2nd VPES.

= Appendix D noted that the provisions of the 10 CFR Part 21_ also apply to the company's authorized nuclear inspectors (Anis), and any defects or deviations in their work which may lead to a substantial safety hazard should be considered reportable. Appendix D also stated that, except in those cases which are clear -

and evident, it is not the Company's intent to require the inspectors to evaluate when a deviation may result in a substantial safety hazard, implying that all potentially reportable conditions are evaluated by the 2nd VPES. The following concerns were identified as a result the inspector's review of Appendix D.

(a) Appendix D was based on, and referenced a superseded revision of 10 CFR Part 21 (October 21,1991). Consequently, certain definitions quoted in tha procedure were not consistent with the current revision of the regulation.

(b) Although Appendix D defined a deviatica as a departure from the technical requirements, and required such conditions to be reported to the 2nd VPES, it did not require that nonconformity reports be considered for 3

potential reportability. Since ESM Chapter 4300, ' Control of Nonconformities," defined nonconformance as a deficiency in documentation, procedure, or instruction that renders an activity unacceptable or indeterminate, to comply with the requirements of 10 -

CFR 21.21, such conditions would need to be evaluated for reportability.

(c) Although the procedure stated that the 2nd VPES will handle the Company's reporting obligations, it did not specify time limits associated with these oblig,t5ns (evaluation, initial notification, interim and final reports, _etc.) -

Based on a record review and discussion with HSB management, the inspectors determined that no potentially reportable conditions had ever been identified and -

forwarded to the 2nd VPES, and that nonconformity reports, generat*d as a result of regional office or home office operations, had not be.en considered as potentially reportable conditions (ESM Chapter 4300 does not require that nonconformances be evaluated for potential-10 CFR Part 21 reportability).

The inspectors also determined that neither the HSB Home Office nor the Atlanta Regional Office had complied with the posting requirements specified in 10 CFR 21.6, " Posting Requirements," which state that each organization subject to the regulations in this Part shati post current copies of the regulation, Section 206 of the Energy Reorganization Act of 1974, and procedures adopted pursuant to the regulations in this Part in a conspicuous position on any premises within the United States where the activities subject to this part are -

conducted. None of the documents cited above were posted in HSB's offices.

The inspectors advised HSB management that failure to _ comply with the posting 4

requirements as discussed above and failure to require that nonconformances are evaluated for potential reportability would be identified as violations of 10 CFR Part 21. However, these failures constitute violations of minor signifiw. ice and are treated as Non-Cited-ViGiions, consistent with the NRC Enforcement Policy. During the inspection, HSB management revised Appendix D in -

response to the concerns identified in 3.2(a), (b), and (c), above, and committed -

to provide the required postings at their Home Office and all regional offices.

3.3 Quality Assurance Procram and its Imolementation The program used by HSB to control inspection services provided to NRC licensees is described in their ESM. The ESM commits to providing inspection -

services consistent with the requirements of the ASME Code and ASME QAl-1 series standards. Although the reouirements of 10 CFR Part 50, Appendix B, are invoked through purchase orders (PO) of several licensees, the ESM does 4

not specifically commit to compliance with this regulation. While a detailed review of programmatic compliance with 10 CFR Part 50, Appendix B, was not made during this inspection, HSB's control of the implementatica of selected safety related services was reviewed against the requirements of the ESM as well as the criteria of 10 CFR 50, Appendix B.

3.3.1 ASME Insoection Resoonsibilities

a. Insoection Scooe The NRC inspectors reviewed ESM Chapter 4500, 'ASME Inspection Responsibilities" and the authorized nuclear inservice inspector (ANil) bound

- diaries, inspection logs, and records of QA monitoring activities for selected j- nuclear plants to assess the program controls and their implementation.

i l

b. Observations and Findinos Chapter 4500 of the ESM describes HSB policies and responsibilities for regional managers, supervisors, and inspectors involved in ASME inspections.

Each RMES is assigned the responsibility for assuring that the inspection services meet all specified standards of performance and quality. The RMES is also responsible for the administration of service contracts with the licensees within his jurisdiction.

The direct supervision of Anils and audits of their performance is the responsibility of the authorized nuclear inservice inspector supervisors (anils).

Chapter 4500 also describes the duties and responsibilities of various categories of authorized inspectors consistent with the provisions of ASME -QAl-1.

The implementation of ASME inspection activities was evaluated by reviewing ta:avant documentation (ANil bound diaries, monitoring schedules and reports, qualification records, etc.) _for selected nuclear plants and, in some cases, discussing specific issues with the assigned ANil by telephone.- The NRC inspectors reviewed the ANil bound diaries for Comanche Peak, covering the period from April 10,1995, to March 21,1997; Browns Ferry, Unit 3 (November 22,1995, to April 21,1997); and Hatch (August 16,1995 to October 21,1995). Also reviewed were selected monitoring schedules and reports, records of inspection verifications, and identification and disposition of nonconforming conditions.

5 l

The ANil bound diaries,- in general, were found to contain appropriate entries, consistent with the applicable ASME QAl-1 requirements. Records of n'onitoring activiues indicated that these activities were performed in ;ccordance with the schedules established by the Anils, using supplementary checklists (developed l by the ANil), to identify specific program areas to be monitored and to provide for recording of objective evidence to support their findings for each area monitored. Program deficiencies identified as a result of the monitoring activity -

were recorded in the ANil's bound diary and documented on HSB form 939, "ES Record for Monitoring QA/QC Programs." Form 939 is used to identify specific program areas reviewed, and, in cases of identified deficiencies, to request a response from the NRC licensee. Review of several monitoring records irdicated that these activities were properly documented, and that identified concems and their resolution were documented in the ANil's diary.

A generai observation was that activities described above were performed end

~ documented using different methods at the selected plant sites reviewed,

- apparently because there were no standard implementing procedures available to perform several of these activities. Similarly, it was also noted that tracking -

and resolution of identified concerns was being addressed by different methods.

In some cases, HSB Form 939, "ES Record for M.on toring QA/QC P. grams" was used for all identified concerns,-while, at another plant site, the ANil had apparently developed and was using an "ASME XI Discrepancy Notice" for tracking and dispositioning of isolated (non-programmatic) deficiencies. These forms were being issued to the licensee and dispositioned after achieving resolution of the issue.

c. Conclusions Review of the records of ASME activities performed at selected plant sites indicated that these activities were conducted and documented in accordance -

with the applicable ASME QAl-1 requirements and that the ANil records of these acNities were generally well documented and complete. One observation in t! .s area was discussed with HSB management as a program weakness. This related to the lack of controlled procedures for performing and documenting monitoring activities and dispositioning of isolated (non-programmatic) deficiencies.

3.3.2 Ag[tta

a. Scope The inspectors reviewed Chapter 4400, " Audits," of the ESM, w's .h described the audit and survey requirements for intemal audits, extemal audits, pre-review /

6

l survey audits, and ASME Code required audits, and also described the accompanying requirements for audit report documentation. The following Sections of Chapter A 100 of the ESM were reviewed: I o Section 4420, " Internal Audits"

• Section 4430, " External Audits" o Section 4470, " Nuclear Audits'

b. Observations and Findinas b.1 Intemal Audits - Home Office Enaineerina Services l The Intemal Audit Department (IAD)is responsible for conducting annual audits

of HO Engineering Services (ES) department's ASME activities. The Quality. l l Assurance Manager (QAM) is responsible for conducting annual audits of the various RMES activities to verify compliance with the ESM, supporting procedures, instructions, and the ASME QAl-1 standards, as applicable.

The inspectors reviewed implementation of the above intemal audit processes for compliance to the ESM. While attempting to review LAD's audits of HO _ES activities, the inspectors determined that only one audit (November 1995) had been performed and documented by LAD and was available for review.; This IAD audit identified six findings and recommendations for compliance to the ESM.

The inspectors were told by the QAM that IAD no longer conducts the audits of HO ES activities even though ESM Section 4421 still requires this activity. The failure to audit HO ES activities as required by the ESM was identified as an example of Nonconformance 99900601/97-01-01.

The inspectors also identified that nonconformance reports had not been written by the QAM to document, disposition, and correct the findings that were

-identified during the LAD audit conducted in November 1995. Section 4320 of Chapter 4300 of the ESM, ." Control of Nonconformities," defines a-nonconformance as a deficiency in documentation, procedure or instruction that renders an activity unacceptable or indeterminate. Section 4340 states that the RMES/QAM shall__ prepare a nonconformity whenever a nonconformance is identified. The failure to issue Nonconformance Reports for documented failures to implement the requirements of the ESM was identified as Nonconformance 99900601/97-01-02.

b.2 Intemal Audits - Reaional Manaaer. Enaineerina Services Allintemal audits of the RMES activities are conducted by the QAM. Currently there are four regional offices that conduct ASME Section lli and Section XI 7

w nuclear inspection activities. The inspectc,rs reviewed the schedule for audits of RMES activities for the last four years to select a sample of audit reports to review. When revit, wing the regional office audit schedules, the inspectors noted that many of the scheduled audits had slipped beyond their scheduled date, including the 1996 audit of the San Francisco RMES and the 1995 audit of the Atlanta RMES, which were never performed. While attempting to review a sample of audit reports from each region, the inspectors were told that several of the documented audit reports and the RMES responses to audit report findings

< requested could not be locded in the HSB HO or Regional files for Atlanta, Northeast / Philadelphia, and San Francisco in the 1993 to 1995 time frame.

During the inspect'on the QAM committed to issue a Nonconformance Report to address this issua. The failure to audit all RMES activities on an annual basis as required by the ESM was identified as another example of Nonconformance 99900601/97-01-01.

The audit reports reviewed by the inspectors documented what appeared to be a thorough review of the RMES activities and included pertinent findings against program implementation, when applicable. However, as was the case with the audit findings identified during the IAD audit of the Home Office Engineering Services ASME activities, no nonconformance reports or appropriate tracking and completion of corrective actions were oocumented by either the QAM or by

-the RMES. The inspectors determined there was only one instance (1996 Atlanta regional office audit) where the RMES documented and dispositioned the audit findings on nonconformance rer,rts. This issue was identified as another example of Nonconformance 99900f.~1/97-01-02.

When reviewing the audit reports that were available the inspectors determined

, that the Engineering Services Audit Checklist used by the QAM for RMES audits was not an approved and controlled quality document ano was not referenced in ESM Section 4420 or in Appendix H, " Forms." During the 1 inspection HSB committed to approve and control the checklist under the ESM j program requirements. This issue was identified as an example of '

Nonconformance 99900601/97-01-03.

The inspectors also noted that Section 4420 of the ESM does not include in-depth guidance and detail for conducting and documenting internal audits and i accompanying audit findings and that there was limited documented QA program requirements for accomplishing this quality activity. No implementing ?rocedure existed that addressed the internal audit process. This issue was identified as another example of Nonconformance 99900601/97-01-03.

8 The inspectors also determined that none of the audit reports reviewed contained documentation of any follow-up review for corrective actions and

' disposition to previous audit findings, and that neither the ESM nor the checklist used for the QAM audits of the RMES included provisions that required -

evaluation and review of the closure of audit findin;'s identified during previous audits. - During the inspection and in follow-up documentation submitted to the NRC on July 25,1997, HSB committed to implement Regional Office Audit Procedure, ES QP 03, Revision 0, to address these concerns.

b.3 ASME Nuclear Audits I l

Section 4470 of the ESM documents HSB requirements for the conduct of audits to comply with ASME QAl-1,1995, ASME Section Ill, Division 1, and ASME Section XI. This included ANI and ANil audits of nuclear Section lll and Section L

XI work activities, and ANI and ANil performance audits conducted by the '

authorized nuclear inspector and inservice inspector supervisors (ANIS and anils). Chapter 4500 of the ESM,"ASME Inspection Responsibilities,"

documents the inspection responsibilities (including audits) for the RMES, ANIS, and the anils.

The inspectors reviewed the current ANI " Nuclear Shop _ Assignments" and

" National ANil Assignments

• listing for ASME Section til shops in the Atlanta region and Section XI plant sites (including Atlanta region).' The ASME Section -

lli shop and Section XI site performance audits were reviewed for ASME Code

- compliance and for the implementation of the various sections of the shop's and site's applicable QA manuals. The nuclear shop ANI and ANil performance audits are required to be performed twice a year and are conducted using the

- ES Inservice Report checklist (Form 2163). National Board forms NB-71 and -

NB-178, " Audit Verification Record," forms are referenced in Sections 4471 and -

4475 of the ESM and are used by HSB as the method to notify the National Board of completion of these required audits of ANI and the ANil by the ANIS and the anils. These forms are not included in Section H,

• Forms," of the ESM.

This issue was identified as another example of Nonconformence 99900601i.7-01-03.

The inspectors reviewed a sample of semiannual inspector's performance audits conducted by the ANIS at Section lll nuclear shops and by the anils at Section -

XI sites. The audit reports included the appropriate res,ew and documentation as required by the HSB ESM with the exception _of an issue pertaining to audit requirements included in Sections 0-2.2.7 and 1-2.2.7 of QAl-1,1995. These paragraphs require the audits to be recorded in writing and to contain a written comment regarding the status of each item audited. The audits reviewed by the inspectors were accomplished using the Form 2163 checklist described above, 9

l but did not contain a written comment regarding the status of each item audited or any real documented objective evidence of what was reviewed. This issue was identified as Nonconformance 99900601/97-01-04.

During the above review, the inspectors also identified that the ESM does not include any requirement for the ANIS and the anils to document and implement an audit schedule for their applicable sites on an annual basis. This issue is identified as another example of Nonconformance 99900601/97-01-04,

c. . Conclusions Based on the above, the inspectors concluded that Hartford had failed to implement an adequate audit program.

3.3.3 Qualificaticn ofl.ead Auditors

a. Scope The NRC inspectors reviewed five HFa lead auditor qualification fiies to determine if the provisions container,in ESM Appendix C," Qualification of Lead Auditors" were veing met.

b. Observations and Findinas HSB uses lead auditors, qualified in accordance with ESM Appendix C, to perform internal audits of the implementation of its quality assurance program.

HSB also uses these lead auditors to perform safety-related, non-ASME Code activities, such as independent reviews, assessments, and audits.

The lead auditor qualification provisions contained in Appendix C of the ESM are based primarily on the adoption of ANSI N45.2.23-1978," Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants," and NOA 1989, " Quality Ass. rance Program Requirements for Nuclear Power Plants."

The inspector's review revealed that Appendix C contained a provision that was not fully consistent with Section 3.2," Maintenance of Proficiency," of ANSI N45.2.23. ANSI N45.2.23 permits lead auditors to maintain their proficiency by several methods including the review and study of codes, standards, procedures, instructions, and other documents related to quality assurance programs and auditing. However, Section C-13. " Annual Recertification," of Appendix C to the ESM states, in part, that in lieu of audit participation, the lead auditor may demonstrate that his proficiency has been maintained by having actively participated as a member in an ASME Code Committee that reports to 10

i the Board on Nuclear Codes and Standards. The NRC inspectors and HSB ,

discussed the fact that even though a lead auditor participates on a committee i that reports to the Board on Nuclear Codes and Standards, this does not ensure that the lead auditor's participation in the committee process requires a review and study of quality assurance programs and the auditing process. It was also discussed that depending on the committee or subcommittee, the lead auditor -

may not be involved in committee activities associated with the auditing proce(s.

The NRC inspectors considered the alternative to audit participation (having

participated on a committee reporting to the Board on Nuclear Codes and  ;

Standards) to be a potential area of deviation from ANSI N45.2.23 and a j weakness in the HSB lead auditor qualification program, i i The NRC inspectors' review of the five HSB lead auditor qualification files idennfied that 3 of the qualification packages did not appear to meet the provisions of Appendix C. Specifically, Section C-3," Qualification of Lead Auditors," requires that the lead auditor shall have verifiable evidence that a -

minimum of 10 points under the score system provided in Appendix C have been acumulated. The maximum number of points permitted for various categories are: a) Education,4 points, b) Experience,9 points, c) Professional Accomplishments, 2 points, and d) Management, 2 points.

Section C-5, " Experience," of Appendix C to the ESM, permits 9 points maximum for related experience and states that time spent in various activities will be -

awarded points on a reasonable basis in line with ANSI N45.2.23 and NOA-1, Appendix 2A-3. Further, Section C-5 contains a provision to score one (1) point maximum for each full year's experience with other companies in other capacities classified as " Industry"if it meets the requirements of Paragraph 2.3.1.2 of ANSI N45.2.23 and Paragraph 2.2 of Appendix 2A-3 of NQA-1.

Section 2.3.1.2 of ANSI N45.2.23 states:

Experience (9 points Maximum). Technical experience in engineering, manufacturing, construction, operation, or maintenance, score one (1) credit for each full year with a maximum of five (5) credits for this aspect of experience.

Section 2.3.1.2 continues by providing guidance on scoring additional points for specific nuclear, quality assurance, and auditing experience.

The review indicated that HSB had credited each of three lead auditors with 5 -

points for work experience, based on two of the lead auditors having 5 years experience in the US Navy and one having 5 years experience with another company. The NRC inspectors questioned whether any of the 5 years experience was technical, or was in the nuclear or in the quality assurance or 11 i

auditing disciplines. HSB presented a resume of one of the thtee lead auditors, 2 HSB Identification Number 02562, which indicated for the per,od 1972-1977, the l individual had 4 years experience as a radiographer and welder and 1 year.

! experience as a. quality control manager. No records or resumes were

. presented for the other lead auditors (Identification Numbers 02026 and 02664).

- c. Conclusions The NRC inspectors determined that the lead auditor files for HSB employees l Identification Numbers 02026 and 02664, initially qualified as lead auditors on

! June 28,1991, and January 31,1996, respectively, were inadequately 4

processed by crediting the maximum of points for work experience without any

j. objective evidence that the provision = contained in Section 2.3.1.2 of ANSI N45.2.23 or Appendix 2A 3 of NOA-1 had been met. This issue was ;dentified -

as Nonconformance 99900601/97-01-05, 3.4 Service Contract Provisions and Acolicability of 10 CFR Part 21 to HSB Activities

a. Scope The NRC inspectors reviewed several licensee purchase orders (POs) issued to HSB for Authorized Inspection Agency inspection services to assess licensee control of subcontracted inspection activities and the extent to which applicable NRC requirements are passed down to the providers of these activities,

b. Observations and Findinas.

The review identified that certain NRC requirements were not consistently invoked by licensees using HSB to perform third party inspections required by the ASME Section XI Code. Specifically:

(a)- Virginia Power PO BKl 483582, dated November 11,1995, stated that the HSB inspection services were nuclear safety related and required HSB to implement quality control and quality assurance programs that comply with the requirements of Appendix B to 10 CFR Part 50 and ANSI N45.2.

The Virginia Power PO also invoked the requirements of 10 CFR Part 21 for this contract.

(b) Section 26," Quality Assurance Requirements" of Carolina Power & Light

- Company (CP&L) Document UFl No. PTC00004, " Contract No.

XT00000026 between Carolina Power & Light Company and the Hartford Steam Boiler Inspection and insurance Co.," hand dated April 5,1990, 12

-3 6-

i stated that the work to be performed by HSB had been determined to be non-nuclear safety related. , in Soction 27, "10CFR21," the C.P&L contract stated that the provisions of Part 21 shall apply to any work within the definition of basic component in 10 CFR 21.3 and that CP&L shall be promptly notified of any reports made to the NRC pursuant to 10 CFR 21.21. However, Section 27 of the CP&L contract continued by stating that CP&L recognizes that the Contractor will not be required to perform nuclear. safety related work.

(c) Georgia Power Co. PO 60120550000 (latest revision dated May 21, 1997) did not invoke the requirements of either 10 CFR Part 50 or 10 CFR Part 21, but required the labor arcJ supervisory personnel to meet the requirements of ANSI N18.1, and the AIA to meet the requirements of ASME 626.1-1982.

(d) Duke Power Co. PO MN 12553, dated March 12,1996 (for McGuire Nuclear Station) stated that the services to be supplied are safety related i and that they shall be supplied in accordance with the suppliers quality assurance program, approved by Duke Power Co. It also stated that, if lower tier procurement is required, the applicable QA requirements must also be invoked on the lower tier subcontractors / suppliers.

The NRC inspectors and HSB discussed the applicability of both Appendix B to 10 CFR Part 50 and 10 CFR Part 21 to the inspection services provided by HSB at nuclear power plants. During these discussions the NRC inspectors identified that Section 21.3(1)(ii)(3) of 10 CFR Part 21 states: "In all cases, basic component includes _ safety related design, analysis, inspection, testing, fabrication, replacement of parts, or consulting services that are associated with the component hardware whether these services are performed by the component supplier or by others." It was also discussed that Section 50.55a,

" Codes and Standards," of 10 CFR Part 50 identifies the applicable codes for tha design, fabrication, erection, construction, testing, and inspection of systems, structures, and components and mandates the use of certain editions and addenda of the ASME Section XI Code.

it was further discussed that because Section IWA 2110 " Duties of the inspector," of Article IWA 2000, " Examination and Testing," of the ASME Section XI Code identifies the duties of the Authorized Inspection Agency's inspector (the inspector) assiped to perform IWA 2110 activities and these activities are mandatory by the ASME Section XI Code in order to assure compliance to the code, they are considered to be safety-related activities. It was further 13

discur. red that because the Inspector's performance of certain verifications and review'* is a safety-related activity, the services of the inspector are considered a basic component.

The inspectors also determined, from discussions with HSB management, that no licensee audits of HSB ASME authorized inspection activity implementation had ever been performed at any of the HSB offices.

c. Conclusions The inspectors determined that major inconsistencies existed in licensee safety classification of the services provided by HSB and in the imposition of applicable NRC requirements (10 CFR Part 21 and Appendix B to 10 CFR Part 50) for the performance of these services. The inspectore also determined that licensees were not performing QA program implementation audits on HSB as a provider of safety-related services.- 3 3.5 Entrance and exit meetinos in the Entrance Meeting on July 14,1997, the NRC inspectors discussed the scope of the inspection, outlined the areas to be inspected, and established interfaces with HSB management. In the exit meeting on July 18,1997, the inspectors discussed ineir findings and concerns.

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PARTIAL LIST OF PERSONS CONTACTED Wilfred LaRochelle, Corporate Quality Assurance Manager Barry Bobo, National Manager, Engineering Services Sidney Montgomery, Regional Supervisor, Engineering Services Harold Robinson, National Safety and Training Manager '

Lashanta Lewis, Engineering Support Assistant L

ITEMS OPENED Opened 99900601/97-01-01 NON Failure to perform required audits 99900601/97-01-02 NON Failure to issue nonconformity reports 99900601/97-01 NON Inadequate procedures s 99900601/97-01-04 NON Inadequate control of audits 99900601/97-01-05 NON Inadequate documentation of auditor qualification l

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p f noog/* , UNITED STATES

,a j

2 NUCLEAR RECULATORY COMMISSION WASHINGTON, D.C. 2006M001

'+,. . . . . ,o September 20, 1997 Dama Winies, President NUS Instruments, Inc.

440 W. Broadway Idaho Falls, ID 83402

SUBJECT:

NRC INSPECTION REPORT 99901320/97 01 AND NOTICE OF NONCONFORMANCE

Dear Ms. Wirries:

This letter addresses the inspection of your f.'.ility at Idaho Falls, Idaha, conducted .y Bill Rogers and Robert Pettis, of this of6ce on August 19 21,1997, and the discussions of their 6ndings with John McGimpsey, Cheryl Allen, and other members of your staff at the conclusion of the inspection.

Areas examined during the inspection and our 6ndings are discussed in the enclosed report. This inspection consisted of an examination of procedures and representative records, interviews with personnel, and observations by the inspector During this inspection it was found that the implementation of your Quality Assurance (QA) program failed to meet certain NRC requirements. It was determined that NUS Instruments, Inc.,

(NUS) did not adequately document the corrective actions taken to prevent recurrence of the inadequate soldering identified in modules shipped to Public Service Gas & Electric in the July 1995 to July 1996 time period. In addition, NUS did not take corrective actions for an extended period of time following the identi6 cation ofless than acceptable work quality in the area of Quality Control inspection. The speci6c 6ndings and references to the pertinent requirements are identined in the enclosures of tl..s letter.

Please provide us within 30 days from the date of this letter a written statement in accordance l with the instructions speciSed in the enclosed Notice of Nonconformance. We will consider l I

extending the response time if you can show good cause for us to do so.

l The respenses requested by this letter and the enclosed notice are not subject to the clearance j procedures of the Of6ce of Management and Budget as required by the Paperwork Reduction J Act of l980, Pub. L. No.96-511.

Ms. Wirries In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter, its enclosures, and your response will be placed in the NRC Public Document Room (PDR). To the  !

extent possible, your response should not lic'ude any personal privacy, proprietary, or safeguards information so that it can be placed in the P7R without redaction. However, if ou find it necessary to include such information, you should clearly indicate the specific information  ;

that you desire not to be placed in the PDR, and provide the legal basis to support your request for withholding the information from the public.

Sincerely, Q K~B m c. -

Stuart A.- Richards, Chief Special Inspection Branch Division ofInspection and Support Programs Office cf Nuclev Reactor Regulation Docket No. 99901320

Enclosures:

1. Notice of Nonconformance

2. Inspection Report 99901320/97-01

NOTICE OF NONCONFORMANCE NUS Instmments, Inc. Docket No. 99901320 Idaho Falh, Idaho Based on the results of an inspection conducted on August 19 through 21,1997, it appears that certain of your activities were not conducted in accordance with NRC requirements.

A. Criterion XVI, " Corrective Action," of Appendix B to 10 CFR Part 50, requires, in part, that for significant conditions adverse to quality, measures will be established to assure that the cause of the condition is determined and corrective action taken to preclude repetition. The identification of the significant condition adverse to qualit , the cause of the condition, and the corrective action taken shall be documented and reported to appropriate levels of management.

Section 16 of the NUS Intruments, Inc., (NUS) Quality Assurance manual, " Corrective Action," Fifth Issue, Revision 0, dated September,1994, requires the prompt identification, documentation, and correction of conditions adverse to quality and, in the case of significant conditions adverse to quality, documentation of corrective actions to preclude recurrence.

Contrary to the above, (1) NUS did not adequately document the corrective actions taken in response to the identified occurrences ofinadequate soldering on modules manufactured and provided to Public Service Gas & Electric in the July 1995 to July 1996 time period, and (2) NUS did not take prompt corrective action following the identification ofless than acceptable work quality in the area of Quality Control inspection in July of 1996, for significant conditions adverse to quality.

(Nonconformance 99901320/97-01-0l)

Please provide a written statement or explaation to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555, with a copy to the Chief, Quality Assurance, Vendor Inspection, and Maintenance Branch, Division of Reactor Controls and Human Factors, Office of Nucle.ar 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 Nonconfc mance" and should include for each Nonconformance: (1) the reason for the nonconformance, or if contested, the basis for disputing the nonconformance, (2) the corrective steps that have been taken and the results achieved, (3) the corrective steps that will be taken to avoid further noncompliances, and (4) the date when ycur corrective action will be completed. Where good cause is shown, consideration will be given to extending the response time.

Dated at Rockville, Maryland this A"- day of $4pbt 1997 Enclosure 1 1

1

U.S. NUCLEAR REGULATORY COMMISSION I

i l

- Docket No: 99901320 l

t i

Report No: 97-01 Facility: NI'S Instruments, Inc.

Location: 440 W. Broadway Idaho Falls, Idaho 83402 Dates: August 19 - 21, 1997 i

Inspectors: B. Rogers, Team Leader R. Pettis, Senior Reactor Engineer Approved by: G. Cwalina, Section Chief VendorInspectior Section Special Inspectica Branch Enclosure 2

1 Inspection Summary-

-1,1 Hgkaround and Basis' NUS Instruments, Inc., (NUS) provides engineering and manufacturing capabilities for the replication, refurbishment, and redesign of electronic circuit card assemblies and modules which are provided to NRC licensees as safety-related equipment _ and services.

NUS is owned by Scientech, Inc. (Scientech) which acquired NUS in October of 1996.

- The basis for the NRC inspection of NUS included:

• 10 CFR Part 50, Appendix B e- 10 CFR Part 21

• American National Standards Institute /American Society of Mechanical Engineers Standard N45.2.6-1978, " Qualification ofInspection, Examination, and Testing Personnel for Nuclear Power Plants" as committed to in the NUS Quality Assurance Manual 1.2 Nonconformances Nonconformance 99901320/97-01-01 was identified and is discussed in Sections 3.1 and 3.3 of this report 2 Status of Previous Inspection Findings No previous inspections have been conducted at this facility.

3 Inspection Findings and Other Comments 3.1 - 10 CFR Part 21 Program and Corrective Actions

a. Smpg The inspectors reviewed the NUS 10 CFR Part 21 implementing procedure,10 CFR Part 21 Posting, and the NUS records related to 10 CFR Part 21 evalrations performed by NUS or performed by customers and subsequently provided to NUS. The purpose of the review was to verify that NUS was meeting the requirements of 10 CFR Part 21 in the applicable activities.

2

b, Observations and Findinot On August 8,1996, Public Service Electric & Gas Co. (PSE&G) notified the NRC that numerous NUS Model OCA801 Signal isolator Modules and Model MTH801 Signal

- Summeor Modules, used in the Salem Nui ear Plant, Reactor Control and Protection System (RPS), were found have to unsolden d or insufficiently solder-d internal electrical connections.

I The documentation provided by PSE&G indicated that the isolator module function is to provide _ electrical separation between the Reactor Protection System (Class lE) and the Process Control System (non-Class lE) portions of the RPS instrument loops, and the l_ summator module function is to algebraically combine analog signals. The reactor i protection system is designed to assure that the system can perform its required functions in the event of a design basis earthquake. The defect could affect the modulei ability to

(

l function during a design basis earthquake and could result in a loss of .Ldancv sufficient to constitute a major degradation of essential safety-related equipment.

NUS documentation indicated that two reviews of the PSE&G modules had been performed: (1)" Corrective Action / Root Cause Evaluation," which was completed in 1 September of 1996 by the NUS Quality Assurance Supervisor and (2)" Quality Assurance Review of NUS Instruments," which was completed in October 1996 by a three pe son review team contracted by Scientech (NUS's parent company) .

NUS Review l The NUS review listed the root cause of the inadequate soldering to be an organizational  ;

breakdown caused by inadequate prioritization of work and inadequate job skills, work practice, and decision making. Contributing causes included inadequate interface between organizations, inadequate supervisory program monitoring, inadequate self-verification practices, and differences in the soldering workmanship criteria of NUS and PSE&G. Corrective actions were specified to include (1) removal of the dual usignment of the manufacturing lead person, (2) increasing manufacturing supervision,

- (3) hiring an additional QC Inspector, and (4) providing supervisory skills traming to supervisors. The performance of the NUS review had occurred just prior to the Scientech purchase of NUS (October 1996). Scientech and NUS management indicated in a telephone conversatio'n with the NRC inspection team on August 26,- 1997, subsequent to the completion of the inspection, '. hat they considered the NUS review's root cause investigation to be generally accurate but that the conclusions and corrective actions had not been specific enough to be useful to Scientech management to prevent recurrence of the situation.

In addition, the NUS review indicated that manufacturing operating sheets (which specified the manufacturing and inspection steps) were not being followed and that 3

partial inspections were being performed in accordance with an agreement, between the manufacturing and inspection staff, to modify the operation process sheet Dow. The hianufacturing Supervisor stated that there had been some indication during the period of the PSE&G rnodule productior. (approximately July 1995 to july 1996) that there had been si me rnodincation of activities occur ing at the stafflevel but that reorgniization of the work activities and reassignment of stafT'ind addressed any potential concerns in this area. Other NUS personnel, assigned to NUS management positions during this time period, did not have indication that modification of activities had occurred at the staff level. Scientech management stated that they had reviewed documentation and observed activitiem.d did not agree with thia conclusion of the NUS review. The inspectors noted that modi 0 cation of the operation process sheet now, to alter the specified inspections, would be, as defined by the NUS Quality Assurance hianual (QAht), a Level 11 inyection function and that the QC Inspector performing the work at the time was certined to Level i The inspectors reviewed applicable work records and interviewed numerous personnel and determined that there was documentation that the work had been performed in accordance with the appropriate procedures. Funher, NUS had taken several corrective actions which would affect this area to prevent any potential for reoccurrence on modi 0 cation of activities that might have occuned at the stafflevel.

'cientech Review The Scientech review had some overlap with the conclusions and recommendation of the NUS review but was more detailed and specific. The contributing factors were determined to be related to organizational structure, training, attitude and awareness, fabrication overv!ew, personnel qualification, and work station adequacy. The recommendations made in the Scientech review incbided evaluation of client action requests, evaluation of work station ergonomics, increasing the frequency of trend analysis, improving the coordination of training activities, re6ning the wave soldering process, and improving upervisory skills. Discussion with Scientech and NUS managem:nt indicated that there was agreement with the contents of the Scientech review and that the Scientech review (as opposed to the earlier NUS review) was the basis for the majority of the corrective actions taken by VJS to address the PSE&G soldedng iscie.

Immediate Corrective Actions The inspectors reviewed the corrective actions taken by NUS, following the PSE&G notification, to address the inadequate setdering of the PSE&G hiodules. The NUS hianufacturing Supervisor indicated that when PSE&G notined NUS in July of 1996 of l the inadequate soldering, NUS had reviewed the manufacturing and inspection process and had taken numerous, immediate corrective actions.

4 l

-4 6-

! The Manufacturing Supervisor had reviewed the soldering process and had determined I

that the majority of the manufacturing line (refened to as " assemblers" by NUS) were soldering at the " preferred" level while two assemblers were soldering at the

" acceptable" level (a lesser quality level than " preferred"). While NUS had determined that ti e " acceptable" level was adequate fa functi ning electri,al connectier.s h .'M not l meet the :ustomer's (PSE&G) expectatio - All of th: assemblers were retrain d in 'he soldering technique by being provided physical examples of"prefened" soldering, studying the applicable procedures, and undergoing a supervisory assessment of each person's soldering ability. In addition, the Manufacturing S"nervisor assumed the manufacturing lead person's responsibilities (that perscm a reassigne 8), which allowed the supervisor to directly observe the manufacturing work being performed. Of the two assemblers who were soldering at the " acceptable" level, one person was adequately 4 retrained and the other was removed fiom the manufacturing activities and within one l month of the PsE&O notification the entire manufacturing line was soldenng at the )

" preferred" levei. Th inspectors noted that the requirement to solder at the " preferred" level of quality had not been procedurealized by NUS.

During the period of July 1995 to July 1996, all of the NUS assemblers had performed l work on the PSE&G modules. The two assemblers previously identified as soldering at the " acceptable" level had not performed any work on projects other that the PSE&G modules manufactured during that period. Work for customers other than PSE&G had been performed by two other assemblers, both of whom the Manufacturing Supervisor had considered excellent performers. NUS reviewed the work performed in the July 1995 to July 1996 time period, and had concluded that there was not a concern with work performed for customers other than PSE&G.

The inspectors noted that NUS had not documented the inadequete soldering discovered in the PSE&G modules in accordance with the requirements of the NUS QA program and had nut documented the immediate corrective actions previously discussed.

Additional Corrective Act:- u In addition to the immediate corrective actions taker. by NUS to raise the level of soldering quality, NUS had taken additional corrective actions, based on the Scientech review, which included rc:rganizing the manufacturing line, installing a mechanical wave soldering machine, raining the manufacturing supervitory and lead personnel, and upgrading the QC Inspector position to Level II. NUS also indicated that ongoing

)

training of assembly and inspection personnel would occur, peer reviews had been instituted for manufacturing personnel, and deficiency logs had been established to track soldering rework, although these conective actions had not yet been proceduralized at the time of the inspection.

5

c. Conclusions The inspectors noted the indication that modincation of the operation process sheets, to alter specined inspections, had occured in the July 1995 to July 1996 puiod, that such a rx dincation would have been a R m h mspection function, and that the QC Inspector performing the work at the time was i ei ined to Level 1. Indication of this modi 0 cation highlighted a potential weakness in the NUS Quality Assurance program.

The inspectors concluded that NUS had taken reasonable corrective actions in response i to the identincation ofinadequate soldering in the Salem modules. However, NUS had not adequately documented the corrective at tions which NUS indicated were taken upon identi0 cation of the inadequate soldering. This was identined as an example of Nonconformance 99901320/97 0101, 3.2 Manufacturing Procedure and Implementation of Soldering Technione

a. Scope The inspectors reviewed the applicable procedures, discussed activities with manufacturing and inspection personnel, and observed ongoing inspection and manufacturing activities, to assess whether soldering activities were being adequately controlled.

b. Observations stid Findings The inspectors reviewed the training records foi several asserublos to determine their quali0 cations to perform soldering. Training was documented on a "Certi0 cation of Training" which indicated that assembler was qualined to solder in accordance with the "Beckwith Training Course E- 5" and in accordance with the NUS Operating Procedures Manual, Appendix E, " Soldering," Revision 0, dated March 12,1992. The certincate documented that the assemb'er had received the required instmetion, had passed a written examination, and com, 'eted a practical demonstration of soldering ability.

The inspectors observP "~mblers installing components, wrapping wire, and performing hand soldering ihe inspectors also observed a demonstration by the QC Inspector on performance of an inspection of solder connections and discussed the levels of classincation of soldering quality. The level of solder quality required for NUS product is the " preferred" level which is the premium level of solder quality, NUS indicated that the "preferTed" level of solder quality was mandated and strictly adhered to as a corrective action to the PSE&G Part 21 report (See section 3.1.2).

6

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c. Conclusions The inspecars co:1dujed that NUS had in place a program to train manufacturing personneli solder at the required level of soldering quality and to verify the nenufacturing personnel's ability at :l hat this program was being adequately implemented. In addition, the QC Inspector possessed the ability to verify the adequacy of the soldered connections.

3.3 Review of Qualincations for QC Inspectors

a. Scope The inspectors reviewed QC Inspector quali0 cation records and se':cted documentation to determine corapliance with American National Standards Institute /American Society of Mechanical Engineers (ANSI /ASME) N45.2.6-1978,"Qualincah .: afinspection, Examinamon, and Testing Personnel for Nuclear Power Plants" as committed R in the NUS QAM. The sequirements included in the ANSI Standard are for the quali0 cation of personnel who perform inspection, examination, and testing of nuclear power plant items used in safety related applications.

l b. Observations and Findings NUS required that the capabilities of a candidate for certiGcation as a Level 1, II, or 111 QC Inspector be initially determined by a suitable evaluation of the candidate's j

education, experience, training, test results, or capability demonstration. Once cenined, l

the QC Inspector's job performance was reevaluated at periodic intervals not to exceed  !

three years. l The NRC inspectors reviewed the training files of all QC Inspectors to verify that the above requirements were met. A total of seven QC Inspector Gles were reviewed and all appeared to be in compliance with the requirements for certin:ation. One QC Inspector was certi6ed to Level 1, three QC Inspectors were certified to Level 11, and three were certined to Level 111. liowever, as of the inspection, only one Level 11 QC Inspector remained active since the others had either been reassiF ned to other positions or were no longer employed at NUS. The present NUS QA Manager, certined to Level 111, supervised the work of the Level 11 QC Inspector A review of the Oles indicated that the QC Inspectors were certined primarily on the basis of education and related experience and that the qualineation records were in compliance with ANSI Standard N45.2 6.

The inspectors reviewed the work being performed by the QC Inspector to determine whether the work met the dennition of Level I work as denned by ANSI standard N45.2  !

6 which states, in part, that Level 1 persons shall be capable of performing inspections, examinations, and tests in accordance with documented procedures. Level 11 persons 7

shall have, in addition to the Level I capabilities, demonstrated capabilities in planning, setting up, and supervising inspections and tests. The inspectors reviewed documentation for several POs which indicated that the QC Inspector had performed activities consistent with the Level I definition s# ; verifying component placement, ver'Scation of soldering, and final visual in pections. Ilowever, the inspectors cautioned NUS management that allowing manufactur ig end inspection starf to modify the operation process sheet Dow without management approval, as discussed in detail in Section 3.1.2, would be a Level 11 inspection function and inappropriate if the QC Inspector performing the work at the time was certified to Level I or irsuch operation process modifications were disallowed by the NUS QA program.

During review of the training files to verify QC Inspector qualifications, two doaments were id ntified which indicated a quality control inspection activity weakn.ss Two memoranda from the QA Supervisor identified weaknesses in the QC Instector's performance and one recommended management action. These memoranda, ii..tiaied in January and July of 1996, anc' were currently in the QC Inspector's training Ole.

Discussion with the current NUS ud Scientech management indicated agreement that the QC Inspector's performance was not adequate and that the QC Inspector had been subsequently reassigned shortly after the current management was made aware of the QC Inspector's performance in December of 1996. liowever, during the period cf time from July 1996 until December 1996, NUS had been made aware of the inadequcte performance of the QC Inspector and had not taken any corrective action. Subsequent to the new management being placed, NUS had reassigned the QC Inspector, and had employed two additional persons in that position. The inspectors reviewed the training files of the two most recent QC Inspectors, discussedt heir performance with the QA Supervisor and the Manufacturing Supervisor, reviewed work documentation, and observed work performar,ce of the current QC Inspector. The inspectors did not observe any indication ofinadequate performance in the documentation or work obs rvations of the current QC Inspector.

C. (QDClusion Although NUS had been made aware of the potentially inadequate work performance of a QC Inspector by the inspector's direct supervisor in January and July of 1996, and that this information was available in the QC Intpector's training file, NUS had not taken any corrective action for this situation adverse to quality until December of 1996. The failure to take corTective action to correct a significant condition adverse to quality was identified as an example of nonconformance 99901329/97-01-01.

8 50-

4- Personnel Contacted Dama Wirries, President, NUS John McGimpsey, General Manager, NUS

- Cheryl Allen, QA Supervisor,- NUS Shauna lloyack, Manufacturing Supervisnr, NUS Heath Buckland, Testing Supervisor, NUS Ron Todd, Quality ControlI.ead, NL3

- Paul Sturm, Principal Engineer, Scientech, Inc.

.)

i

@ **o

,y 4 UNITED STATES

!! NUCLEAR RECULATCRY COMMISSION WASHINGTON O.C. 30e06-0001 g....+

September 9, 1997 I Mr. Gregory M. Rueger Pacific Gas & Electric Com;,any NPG Mail Code A10D PO Box 770000 San Francisco, CA 94177

SUBJECT:

NRC INSPECTION OF THE DIABLO CANYON POWER PLANT (REPORT NOS. 50-275/97201 AND 50 323/97201)

During the period June 9 12. 1997, the Special Ins)ection Branch of the U.S.

Nuclear Regulatery Comission's (NRC's) Office of luclear Reactor Regulation (NRR) performed an inspection of Pacific Gas & Elect"ic Company's (PG&E's) activities related to the procurement, modification, testing and installation of replacement 4 kV circuit breakers at the Diablo Canyon Power Plant (DCPP),

Units 1 and 2.

The primary >urpose of the inspection was to determine if design verification testing of t e modified breaker was accompiished in accordance with applicable requirements of American National Standards Institute (ANSI)/ Institute of Electrical and Electronic Engineers (IEEE) Standards. A seco-!ary purpose of the inspection was to examine several related issues invo1M ng production testing, modifications, and post installation testing in-s ice failures.

The results of this inspection are contained in the enclos'4 nspection report, With respect to design verification, the inspectors determined that PG&E's approach was generally consistent with the applicable NRC regulations in that PG&E undertook to verify, by engineering analysis and a testing program, that DCPP's safety-related 4 kV electrical distribution system as converted and modified would perform its safety functions under all design basis conditions.

By NRC letter dated July 24, 1997 (Attachment 1 to the enclosed inspection report), the question of consistency of the approach with applicable industry standards, in particular, taking credit for certain design verification tests done by the manufacturer of the circuit breakers used in the conversion, was referred to the Standards Board of the IEEE for consideration by the appropriate subcommittee. The details of the IEEE response, contained in an August 21, 1997, letter from the IEEE Power Engineering Society Switchgear Committee (^ttachment 2 to the enclosed report), are discussed in the report, but in summary, the PG&E a>proach is considered consistent with the intent of the standards. Although t1ere is no regulatory guide endorsing the standards in question. PG&E has cmitted to them in the Final Safety Analysis Report (FSAR) for DCPP, and they are therefore relevant to the design and licensing basis of the plant. With regard to the engineering evaluations necessary to support the PG&E approach. the inspectors could not, on the basis of documentation available for review at DCPP, conclusively determine the adequacy of all the justifications for not reperforming certain design tests.

Mr. Gregory H. Rueger Therefore, the inspectors identified the need for further review at National Technical Systems-(NTS). Inc., Acton. Massachusetts: and possibly also Power Distribution Services-(PDS). Inc.. Cincinrati Ohio, in accordance with 10 CFR 2.790 of-the Comission's regulations, a copy of this-letter and inspection report will be placed in the NRC Public Document Room. Should you have any questions concerning the attached inspection report. please contact the inspection team leader. Mr. Stephen Alexander, at (301) 415 2995.

Sincerely, j

i Stuart A. Richards. Chief Special Inspection Branch Division of Inspection and Support Programs Office of Nuclear Reactor Regulation cc: See next page Docket Nos. 50 275 and 50 323 License Nos.: DPR-80 and DPR 82

Enclosure:

Inspection Report No. 50-275.323/97201 Mr. Gregory M. Rueger -3 CC:

NRC Resident Inspector Regional Administrator. Region IV Diablo Canyor Nuclear-Power Plant U.S. Iluclear Regulatory Commission c/o U.S. Nuclear Regulatory Commission Harris Tower & Pavilion P.O. Box 369 611 Ryan Plaza Drive. Suite 400 .

Avila Beach, California 93424 Arlington Texas 76011 8064 t i

Dr. Richard Ferguson Energy Chair Christopher J. Warner Esq.

Sierra Club California Pacific Gas & Electric Company  :

1100 lith Street. Suite 311 Post Office Box 7442 Sacramento. California 95814 San Francisco. California 94120  ;

Hs. Nancy Culver Mr. Robert P. Powers San Luis Obispc Vice President and Plant Manager Mothers for Peace Diablo Canyon Nuclear Power Plant P.O. Box 164 P.O. Box 56 Pismo Beach. California 9344B Avila Beach. California . 93424 Chairman Telegram-Tribune San Luis Obispo County Board of ATTN: Managing Editor ,

Supervisors 1321 Johnson Avenue Room 370 P.O. Box 112 San Luis Obispo. CA 93406-County Government Center San Luis Obispo. California 93408 1

Mr. Truman Burns Mr. Robert Kinosian' California Public Utilities Commission 505 Van Ness. Room 4102 San Francisco California 94102 ,

Mr. Steve Hsu Radiologic Health Branch-State Department of Health Services .

Post Office Box 942732 -

Sacramento. California 94232 Diablo Canyon Independent Safety Committee ATTN: Robert R. Wellington. Esq.

Legal Counsel 857 Cass Street. Suite D Monterey. California 93940 r

Mr. Gregory M..Rueger cc w/ encl:

NRC Resident Inspector Regional Administrator, Region IV Diablo Canyor Nuclear Power Plant U.S. f'uclear Regulatory Commission c/o U.S. Nuclear Regulatory Comission Harris Tower & Pavilion P.O. Box 369 611 Ryan Plaza Drive, Suite 400 Avila Beach, California 93424 Arlington, Texas 76011-8064-Dr. Richard Ferguson, Energy Chair Christopher J. Warner Esq.

Sierra Club California Pacific Gas & Electric Company 1100 lith Street, Suite 311 Post Office Box 7442 Sacramento, California 95814 San Francisco, California 94120 .

l Ms. Nancy Culver Mr. Robert P. Powers San Luis Obispo Vice President and Plar.: Man.ger ,

Mothers for Peace 016blo Canyon Nuclear Power Plant  :

P.O.-Box 164 P.O. Box 56 Pismo Beach, California 93448 Avila Beach, California 93424 Chairman Chief. Fuel Cycle and Decommissioning San Luis Obispo County Board of Branch, Region IV Supervisors U.S. Nuclear Regulatory Commission Room 370 611 Ryan Plaza Drive Suite 400 County Government Center Arlington, Texas 76011-8064 San Luis Obispo, Ca'.ifornia 93408 Mr. Peter H. Kaufman

) Mr. T'uman Burns Deputy Attorney General Mr. Robert Kinosian State of California

' California Public Utilities Comission 110 West A Street, Suite 700 505 Var, Ness, Room 4102 San Diego, California 92101 San Francisco, California 94102 Mr. Thomas A. Moulia

-Mr. Steve Hsu. Humboldt Ba Plant Manager Radiologic Health Branch Humboldt Ba Nuclear Power Plant State Department of Health Services 1000 King S 1 mon Avenue Post Office Box 942732 Eureka, California 95503 Sacramento, California 94232 Chairman, Humboldt County Board Diablo Canyon Independent Safety of Supervisors Committee County Courthouse ATTN: Robert R. Wellington, Esq. 825 Fifth Street Legal Counsel Eureka, California 95501 857 Cass Street. Suite 0 Monterey, California -93940 Redwood Alliance P.O. Box 293 Arcata, California 95521

s U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION Docket Nos.: 50-275, 50 323 License Nos.: DPR 80, OPR 82 Report No.: 50-275/97201, 50 323/97201 Licensee: Pacific Gas & Electric Company Facility: Diablo Canyon Power Plant, Ur .51 and 2 Location: Avila Beach, California Dates: June 9 12. 1997 Inspectors: Stephen D. Alexander Team Leader. NRR Billy H. Rogers Reactor Engineer, NRR Approved by: Gregory C. Cwalina. Section Chief Vendor Inspection Section Special Inspection Branch Division of Inspection and Support Programs Office of Nuclear Reactor Regulation Enclosure 1

EXECUTIVE

SUMMARY

During the period of June 9-12. 1997. representatives of the U.S. Nuclear .

Regulatory Commission's (NRC's) Special Inspection Branch conducted an inspectton of Pacific Gas 8. Electric Company's (PG&E's) activities related to- -

the procurement, modification, testing and installation of replacement 4 kV '

circuit breakers at the Diablo Canyon Power Plant (DCPP). Units 1 and 2.

- The inspectors reviewed engineering and quality assurance docu M ation, '

interviewed cognizant staff and examined equipment in order to waluate PG&E's dedication and modification of 4 Kv Yaskawa circuit breakers for use in safety 4

related applications at DCPP. The primary purpose of the inspect'.on was to i determine if prototype design verification testing of the modified breaker was accomplished in accordance with applicable NRC regulations and industry

' standards. Although there is no regulatory guide endorsing the standards ir ,

question. PG&E had committed to them in the Final Safety Analysis Report (FSAR) for DCPP. and they are therefore relevant to the design and licensing

• basis of the plant.

A secondary purpose of the inspection was to exwine several related issues involving production testing, modifications, post installation testing, and in service failures. The review iricluded all documentation pertinent to the design, design verification (including prototype testing), conversion / modi-fication, fabrication of adapting hardware, production testing rocurement and dedication done by PG&E. its principal contractor Nationai echnical Systems (NTS). Inc., of Acton Massachusetts, and the NTS subcontractor who actually )erformed the conversions, Power Distribution Services (PDS). Inc.,

cf West C1 ester. Ohio.

The inspectors determined that PG&E's approach to design verification was generally consistent with the a)plicable NRC regulations, primarily Criterion 111. " Design. Control." of 10 CFR Por 50. Appendix B. in that PG&E undertook to verify. by engineering analysis and a testing program, that DCPP's safety-related 4 kV electrical distribution system as converted and modified would perform its safety functions under all design basis conditions.

The question of consistency of PG&E's ap3 roach to the' conversion and its design verification process with apfilca)le industry standards, in particular, taking credit for certain design verification tests done by the manufacturer of-the circuit breakers used in the conversion, was referred to the Standards Board of the IEEE for consideration by the appropriate subcommittee. In a letter to the NRC. dated August 21, 1997, the Chairman of the High-Voltage Circuit Breaker Subcommittee of the Switchgear Committee of the IEEE Power Engineering Society (sponsor of the principal applicable standard. C37.59) confirmed the licensee's understanding of the intent of the standard regarding required design verificatior, testing for conversions using modular assemblies.

The letter further stated that if the modular assembly is in no way altered with respect to the coupling of the interrupting chambers and the operating mechanism. then some of the original design tests performed by the manufacturer, such as short circuit current interruption, load switching and capacitance switching tests, need not be repeated. However, the letter i 2

stated, to apply this waiver, it must be shown that the mechanical operating characteristics of the interrupting chamber, such as contact parting times and contact travel, are still within the range specified for the original module prior to the conversion.

With regard to the engineering evaluations necessary to support the PG&E approach, the inspectors could not, on the basis of documentation for review at DCPP. conciusively determine the adequacy of all the justifications for not reperforming certain design tests. Therefore this issue remains unresolved and the inspectors identified the need for further review at NTS and possibly also at PDS. Accordingly, this issue is designated Unresolved item 50 275.323/97210 01.

3

Reoort Details III. Enaineerina E1 - Conduct of Engineering E.1 4 kV Swi'choear Conversion Desian Vetification

a. Insoection Scooe in order to verify that PG&E and its contractors. NTS and PDS had properly verified the interrupting capacity of the 350 HVA Yaskawa SF6 interrupter breakers, adapted to fit into 250-HVA GE Magne Blast switchgear, the inspectors reviewed engineering and quality assurance documentation.

interviewed cognizant staff, examined equipment, and evaluated PG&E's dedication and modification of 4 Kv Yaskawa circuit breakers for use in safety related applications at DCPP.

The primary issues examined were:

e Whether the design verification testing conducted on the complete conversion (consisting of the Yaskawa " modular assembly" plus the j hardware to adapt it to the Magne Blast cubicle) for PG&E (conducted for i NTS at PSM Inc.. of Pittsburgh, Pennsylvania) was consistent with ANSI /IEEE Std C37.59-1991. Paragraph 5.1.4.2(2). for conversions using adapted modular assemblies, e Whether taking credit for the Yaskawa ANSI testing of t'e modular assembly supplemented by technical evaluations was consistent with Section 5 of C37.59, e

Whether the Yaskawa technical evaluations of its modifications made to the modular assembly after the original design verification tests were adequate to demonstrate that the original test results were still valid after those factory ::d ifications, e Whether the technical evaluations by NTS to demonstrate that its additional testing of the complete conversion as required by C37.59 was not invalidated by mouifications made in response to installation, setup and operational problems made subsequent to the tests, and l

e Whether PG&E adequately resolved the findings identified in its audits of NTS. PDS and Yaskawa.

Also examined were:

e The material of the breaker secondary disconnect pins and e :The potential overtravel situation involving the stationary auxiliary switch (SAS) in the cubicle possibly preventing full closure of the breaker, and several other interface and operation issues brought to the VIS inspectors' attention by personnel from DCPP Operations.

4 i

The inspectors reviewed the PG&E documentation related to the design tests of  :'

the-SGYB 1-1200 AND SGYB 1 2000 circuit breakers performed by Yaskawa PDS.

and PSM: the evaluations performed by NTS which demonstrated the equivalenc

• of the circuit breakers supplied to PG&E to the prototype circuit breakers y tested by Yaskawa and the two circuit breakers tested by PDS and PSM: and the review of the production modifications made by Yaskawa and PDS following the ANSI /IEEE testing. The inspector'c r;".: / was performed to verify that the circuit breakers su) plied to PG&E were equivalent to the circuit breakers tested by Yaskawa >DS. and PSM and that any modifications made to the circuit breakers by Yaskawa or PDS did not invalidate the original design testing,

b. Observations b.1 Validity of Design Verification Approach During the initial stages of the original circuit breaker production. Yaskawa  :

had performed the ANSI /IEEE design testing on prototype breakers as represen-tatives of the circuit breaker which Yaskawa would subsequently sell as  ;

commercial grade products (non safety related). This circuit breaker was the '

Yaskawa SF 6 gas "Fluopac" Series, medium voltage (4.76-kV rated), rotary arc circuit breaker of 350 MVA interrupting capacity. PG&E contracted with NTS to <

provide the Yaskawa circuit treakers as Class 1E (safety related) equipment.

NTS purchased breakers from Yaskawa and subsecuently subcontracted their modification and additional testing to PDS anc PSM. Inc. (a high-energy test facility in Pittsburgh), to be performed under the 0A coverage of NTS.

The particular models of these breakers that underwent design verification testing were Yaskawa Models SGYB-1 1200 and 5GYB 1-2000NTS. The conversions performed by PDS used the interrupting chambers and their attached operating mechanisms and chassis (frame) from the circuit breakers. According to the applicable industry standard. ANSI /IEEE Standard C37.59 1991, these components constitute a " modular assembly." PDS adapted the modular assemblies for retrofit into the existing GE Magne Blast cubicles at DCPP by mounting each assembly in a custom fabricated enclosure and truck unit containing the necessary hardware with which to make the primary and secondary electrical connections in the cubicles to the 4 kV busses and 125 Vdc control power respectively, as well as the mechanical interfaces with the cubicle vertical lift a>

and cu)paratus, stationary auxiliary switch, truck-operated cubicle switch, 1cle interlock devices.

The applicable NRC regulations were Appendix A. " General Design Criteria for Nuclear Power Plants." to Part 50 of Title 10 of the Code of Federal ,

Regulations (10 CFR Part 50. Ap>endix A): 10 CFR Part 50. Appendix B. " Quality Assurance Criteria for Nuclear )ower Plants and Fuel Reprocessing Plants'"

10 CFR 50.59 " Changes. Tests. and Ex>eriments " 10 CFR Part 21. " Reporting of Defects and Noncom)liance." The 10 C:R Part 50. Appendix B. criteria that are l

especially applica)le to this project are Criterion 111. " Design Control." and l Criterion Vll. " Control of Purchased Material. Equipment and Services."

l The principal t A stry standards ap)11 cable to switchgear conversions are '

ANSI /IEEE Stane.m C37.59-1991. "lEEE Standard Requirements for Conversion of Power Switchgear Equipment:" C37.09 1979. "IEEE Standard Test Procedure for AC 5

i 60-

High Voltage Circuit Breakers Rated on a Symmetrical Current Basis " and C37.04 1979, "lEEE Standard Rating Structure for AC High Voltage Circuit Breakers Rated on a Symmetrical Current Basis." Although there is no regulatory guide endorsing the standards in question, the inspectors determined that PG&E had committed to them in the Final Safety Analysis Report (FSAR) for DCPP, and they are therefore relevant to the design and-licensing basis of the plant.

Paragraph 5.1.4.2(2) of C37.59, which deals with conversions using modular assemblies, requires that the modular assembly-undergo the entire series of design tests-in accordance with C37.09.- PG&E and NTS interpreted this to mean that it must be verified that the complete series of design tests has been performed on a prototype (s) of the modular assembly,, by e.g.per the C37.09 manufacturer as part of original testing, with satisfactory results. However, on the basis of the general guidance in the beginning of Section 5 of C37.59.

PC&E further interpreted the standard to provide that if engineering evalua-

! tions of all :ubsequent modifications can demonstrate that function, or

) characteristics (e.g., interrupting capacity) of the converted breaker are not f

l adversely affected by the modifications such that the original tests would be invalidated, then the C37.09 design tests of those functions or characteris-tics need not be repeated on the complete conversion.

According to the documentation provided to PG&E by NTS (and reviewed by the

-ins)ectors), the original circuit breaker design tests had been performed by Yascawa in accordance with ANSI /IEEE C37.09 1979. C37.20.2-1987, C37.59 1-1991 and PG&E Specification 1001-E NPG Section 12.1. The purpose of the tests was-to determine the adequacy of the design of this particular type and model and 4

d its components parts to meet its assigned ratings and operate satisfactorily  ;

under normal service conditions /special conditions defined by the PG&E specification. NTS Report No.60431 95N C, " Equivalency Evaluation'of ANSI Type Tests and ANSI C37 Test Re> orts," Revision 7. dated June 4, 1997, documented the required ANSI /IEEE tests aerformed, specifying a descri> tion of each test, which company had performed tie test, and including applica)le test report.

-Paragraph 5.1.4.2(2) of C37.59 then recuires specific additional testing of the complete conversion, which includec dielectric, momentary (C37.20.31987),

continuous current, interlock and u.ailiary functions, and mechanical endurance testing (C37.06 1987). The inspectors confirmed by review of the test documentation that these tests were performed on two representative complete conversions with satisfactory results.

In reviewing the design verification of the breaker conversion, the inspectors learned that PG&E had consulted with several industry ex>erts, including, most notably. Dr. Ward Laubach of the low-Voltage Switchgear )evice Subcommittee of the IEEE Switchgear Committee of the_lEEE Power Engineering Society, sponsor of ANSI /IEEE C37.59-1991, who asserted that PG&E's approach was consistent with the provisions of the standard. However, the inspectors noted that Dr.

Laubach was also employed as a consultant to NTS, PG&E's primary contractor on-the project. In addition, the inspectors noted that two other parties

_ consulted by PG&E were-(l) one of the other bidders on the project, an employee of Pacific Breaker Systems. Inc. (who had used the same approach in a 6

-61

, >roject involving French Merlin Gerin breakers for the Quad Cities and Dresden luclear Stations), and (2) an employee of Square D Company, the U.S.

re)resentative for Merlin Gerin (both of which companies are owned by the Scineider Electric conglomerate). The inspectors determined that the industry experts consulted by PG&E were not totally disinterested parties because they were directly or indirectly involved in this or other similar projects.

Therefore, by NRC letter dated July 24. 1997 (Attachment 1 to this report),

the question of consistency of the approach with applicable industry standards, in particular, taking credit for certain design verification tests done by the manufacturer of the circuit breakers used in the conversion, was referred to the Standards Board of the IEEE for consideration by the appropriate subcommittee. In a letter to the NRC, dated August 21. 1997 (Attachment 2 to this report), the Chairman of the High Voltage Circuit Breaker Subcomittee of the Switchgear Committee of the IEEE Power Engineering Society (sponsor of the principal applicable standard. C37.59) confirmed the licensee's and the inspectors' understanding of the intent of the standara regarding requir'd design verification testing for conversions using modular assemblies.

The letter stated that while it was intended that all design tests be performed on the converted equipment in the case of a conversion using a modular assembly, in addition to the specific tests explicitly required to be performed on the complete conversion by Paragraph 5.1.4.2(2) of C37.59-1991.

only those design tests that cover an area of performance affected by the modifications associated with the conversion must be repeated on the complete conversion. The letter further stated that if the modular assembly is in no way altered with respect to the cou) ling of the interrupting chambers and the operating mechanism, then some of tie original design tests performed by the manufacturer, such as short circuit current interruption, load switching and capacitance switching tests need not be repeated. However, the letter stated, to apply this waiver. it must be shown that the mechanical operating characteristics of the interrupting chamber, such as contact parting times and contact travel, are still within the range specified for the original module prior to the conversion. Although there is no regulatory guide endorsing the standards in question. PG&E had committed to them in the FSAR for DCPP. and they are therefore relevant to the design and licensing basis of the plant.

b.2 Equivalency Evaluation The inspectors found that the original Yaskawa breaker required extensive M substantial modification in order to successfully adapt it for use in the Magne Blast switchgear; modification for which neither Yaskawa. nor PG&E, nor its subcontractors had originally or promptly provided comprehensive engineering evaluation (s) (at least in english) to est3blish that the modifications would not adversely impact the tested interrupting capacity, and would not invalidate the original design tests performed by Yaskawa.

According to the certifications provided to PG&E, NTS had established that several areas were critical to determining that the circuit breakers supplied to PG&E were equivalent to the circuit breakers tested by Yaskawa. PDS and PSM. The documentation further certified that NTS had maintained the design 7

control of materials dimensions, and processes, verified that all ANSI /IEEE testing was performed by approved vendors (or surveilled) and performed in a calibrated test facility, and materials on the supplied circuit breakers were  !

appropriately dedicated. In addition. NTS reviewed all design changes made by Yaskawa and PDS and .rtified that the modification did not invalidate the ANSI /IEEE telting performed by Yaskawa and PDS (and for NTS/PDS at PSM).

Yaskawa had provided NTS information on all design, parts, or material changes made to the circuit breaker since May of 1993, when the ty)e tests had been performed, to the time of the NTS purchase. The_Yaskawa c1anges were contained in Yaskawa document no. GA9400864 Statement of Design Change.

Revision 2. dated July 31, 1995, in section 5.0 of NTS Report No. 60431 95N C.

Yaskawa had made numerous changes to the operating mechanism the interrupter and the general assembly including items such as material changes, dimensional changes. and drawing changes. Yaskawa indicated in the document that none of the indicated changes would impact the results of the ANSI /IEEE type tests which had been performed by Yaskawa in 1993, and provided  !

certification to that effect with the circuit breakers shipped to niS. In addition. NTS had reviewed the changes and performed and evaluation of those which NTS considered had the potential to affect critical characteristics of the as-tested design and had concluded that the Yaskawa changes had not affected the results of the type tests originally performed by Yaskawa.

In addition to Yaskawa's modifications to the modular assemblies supplied to NTS. PDS. under NTS controls, had also made modifications to the circuit breakers to facilitate the modular conversion and allow them to operate in the installed DCPP Magne Blast switchgear. The PDS engineering change notice (ECN) table for Job #1466. included in NTS Report No. 60431-95N C. contained all of the changes that NTS considered relevant to the conversion design. The PDS modifications were primarily mechanical changes to the circuit breaker frame, wheels, and hardware. Each ECN was accompanied by an evaluation for impact on the circuit breaker ANSI /IEEE testing which had been performed by Yaskawa. PDS or PSM. NTS Report No. 60431-95N C concluded that none of the PDS modifications invalidated the circuit breaker testing performed by Yaskawa. PDS or PSM.

b.3 Low Voltage Fault Current Test The inspectors found that although PG&E did not conduct a rated ' voltage.

rated fault current interrupting capacity test on the complete conversion. it '

ordered a special fault current test at PSM (in addition to the additional testing on the complete conversion required by C37,59), which was conducted at 480 volts instead of 4760 volts. PG&E explained that-there might be adverse effects of fault-current magnetic fields on the components added to the i complete conversion to adapt it to the auxiliary switches and mechar.ical i interlocks of the Magne Blast cubicle. These effects might cause che added i hardware to impede breaker tripping on a fault: effects that maf not have been covered by the testing required by the standards. Therefore. PG&E decided to conduct an interrupting capacity test of a prototype of the ,omplete conversion unit that was already undergoing the testing required by Paragraph 5.1.4.2(2) of ANSI /IEEE Std C37.59 in July 1996 at PSM. However, due to some problem with or unavailability of PSM's main high current test facility 4

8 generator, the test was being done at 480 Vac (although presumably at the required current level) instead of the 4760 Vac reportedly required for 4 kV breakers by C37.09 (Referenced in C37.59). PG&E argued, that even though the test was at low voltage, the fault current would produce magnetic fields to adequately simulate the fault interrupting conditions that might conceivably affect the operation components in question irrespective of the voltage at which the 41.000 amp test was cont:t e , as only the current, not the voltage gives rise to the magnetic fields.

c. Conclus120 l The inspectors concluded that PG&E's approach to design verification was 1 generally consistent with the a)plicable NRC regulations, primarily Criterion Ill. " Design Control." of 10 CF1 Par 50. Appendix B. in that PG&E undertook to verify relatedby4-kV engineering electricalanalysis andsystem distribution a testing as program convertedthat andDCPP's safety modified wou ld perform its safety functions under all design basis conditions. On the basis  !

of the IEEE interpretation of the intent of C37.59. the inspectors further concluded that the PG&E approach was consistent with applicable industry standards, provided the required supporting engineering evaluations were adequate. On the basis of the review of the supporting documentation supplied by Yaskawa, NTS, and PG&E. the inspectors further concluded that the applicable ANSI /IEEE tests had been performed with satisfactory results.

However, although PG&E had certifications trom NTS that equivalency had been established between the tested circuit breakers and those supplied to PG&E.

the inspectors could not conclude on the basis of documentation available for review at DCPP that the technical evaluations aerformed 3rimarily by NTS demonstrated that the modification made by Yastawa and P]S had not impacted the validity of the ANSI /IEEE tests performed on the circuit breakers; therefore, the inspectors could not conclusively determine the adequacy of all the justifications for not reperforming certain design tests. Accordingly, the inspectors identified the need for further review at NTS, and possibly also at PDS and designated this issue as Unresolved item 50 275,323/ 97201-01.

E.2 4 kV Production Breaker Installation and Performance Concerns

a. Insoection Scope In order to address three areas of concern that had been identified by the inspectors in preparation for this inspection: (1) 4-kV breaker secondary disconnect c: itact pin material. (2) breaker operation interference due to stationary auxiliary switch (SAS) overtravel, and (3) SAS adjustment /

performance, the inspectors reviewed the associated DCPP Action Requests (ARs) and their dispositions, interviewed cognizant engineering and operations

personnel, and examined affected components, l

b. Observations l

l L

b.1 Secondary Disconnect Pin Material in the case of the secondary contact pin material, the pins on the breakers for DCPP Unit 1, which had not yet been shipped were replaced with pins of a stiffer: more tempered material. The pins on the breakers in Unit 2 were not replaced en masse, but rather inspected and replaced if permanently deformed or otherwise damaged or degraded. In additien. PG&E had discovered that the, reason the pins (of more malleable material than those of the original GE secondary contact blocks) were becoming deformed, in some cases enough to degrade electrical contact. was the manner in which maintenance electricians had become used to removing the secondary contact test position adapter cable and plug assembly, i.e., by yanking it off partially sideways. Accordingly.

PG&E changed procedures and conducted training to ensure that the test cable plugs would be pulled off carefully and only with ver tical force to prevent any future pin deformation. In addition, procedures were changed to require the use of a GE secondary contact pin spreading or gapping tool after each test cable reinoval to ensure that the four segments of each pin were )roperly spread for adequate electrical contact when the breaker was fully racred up into its operate position. The representative from DCPP o>erations who had also related this concern to the team was satisfied that tie corrective action was adecuate. PG&E stated that there were sufficient replacement pins of the improvec, stiffer /more tempered material on site to replace any pins that should become irreparably deformed despite improved handling procedures.

b.2 SAS Overtravel With respect to the concern about the potential for an overtravel condition in the SASS potentially preventing the converted Yaskawa breakers from closing fully (which never occurred in service, only during experimenta. ion), the inspectors determined that PG&E's minimum recuired gap (0.040") between the stationary auxiliary switch operating rod anc the breaker's mechanism operated cubicle plunger (set by adjusting shims in the plunger and thereafter by manually adjusting open breaker elevation in its cubicle), in conjunction with the )rocedures that required checking and establishing this gap each time a breater was racked in, would prevent the overtravel condition from occurring.

b.3 SAS Adjustment and Performance in addition, the DCPP operations re)resentative had related concerns to the inspectors regarding the several otler interface problems that had been encountered relating to the SASS. Having suffered several equipment failures since the installation of the converted Yaskawa breakers attributable to

)roblems with the SASS (all sets of contacts not always changing state with areaker operation). PG&E had determined (through testing) the worst case stroke requirement (they are somewhat variable) among all the SASS (i.e..

stroke of the operating rod required to ensure that all contacts in the SAS a GE SB-12 switch, will fully change state). Some older switches that had actually caused failures or were found through testing to be unreliable or out of tolerance were replaced. PG W then determined the maximum allowable breaker- open, plunger-SAS op. , ting rod gap that would ensure that all SAS contacts would cht.nge stNe wtan the breaker closed given the worst case (largest) required stroke of the all the SB 12 switches, 10 During this period, another related problem presented itself. Upon investigating the failure of a pum) to start PG&E found that another instance of 4 kV breaker SAS adjustment to >e the cause. Inspection revealed that even though the gap had been set by procedure when the affected breaker was last racked in. the'as found gap was too wide. Thus when this breaker was closed, not all of the contacts in its SAS had changed state. Through further testing and investigation. PG&E discovered that wh>n the adjustment screw at the top of the breaker's SAS plunger 1s-retorqued after replacing adjustment shims, the linkages that operatt the SAS plunger become slightly cocked as joints in the linkages expand to their maximun end float. This condition raises the SAS plunger up as much as 0.050 or 0.060 inch above its normal, breaker open, rest position. During the first subsequent closing operation after the gap has been set with the plunger in the slightly raised position, the end floats all take up which effectively shortens the plunger stroke and thereafter not all of the SAS contacts may change state. To prevent this and ensure the normally consistent SAS plunger stroke PG&E changed piacedures and conducted training to ensure that 'the plunger is tapped down into its fully withdrawn, breaker open re:.t position before setting the plunger to SAS operating rori gap by manually adjusting the breaker elevation in the cubicle.

c. Conclusion With respect to the two 4 kV breaker installation, interface, and performance concerns, the inspectors determined that PG&E's corrective action was appropriate and adequate.

l l 11 l

- 66 n --v, , ,------,,. , ,-,n- ,,n. , , - -n-r,--.--- , , , - - - - , - - - . -

-, . - ~ - . - - _ c .. ,

l PARTIAL LIST OF PERSONS CONTACTED Licensee Shawn LaForce RS Engineer Brad Olson USNRC Project Engineer

Don Allen USNRC Resident inspector

, Stan Ketelsen NSAL RS Supv.

Bill Bayne Proc. Svs Supv. .

Bob Whitgell NOS -

Supv.

Charlie Nichols Materials Director Michael Jacobson NOS Sr. Engineer Chuck Lewis NOS Engineer Bill Colry Reg. Sycs. Engineer Tom Bennett OS Director ,

Dave Taggart NOS Engineering Director

• Dir. & Procurement Thon.o W. Packy NOS-PA Lead Auditor Klemme Herman NTS/ DES /EE Elect. Sup.

Ed Kahler NTS/ Tech PM Tom Fetterman NTS/ES Director Pat Colbert NTS/ES Elect. Sup.

Eric Nelson Mechanical Maint. General Foreman David Detley Maint. Serv. Manager Jim Holden Operations Manager Terry Grebel Reg. Services Director Open Items This report categorizes the-inspection findings as unresolved items and inspection follow up items in accordance with the NRC Inspection Manual.

Manual Chapter 0610. An unresolved item (URI) is a matter about which more information is reMred to determine whether the issue in question is an -

acceptable item, a ceviation, a nonconformance, or a violation. The NRC Office of Nuclear Reactor Regulation will issue any enforcement action resulting from their review of the identified unresolved items, An inspection follow up item (IFI) is a matter that requires further inspection because of a

)otential problem, because specific licensee or NRC action is pending. or

)ecause. additional information is needed that was not available at the time of the inspection, Item Number Finding Title 50-275:50-323/97201 01 URI Meeting ANSI /IEEE C37.59-1991 (Section 5) e 12

p p.a..,k UNITED STATES j NUCLEAR REDULATCRY COMMISSION wAsmwatow o.c. seass aen y July 3, 1997 t

Mr. D. M. Smith, President PECO Nuclear PECO En9tgy Company ,

i Nuclear Group Headquarters Correspondence Control Desk ,

P.O. Box 195 Wayne, Pennsylvania 19087 0195

SUBJECT:

LIMERICK GENERATING STATION ASSURANCE OF VENDOR QUALITY INSPECTION (REPORT NOS. 50 352/96 201 AND 50 353/96 201)

During the periods August 5 through 9,1996, and March 10 through 14,1997, the U.S. Nuclear Regulatory Commission's (NRCs) Office of Nuclear Reactor Regulation (NRR) performed a pilot Assurance of Vendor Quality inspection at PECO Nuclear Offices in Wayne, Pennsylvania. The inspection was related to activities at the Limerick Generating Station, Units 1 and 2. This was the first of a series of pilot inspections being conducted ,

to evaluate the implementation of licensee safety related procurement programs, f

The results of this inspection are contained in the attached inspection report. Overall, the team found that the procurement of safety related items and services was adequately performed and the procurement process was being implemented per program requirements except for the concerns identified in the report.

An unresolved item was identified concerning the the receipt inspection and acceptance of ASME Code, Section ill items from meterial suppliers without the complete documentation  :

required by Paragraphs NCA 3861(b) and NCA 3862.1(b) of Subsection NCA, Section ill l of the ASME Code, in addition, the staff identified a weakness re9erding PECO Nuclear Quality Assurance  ;

(NOA) group review and acceptance of an audit report performed by another utility under the auspices of the Nuclear Utilities Procurement issues Committee (NUPIC) and use, of this_ review of the audit report as the basis for maintaining ACCUTECH es an approved vendor on the PECO Evaluated Vendors List (EVL). The NOA. review of the NUPIC audit report dd not adequately address the commercial grade dedication sampling issues described in the NUPlc audit report for applicability to PECO procurement requirements

. and did not question the basis for verification of lot homogeneity for finished fasteners purchased from non approved suppliers as described in the audit report.

I i

l Mr. D.M. Smith 2-You are requested to respond to the worskness identified in the inspection report. In your response please address,1) your process for placing vendors on the EVL based upon third party review,2) your process for 8dentifying, reviewing and addressing audit findings and followup correspondence that identify issues applicable to your procurements, including  ;

findings from NRC inspections and 3) your assurance that past procurements from vendors, including ACCUTECH, similarly placed on the EVL are adequate based upon the weakness identified. Please send your :ssponse 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 receipt of this letter, in accordance with 10 CFR 2.7g0 of the Commission's regulations, a copy of this letter j and inspection report will be placed in the NRC Public Document Room. Any enforcement '

action resulting form this inspection will be issued by the NRC Region I office via a separate correspondence. Shouid you have any questions concerning the attached inspection report, please contact the inspection team leader Mr. Richard P. McIntyre at (301)415 3215.

Sincerely, ,

P.obert t iallo, Chief Specialin pection Branch Division of Inspection and Support Programs Office of Nuclear Reactor Regulation Docket Nos. 50 352 and 50 353 License Nos.: NPF 3g and NPF 85

Enclosure:

Inspection Report No. 50 352/g6 201 and 50 353/g6 201 cc: See next page

PECO Energy Company Limerick Generating Station, Units 1 & 2 cc: .

J. W. Durham, Sr., Esquire Chief Division of Nuclear Safety Sr. V.P. & General Counsel- PA Dept. of Environmental Resources PECO Energy Company P.O. Box 8469 2301 Market Street Harrisburg, PA 17105 8469 Philadelphia, PA ~ 19101 Manager Limerick Licensin),62A 1 Director Site Engineering PECO Energy Company Limerick Generating Station 965 Chesterbrook Boulevard P.O. Box A Wayne, PA 19087 5691 Senatoga, PA 19464 Mr. Walter G. MacFarland, Vice President Limerick Generating Station Manager Experience Assessment Post Office Box A Limerick Generating Station Senatoga, PA 19464 P.O. Box A Senatoga, PA 19464 Plant Manager Limerick Generating Station Library P.O. Box A U.S. Nuclear Regulatory Commission Senatoga, PA 19464 Region 1 475 Allendale Road Regional Administrator, Region i King of Prussia, PA 19406 U.S. Nuclear Regulatory Commission 475 Allendale Mood Senior Manager Operations King of Prussle, PA 19406 Limerick Generating Station P.O. Box A Senior Resident inspector Senatoga, PA 19464 U.S. Nuclear Regulatory Commission Iimerick Generating Station Dr. Judith Johnsrud P.O. Box 596 National Energy Committee Pottstown, PA 19464 Sierra Club 433 Orlando Avenue Director Site Support Services State College, PA 16803 Limerick Generating Station P.O. Box A Mr, George A. Hunger, Jr.

Senatoga, PA 19464 Director Licensing. MC 62A 1 PECO Energy Company Chairman Nucient Group Headquarters Board of Supervisors Correspondence Control Desk of Limerick Township P.O. Sort No. .'Or, 646 West Ridge Pike Wayne, PA 19087 0195 Linfield, PA 19468

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

Docket Nos.: 50 352 and 50 353 License Nos.: NPF 39 and NPF 85 Report Nos.: 50 352/96 201 and 50 353/96 201 Licensee: PECO Energy Company Facility Name: Limerick Generating Station, Units 1 and 2 Location: Correspondence Control Desk l O. Box 195 Wayne, Pa 19087 0195 Dates: August 5 9,1996 March 1014,1997 Inspectors: Richard P. McIntyre, Team Leader, NRR Uldis Potapovs, Senior Reactor Engineer, NRR Bill Rogers, Reactor Engineer, NRR Approved by: Gregory C. Cwalina, Section Chief Special Inspection Branch Division of Inspection and Support Programs Office of Nuclear Reactor Regulation

l l

EXECUTIVE

SUMMARY

From August 5 through 9,1996, and March 10 through 14,1997, representatives of the U.S. Nuclear Regulatory Commission's (NRC's) Special inspection Branch conducted an inspection of PECO Energy Company (PECO) activities related to the nrocurement of products and services used in safety-related applications at the Limerek Generating Station, Units 1 and 2 (LGS).

The inspection team reviewed PECO's procurement pogram to assess its compliance with the quality assurance (OA) requirements of Appendix B to Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50). The inspection scope included Regulatory Guide 1.123, " Quality Assurance Requirements for Control of Procurement of items and Services for Nuclear Power Plants," and Regulatory Guide 1.144, " Auditing of Quality Assurance Programs for Nuclear Power Plants." Applicable industry standards include the American National Standards Institute (ANSI) N45.3.13, " Quality Assurance Requirements for the Control of Procurement of items and Services for Nuclear Power Plants," a.id ANSI N45.2.12. " Requirements for Auditing of Quality Assurance Programs for Nuclear Power Plants."

The NRC conducted this inspection, the first pilot inspection in this area, using draft inspection Procedure, " Assurance of Vendor Quality," dated July 16,1996, with the intent to use this inspection as input to finalize the inspection procedure for future inspections of this type. The objective of the inspection was to ascertain whether PECO is effectively monitoring the control of quality of safety related products and services by contractors and subcontractors (hereafter referred to as " vendors"). This was done by assessing attributes of the licensee's vendor oversight program and verifying its implementation with regard to selected vendors based upon licensee and vendor documentation, NRC regulations, regulatory guides, and applicable industry standards.

The team reviewed the PECO program and its implementation for the procurement of items and services used in safety related applications at LGS. The team also reviewed the PECO program and its implementation for determination of verification of suitability of those items for their intended or approved safety reisted applications. The inspection included a review of procedures and representative records, including approximately 40 procurement packages for mechanical, material and electricalitems; interviews with PECO staff, including senior management and LGS site pertonnel; and observations by the inspection team members. The inspection team findings were discussed with PECO's representatives and_ senior management at the interim exit meeting held August 9,1996, and the final exit meeting held on March 14,1997.

Overall, the team found that the procurement of safety-related items and services was adequately performed and the procurement process was being implemented per program requirements. However, the team identified a program weakness that concemed the PECO Nuclear Quality Assurance (NOA) group review and acceptance of an audit report performed by another utility in May 1996 under the auspices of the Nuclear Utilities Procurement issues Committee (NUPIC) and use of this review of the audit report as the basis for maintaining ACCUTECH as an approved vendor on the PECO Evaluated Vendors 1

i

-List (EVL). The NOA review of the NUPIC audit report did not adequately address the commercial grade dedication sampling issues described in the NUPIC audit report for applicability to PECO procurement requiraments and did not question th;, basis for verification of lot homogeneity for finished fasteners purchased from non approved suppliers as described in the audit report. PECO did not identify that ACCUTECH's 1 sampling process needed improvement for them to be maintained on the EVL as an approved supplier. The NUPIC audit report described NRC inspection findings and follow up correspondence issues from the December 1994 NRC Vendor inspection Branch j inspection at Cardinal industrial Products (previous name of ACCUTECH). The sampling

, process reviewed at ACCUTECH in May 1996 was essentially the same process reviewed by the NRC in December 1994. The PECO NQA review used the NUPIC audit report for maintaining ACCUTECH on the EVL without adequately addressing the issues described in

< the NRC Notice of Nor.conformance identified during the December '994 inspection .

Another issue identified concerned the recolpt inspection and acceptance of ASME Code,

Section lilitems from material suppliers without the complete documentation required by l Paragraphs NCA 3861(b) and NCA 3862.1(b) of Section 11! of the ASME Code. Paragraph i

NCA 3862.1(b) states that, when the required chemical analyses, tests, examinations, i heat treatment, etc., are subcontracted, the approved suppliers certification for the i operations performed shall be furnished as an identified attachment to the certified

, material test report (CMTR). In several inst 6nces the approved supplier certifications were not furnished with the CMTR and were not included in the document package, i

o i

4 il 73-

d j panort Detai[3 i

Ill Enginsedng E7 Ou611ty Assursace in Enginsedne Activities 4

}.

a E7.1 Evaluatad Vendors List 1 i

s. Inanaction Scone '

j The inspectors reviewed PECO procedure P C 9, " Evaluated Vendors List," Revision i 1, dated December 15,1993, and representative documentation to verify .

implementation. P-C 9 established requirements, assioned responsibilities and ,

i provided guidance for the properation and maintenance of the PECO Evaluated

! Vendors List (EVL), the EVL Conditional Clauses, and the Alert List. The inspectors reviewed PECO procedure NOA 19, "NOA Evaluation of Vendors," Revision 4, dated February 20,1996, and Revision 5, dated August 5,1996, and representative documentation to verify implementation. NOA 19 established requirements, assigned responsibilities and provided guidance for the PECO Nuclear Quatty Assurance (NOA) evaluation of vendors for acceptance on the PECO EVL.

l l b. Qh3pgntions and Findinos The EVL was a listing of vendors which PECO had evaluated to determine their capabilities c? supplying equipment, components, and services, in accordance with the conditions of a purchase order or contract. NOA-19 indicated that PECO i

evaluated vendors by several methods including assessments, commercial grade surveyn, QA Manual reviews, and annual evaluations. Evaluated vendors were listed on the PECO EVL as Approved (A), Conditionally Approved (C), Commercial i Grade (G), or Sidder (8) Approved vendors had a quality program complying with 10 CFR Part 50, Appendix B or another applicable standard and PECO had verified i

[ implementation of the quality program by an initial qualification audit or by a 1

' triennial follow-up audit. Conditionally Approved vendors were required to comply with a set of quality assurance conditional clauses imposed by PECO as a result of verWor evaluation activities. The applicable "C" vendor quality assurance 4

conditional clausas were docurr.ented on the PECO EVL and imposed on PECO POs j to the vendor. In addition, PECO had verified implementation of the quality

[ program by an initial qualification audit or by a triennial follow-up audit fcr the "C"

! vendor. For Commercial Grade vendors, PECO had identified the vendor-controlled

critical characteristics for the products to be purchased from the vendor, had i

obtained satisfa: tory results from a PECO evaluation of applicable vendor quality program documents, and had verified implementation of the quality program by an initial qualification audit or by a triennial follow-up audit for the "G" vendor. For Sidders (B), NQA reviewed the GA program documents to determine that the vendor had the capt.bility to meet 10 CFR Part 50, Appendix B ovelity requirements. The bidder review process was documented on use PECO Vendor QA Manual Checklist.

1 74-

For vendors listed on the EVL as Conditional, the additional requirements, listed in the Conditional Clauses section of the EVL, were imposed on the procurement documents or purchase orders. P C 9 indicated that this status was applicable to vendors who operated with a timited scope 10 CFR Part 50 Appendix 8 program or who required additional requirements on POs based c.n audit, survel: lance, or survey results, but were capable of supplying safety-related items or servi;es. The Vendor Alert List identified ali es to be reviewed concerning equipment and services and associated vendors based on informnion obtained from NRC Bulletins and -

hformation Notices and PECO 10 CFR Part 21 reportable items.

The inspectors reviewed the PECO EVL dated August 5,1996, to verify-implementation of P C 9. The EVL listed approximately 450 vendors including

, name, address, product or service available for PECO use, EVL status, and the ASME status. The vendors whose audits or purchase documents reviewed by the inspectors were determ!ned to be appropriately listed on the EVL. In addition, the inspectors noted that all vendors reviewed, which were listed as Conditionally Approved, had corresponding entries in the Conditional Clauses section which identified conditions to be met,

c. Conclusions The inspectors concluded that PECO had imposed adequate procedural controls on the EVL and that it was well organized and had been maintained to an up to date condition. ,

E7.2 Review of PECO Vendor Amnessment and Surveillance Process

a. Innnection Scone The inspectors reviewed PECO procedure NOA 20, *NOA Vendor Assessments and Surveillance," Revision 7, dated March 8,1996, and representative documentation to verify implementation. NOA 20 established requirements, assigned responsibilities and provided guidance for the coordination, preparation, performance, and reporting of PECO Nuclorr Quality Assurance (NOA) vendor assessments and surveillance.

b. Observations and Findinan NOA 26 defined Pre award assessments as activities performed to evaluate a vendors's QA program prior to placement on the PECO EVL as an approved supplier and issuance of a PO to that ver dor, and the triennial assessments as the activity performed to initially place and continue maintaining the vendor on the PECO EVL as a supplier. The inspectors noted that PECO used the term assessraent to define the process of performing the on-site review of the applicable portions of the vendor's QA program and its implementation, to determine that it meets the applicable requirements of 10 CFR Part 50, Appendix B. NOA-20 defined

" assessment" as a process equal to or exceeding the requirements of "sudit" as defined in ANSI N45.2.

2

The PECO Assessment Team Leader (ATL) prepared an Assessment Plan and checklist listing based on a variety of sources which would indicate areas needing to be reviewed during the assessment which included the "e ' 1A documents, PECO procurement documents, the PECO Vendor Alert LL 7 Nuclear Network, NRC Vendor inspection Reports, previous asse - . s.  % EVL corrective actions requirements, and nonconformance du TN Asa sment Plan was to be performance based using methods such as cori T a porf t. 1nce objectives and acceptance criteria to the final result and o. A Men mgoing ,

activities. In addition, NOA 20 indicated that the NUPlc Perf. ~ w . wd Supplier Audit Checklist should be used in developing the Asse .r.W Aen when appropriate. The ob}ective in' development of the Assessmc.nt Pbri ' .o determine what vendor program and processes should be evaluated during the assessment to verify product acceptability for its intended application based on the products essential function, product configuration, essential parts and components, vnd critical characteristics. The inspectors reviewed several Assessment Plans and concluded that they effectively implemented program requirements.

The assessment consisted of a pre assessment conference, assessment performance, and a post assessment conference. Nonconformances identified during the assessment were documentr* on Vendor Correctivs Action Requests (VCR) and if the nonconformance couk n otentially affect hardware the ATL would request that the vendor document the identified nonconformance on a vendor Noncoaformance Report if the vendor dispositioned the nonconformance as

" acceptable as is" or " repair" in accordance with the vendor's QA program the Assessor requested that the vendor initiate a Vendor Deviatica Request (VDR) as required by the PECO PO PECO issued the Assessment Report, within 30 days of the completion of the assessment, which included and Assessment Summary

. (scope, assessment results, stronaths and weaknesses, recommendations, and evaluation of activities), Investigt: on Results (activity investigated, summary of acceptable and unacceptable resu.cs, reference to VCRs and recommendations (REC)), and Administrative Details (names of PECO personnel involved, vendor personnel contacted during the assessment, and VCRs and 3ECs). PECO required the vendor to respond to the issued VCRs, reviewed the received responses, and informed the vendor of the results. The inspectors reviewed several Ausssment reports (discussed in Section E.7.3 of this report) and concluded that they effectively implemented the program requirements.

Surveillance were performed to meet a PECO EVL Conditional Approval requirement which indicated that the vendor contact PECO prior to performir's a certain portion of the work so that PECO could perform a surveillance of the activities.

Surveillance were also performed to meet PECO safety related PO requirements which specified vendor surveillance requirements. Advance preparation for the Surveillance include development of a Surveillance Checklist to be approved by the Working Lead and contacting the vendor to initiate ond coordinate activities. The Surveillance performance included observation and evaluation of activities and objective evidence for conformance with the PO requirements and completion of the Surveillance Checklist. Nonconformances identified during the surveillance were documented on a VCR and if the nonconformance could potentially affect hardware the Assessor would request that the vendor document the identified nonconformance on a vendor Nonconformance Report, if the vendor dispositions the as " ace eptable as is" or " repair" in accordance with the vendor's QA program 3

the Assessor requested that the vendor initiate a VDR as indicated in the PECO PO.

After completion of the Surveillance the Assessor completed a Certificate of Surveillance including the PO item number, quantity, description, and PECO nuclear code number for the item being released. A copy of the Certificate of Sarveillance was provided to the vendor for inclusion into the documentation package shipped with the product.. Any VCRs generated by the surveillance were provided to the vendor for response accompanying a PECO Nuclear Transmittal Letter issued within 30 days of the completion of the surveillance. The inspectors reviewed documentation associated with several surveillances and determined that the program requirements were being effectively implemented,

c. Conclusions The inspectors concludad that PECO had developed adequate procedural l requirements to establish a program to effe*:tively perform assessments and surveillances to support verification of quality activities and that PECO had generally implemented these procedural controls on the performance of  ;

assessments and surveillance.  !

l l

E7.3 Review of Selected PECO Assessments and Surveillance

a. Insoection Scone The inspectors review of a listing of PECO performsd assessments indicated that PECO typich:ly performed twelve assessments yearly. A large portion of vendor placemen* on PECO's EVL was based on PECO's formal review and acceptance of NUPIC Joint Utility Audits or NUPIC Member Audits, in addition, PECO had led a number of NUPIC Joint Utility Audits recently including GE-Fenton, Bechtel, Crane Chem Pump, Amerace, ARCOS, Leeds & Northrup, Overly, NUS, and Ingersoll-Rand. To verify implementation of PECO vendor assessments, the inspectors reviewed the following NUPIC member audits and the accompanying PECO NQA reviews of theses audits.-

b. Observations and Findinom b.1 ACCUTECH The inspectors reviewed the Comanche Peak Steam Electric Station TU Electric (TUE) QA Audit Report QAA 96-010 of ACCUTECH dated May 23,1996. The audit, conducted April 2g through May 2,1996, was led by TUE and performed in accordance with the requirements of the TUE OA program, under the auspices of-NUPlc. The audit also included representatives of Iowa Electric Services (lES),

Houston Light and Power, and Northern States Power. The audit was performed to essess ACCUTECH's OA program and its implementation in supplying ASME code and non-code materials to the applicable requirements of 10 CFR Part 50, Appendix B, ASME NOA 1, and ASME Section lit, Subsection NCA 3800. The audit included the areas of Order Entry: Commercial Grade Dedication; Procurement; Material Control / Handling Storage, and Shipping; Fabrication, Assembly, and Special Processes; Tests and inspection; Calibration; Document Control / Procedure Adequacy; Organization / Program Compli6nce; Nonconforming Conditions; Corrective Actions; internal and External audits; and Training / Certification.

4

PECO performed a formal review of the TUE audit of ACCUTECH as documented on PECO Audit Report Review Form (ARRF) dated June 27,1996. The ARRF indicated that TUE had performed an audit of the areas applicable to PECO's planned purchases using the appropriate criteria (10 CFR Part 50, Appendix 8, NOA 45.2 and NOA 1). The ARRF required a review of numerous areas including Lead Auditor certification, audit scope, applicability of CA program, sufficient objective evidence in audit package, audit findings issued and applicability of items on orderiin stock, or installe.1 in the plant, and whether corrective actions for p findings were r dequate for the PECO application. PECO had concluded that the TUE audit was acceptable for PECO's purchases and supported ACCUTECH's status on the PECO EVL, however, the review did require that specific information be added to the EVL for PECO information only. This information dealt with,

. among other things, a description of the sample plans implemented by ACCUTECH. 3 The inspectors reviewed the applicable sections in the TUE sudit report that addressed the nonconformances and issues identified in the December 1994 NRC Inspection (Report No. 99901076/94 01) at Cardinal Industria! Products (former name of ACCUTECH), concerning commercial grade dedication, and sampling relating to commerc;al grade item (CGI) dedication. The sadit reviewed the applicable portions of the ACCUTECH program as it relates to sampling as part of commercial grade dedication; however, it did not appear that the NUPlc audit verified or attempted to verify the ACCUTECH basis for their sampling plans utilized for destructive and nondestructive testing. This issue had been identified as a nonconformance in the December 1994 NRC inspection at Cardinal. Also, when reviewing the documentation on the NUPIC audit checklist, the inspectors did not identify any evidence that the auditors reviewed ACCUTECH's rationale for verifying lot homogeneity for finished fasteners that are purchased from non-approved suppliers. This method, in turn forms the basis for ACCUTECH's selection of the CGI sampling plans.

PECO's review and teceptance of the NUPlC audit report of ACCUTECH did not #

question the sampling plans implemented by ACCUTECH or address the fact that the audit report identified that the NRC had issued a nonconformance to ACCUTECH concerning CGI dedication sampling deficiencies in December 1994 and they were stillimplemer, ting basically the same dedication sampling process in May 1996. PECO stated that part of the reason the TUE audit was accepted was that the audit report concluded that the sampling plan (s) invoked by the ACCUTECH progrcm for the products reviewed were deemed to be adequate. The inspectors stated that the TUE audit report described the NRC findings and follow-up correspondence from the December 1994 inspection at Cardinal and that the sampling process reviewed at ACCUTECH in May 1996 was essentially the same process reviewed in DeceT. bur 1994. In their formal AARF review conducted June 27,1996, PECO did not identify that ACCUTECH's sampling process needed i improvement for them to be maintained on the EVL as an approved supplier.

Review of the August 5,1996, PECO EVL showed ACCUTECH listed as a Conditionally Approved supplier of nuclear fasteners and materials with two vendor conditional approval (CA) clauses describing the requirement of implementation of the Quality Systems Manual to be included on purchase orders and also included, among other items, the various sampling plans used by ACCUTECH under "For PECO information." Since the initial inspection, PECO had made several changes to 5

the CA requirements to include specific information on commercial grade dedication and sampling.- PECO stated that these revisions had been initiated prior to the NRC inspection in August 1996.

PECO stated they received NRC Information Notice (IN) 96-40 on July 31,1996, and they initiated research on actions to be taken to address commercial grade dedication program weaknesses described in IN 96-40. As a result of this evaluation and further review of vendor issues contained in IN 96-40 and in Vendor inspection Reports, PECO revised their EVL CA clauses to define PECO Nuclear expectations for acceptable commercial Stade dedication, particularly in the area of sampling. These CA requirements were invoked on ACCUTECH and Allied Nut and Bolt for all future purchase orders in an August 20,1996, letter. Several other vendors were also notified of the revised EVL CA requirements in the September -

October 1996 time frame, in December 1996, letters were sent to each of the applicable vendors raquesting a copy of their procedures used to implement the CA requirements. The inspectors verified that the vendors who did not submit their procedures to PECO Nuclear NOA for review were either downgraded or removed from the EVL.

During the March 1997 inspection at PECO, the inspe: tors discussed concems with the use of the term " heat lot" traceability in the CA requirements and the resulting potential misinterpretation of what would be required for heat lot traceability regarding sampling to meet these CA requirements. FECO agreed that the CA requirements could be clarificd and did so prior to the completion c1 the inspection.-

l b.1.1 Vendo Insnection at ACCUTICH ,

l After the August 1996 inspection at PECO, follow up vendor inspections were performed at ACCUTECH in November 1996 and January 1997 and documented in inspection Report 99901307/96 01, dated March 4,1997. During the November 1996 inspection at ACCUTECH the inspectors reviewed the sample plan i methodology as part of the commercial grade dedication process currently in place.

The inspectors determined that it was basically the same process that was reviewed during the NRC's December 1994 inspection at Cardinal, in that it places heavy reliance on visual and dimensional inspection to support the verification of lot homogeneity. No revisions had been implemented in the ACCUTECH OA program that supported the sampling process rationale described in various 1305 correspondence to the NRC, especially the August 30,1995, letter. This letter was the last to formally respond to the sampling nonconformance identified in NRC Inspection Report No. 99901076/94 01.

During the November 1996 NRC inspection ACCUTECH could not provide a documented basis to support the sampling information that was described to the NRC in the August 30,1995 letter. ACCUTECH then stated that they had recently written a " white paper" that described proposed changes to the ACCUTECH sample plan methodology for testing and sxamination. This document was dated November 7,1996, but had not as yet been implemented as part of the ACCUTECH OA program.

6

l l

In conclusion, the inspectors determined that ACCUTECH placed heavy reliance on visual and dimensional inspection to support the verification of lot homogenelty.

Based upon this method for verification of lot homogeneity, ACCUTECH then utilized the ASTM A 325 shipping lot sampling plan for destructive testing (material chemistry and mechanical properties) and the EPRI guidelines for nondestructive testing (dimensional). The inspectors determined that visualinspection for shipping damage and manufacturing defects can not assure that allitems in the same product lot were manufactured from the same heat of material or were heat treated under the same conditions and that, as discussed in NRC Inspection Report 99901076/94 01 and related correspondence, the use of ASTM A 325 shipping lot sampling plan is inappropriate for this application.

b.2 Draoon Valves The inspectors reviewed the Illinois Power Company (IPC) Auriit Report of Dragon Valves, Inc. (Dragon), dated October 27,1995, which documented the September 25 28,1995, nuclear assessment of Dragon. PECO had perfo'med a review and accepted the Audit Report on December 13,1995, as the basis for placing Dragon on the PECO EVL as an approved supplier.

IPC had led and porformed the audit as a NUPlc Joint Utility Audit with Commonwealth Edison, Entergy Operations, and Illinois Power providing audit team members. The audit was peiformed to assess Dragon'a QA program and its implementation in supplying ASME code and non-code valves, parts, and components to the applicable requirements of 10 CFR Part 50, Appendix B, and ASME Section ill, NCA 4000. The audit included the areas of Order Entry; Procurement: Material Control / Handling, Storage, and Shipping; Fabrication, Assembly, and Special Processes; Tests and inspection: Calibration; Document Control / Procedure Adequacy; and Program Compliance.

The audit report concluded that Dragon was implementing an effective program for the product to be supplied with the exception of two findings that wue not consioered significant enough to have any major impact on work previously performed by Dragon. When reviewing the NUPlC audit package the inspectors noted that the audit identified that Dragon does not audit vendors who hold current ASME Quality System Certificates (OSCs) and also that Dragon does not perform any commercial grade dedication. It further stated that CMTRs for material received from unapproved suppliers and OSC holders is validated through receip't inspection and independent chemical analysis performed by an approved supplier.

PECO had performed a formal review of the IPC audit of Dragon as documented on PECO ARRF dated November 13,1995. The ARRF indicated that IPO had performed an audit of the areas applicable to PECO's planned purchases using the appropriate criteria (10 CFR Part 50, Appendix B, and ANSI N45.2 ). The ARRF required a review of numerous areas including Lead Auditor certification, audit scope, applicability of QA program, sufficient objective evidence in audit packace, audit findings issued and applicability of items on order, in stock, or installed in the plant, and whether corrective actions for findings were adequate for the PECO application. PECO had concluded that the IPC audit had been applicable to PECO's purchases, that the findings and PECO's concerns had been adequately addressed, and PECO had sufficient basis for p;ocement of Dragon on the PECO EVL with 7

80-

conditional PO requirements. Review of the August 5,1996, PECO EVL showed Dragon as a conditionally approved supplier of nuclear valves and replacement parts, with certain conditional clauses listed.

During the NRC inspection at Dragon in September 1996, the NRC reviewed specific program implementation areas and identified severalinstances whue activities were not conducted in accordance with NRC and ASME code requirements. These included; (1) failure to provide adequate control of material ,

procurement procedures and to prepare the Unqualified Source Materiallist; (2) i failure to adequately verify supplier qualifications for certain ASME quality system certificate holders; (3) failure to establish and document the basis for chemical sampling plans used for dedicating unqualified material and; (4) failure to perform and document the required actions for a significant condition adverse to quality.

However, these specific findings were not against PECO purchases, b.3 ASTA Enoineerinac Inc.

The inspectors reviewed the PECO Assessment Report of ASTA Engineering. Inc.

(ASTA), dated September 22,1995, which documented the August 22 23,1995, assessment of ASTA. PECO had performed the assessment using a performance based approach using the NUPIC Joint Audit Checklist, Revision 6, dated March 26,1995. The Audit Team Leader and the Technical Specialist were both -

Gilbert / Commonwealth, Inc., personnel contracted by PECO to perform the assessment and prepare the Assessment Report. l l - The assessment was performed to address ASTA's OA program and its implementation as applicable to engineering ervices supplied to PECO. The portion of the assessment covered by the section of the Assessment Report titled

" investigative Summary" included the areas of Order Entry: Design; Software Quality Assurance; Procurement: Material Control, Handling, Storage, and Shipping; Fabrication, Assembly, and Special Processes; Tests and Inspection; Calibration; Document Control; and Program Control. The portion of the assessment covered by the section of the Assessment Report titled " Technical Specialist Summary"-

discussed the Technical Specialist's review of three ASTA packages for the dedication of molded case circuit breakers, dedication of Potter & Brumfield MDR relays, and the dedication of control transformers.

The PECO assessment concluded that ASTA was implementing an effective program for the services to be supplied to PECO with the exception of two areas identified in Vendor Corrective Action Requests (VCR). The VCRs documented that ASTA had not passed down a requirement (10 CFR Part 21), imposed by customer PO, to a sobtier supplier as required by the ASTA QA program and ASTA had not taken actions to approve a calibration facility used in quality activities.

PECO provided documentation of the correspondence between PECO and ASTA documenting acceptable closure of both findings. Review of the August 5,1996, PECO EVL showed ASTA as an approved supplier of engineering consulting services and commercial grade dedication of equipment. There were no conditional clauses listed for ASTA.

8

b.4 Namco Controls Corooration The inspectors reviewed the IES Utilities (lES) Audit Report of Namco C. rois Corporation (Namco), dated January 25,1996, which documented the January 812,1996, assessment of Namco. PECO had performed a review and '

accepted this Audit Report as a basis for placing Namco on the PECO EVL as an approved supplier.

IES had led and performed the audit as a NUPIC Joint Utility Audit with lilinois Power, Tennessee Valley Authority, and IES providing audit team members. The audit was performed to assess Namco's OA program and its implementation in  !

supplying electromechanical limit switches, limit switch receptacle assemblies, plug-in connector assemblies, and cable assemblies. The audit ir.cluded the areas of Order Entry; Design; Procurement; Material Control, Handling, Storage, and Shipping; Fabrication, Assembly, and Special Processes; Tests and Inspection; 1 Calibration; Document Control; and Program Control.

The audit report concluded that Namco was implementing an effective program for the product to be supplied with the exception of four areas identified in one finding and three observations. The finding indicated that Namco did not have procedural controls which adequately established storage temperature requirements. The observations indicated that Namco did not have documentation to support a

- referenced activation energy, did not specify certain management personnel in the organization chart, and the Namco QA program did not specify the frequency of interna! audits and how resultant findings were to be addressed. - PECO provided documentation of the correspondence between IES and Namco documenting acceptable corrective action for the finding and three observations.

PECO had performed a formal review of the IES sudit of Namco as documented on PECO Audit Report Review Form (ARRF) dated February 20,1996. The ARRF indicated that IES had performed an audit of the areas applicable to PECO's planned purchases using the appropriate criteria (10 CFR Part 50, Appendix B, ANSI N45.2 and NOA 1). The ARRF required a review of numerous areas including Lead Auditor certification, audit scope, applicability of QA program, sufficient objective evidence in audit package, audit findings issued and applicability of items on order, in stock, or installed in the plant, and whether corrective actions for findings were adequate for the PECO application. PECO had concluded that the IES audit had been applicable to PECO's purchases, that the findings and observations had been adequately addressed, and PECO had sufficient basis for placement of Namco on the PECO EVL. Review of the August 5,1996, PECO EVL showed Namco as an approved supplier of nuclear limit switches and spare and replacement parts, with no conditional clauses listed.

b.5 Sorrento Electronics The inspectors reviewed the Wolf Creek Nuclear Operating Corporation (Wolf Creek) Audit Report of Sorrento Electronics, Inc. (Sorrento), dated July 7,1994, which documented the June 20 24,1994, audit of Sorrento. PECO had performed a review and accepted this Audit Report as a basis for placing Sorrento on the PECO EVL as an approved supplier.

9

l Wolf Creek had led and performed the audit as a NUPIC Joint Utility Audit with APS, TUE, WCN, and PSE providing audit team members. The audit was performed to assess Sorrento's QA program and its implementation in supplying radiation monitoring devices. The audit included the areas of Order Entry; Design; l Software Quality Assurance; Procurement: Material Control, Handling, Storage, and '

Shipping; Fabrication, Assembly, and Special Processes; Tests and Inspection:

Calibration; Document Control; and Program Compliance, l'

The audit concluded that Sorrento was implementing an effective program for the product to be supplied with the exception of the four findings identified. The findings included not including functional testing as a critical characteristic of an active electrical component; not assuring that items and services conformed .o the procurement documents; inconsistencies in the software quality assurance program; and a measurement gage being calibrated at longer intervals than specified in the applicable QA procedure. PECO provided documentation of the correspondence between Wolf Creek and Sorrento documenting acceptable corrective action for the findings.

i PECO had performed a formal review of the Wolf Creek audit of Sorrento as documented on PECO Audit Report Review Form (ARRF) dated October 21,1994.

The ARRF indicated that Wolf Creek had perfctmed an audit of the areas applicable to PECO's planned purchases using appropriate criteria (10 CFR Part 50, Appendix B, and 10 CFR Part 21). The ARRF required a review of numerous areas including Lead Auditor certification, audit scope, applicability of QA program, sufficient objective evidence in audit package, audit findings issued and applicability of items on order, in stock, or installed in the plantiand whether corrective actions for findings are aderwate for the PECO application. PECO had concludert that the Wolf Creek audit was had been applicable to PECO's purchases, that the findings had i been adequately addressed, and PECO had sufficient basis for placement of Sorrento on the PECO EVL.

Review of the August 5,1996, PECO EVL showed Sorrento as a conditionally approved supplier of nuclear radiation monitoring systems and spare and replacement parts. A conditional clause in the EVL specified four requirements to included on PECO PCs to Sorrento: (1) Specific revision of Sorrento QA manual to be applied, (2) Commercial grade items were to be dedicated by a specific Sorrento procedure, (3) All material and documentation was to be shipped from San Diego Manufacturing facility, and (4) a PECO surveillance was required for Full Assembly Radiation Monitor Pos. The inspectors reviewed a recent purchase order to Sorrento PO No. LS 607244, dated March 13,1996, for a General Atomic Processor component and determined that it contained the four conditional clauses as required by the PECO EVL.

c. Conclusions c.1 ACCUTECH The inspectors concluded that the NQA third party review of the NUPIC audit report did not adequately address the commercial grade Jedication sampling issues described in the NUPIC audit report for applicability to PECO procurement requirements and did not questior' the basis for verification of lot homogeneity for 10

finished fasteners purchased from non approved suppliers as described in the audit report. PECO did not identify, in their formal AARF review conducted June 27, 1996, that ACCUTECH's sampling process needed improvement for them to be maintained on the PECO EVL as an approved safety related supplier. The inspectors also concluded that PECO reviewed and accepted the NUPIC audit report without adequately addressing the NRC Notice of Nonconformance and follow up issues identified during a December 1994 Vendor inspection Branch inspection at Cardinal Industrial Products (previous name of ACCUTECH). This issue was identified as a program weakness, c.2 Drsoon Vgyag The inspectors concluded, based on the documentation t. <iewed, that PECO had taken adequate actions to place Dragon on the PECO EVL as a conditionally approved supplier of ASME code and safety related non-code valves, parts, and components and had adequately documented these actions. Also, the inspactors verified during review of PECO POs to Dragon, that PECO had placed the appropriate conditional clause requirements in the Pos.

c.3 ASTA Enoineerino. Inc.

The inspectors concluded, based on the documentation reviewed, that PECO had taken adequate actions to place ASTA on the PECO EVL as an approved supplier of engineering consulting and commercial grade dedication services and had adequately documented these actions.

c.4 Namco Controls Corocration The ibspectors concluded, based on the documentation reviewed, that PECO had taken ader.uate actions to place Namco on the PECO EVL as an approved supplier of nuclear limit switches and spare and replacement parts and had adec.uately documented these actions.

c.5 Sorrento Electronics The inspectors concluded, based on the documentation reviewed, that PECO had taken adequate actions to place Sorrento on the PECO EVL as an approved supplier of nuclear radiation monitoring systems and spare and replacement parts and had adequately documented these actions.

E7.4 Reauisition of items From Other Utilities

a. Insoection Sagga The inspectors reviewed PECO procedure P-C-1, " Material Requisition Process,"

Revision 3, dated August 27,1995, and representat!ve documentation to determine the process used to requisition safety-related material from other utilities.

11

p. Observations and Findinas P-C 1 established purchase classification requirements and stated in Section 4.16 that a Purchase Class 1 item is a safety related item that shall be procured from a vendor listed as approved or conditionally approved on the EVL or from a licensed commercial nuclear facility. Section 7.3, " Requisition of items from Other Utilities," describes the documentation and quality requirements necessary for purchas!rm . safety-related items from other Utilities. During the first phase of the inspection PECO stated that they do not wdit other nuclear utilities for placement on the EVL, based upon the fact that they consider their QA programs as approved by the NRC by acceptance of Chapter 17 of their FSAR/USAR and done in combination with the quality and documentation requirements invoked BY PECO on the selling utility.

The inspectors determined that: PECO compares the PO used by the original purchasing utility to PECO's requirements; copies of the vendor PO documentation is obtained; the supplying utility is required to certify the item and its documentation was procured, received, and stored in accordance with the quality program and was not modified and; PECO ensures that the PO includes a statement requiring that PECO be notified of any deviations from the purebase requirements i prior to shipment.

l

c. Conclusions The inspectors concluded that this method was consistent with 10 CFR part 50, Appendix B requirements.

E7.5 Control of Purchased Material

a. lDangetion Scone The inspection scope incicded a reviev? of PECO's control of purchased material with emphasis on the effectiveness of measures for assuring conformance with procurement document requirements. This phase of the inspection focused on receiving inspection activities for verifying the a64uscy and completeness of vendor supplied documentation. The control of these activities is described in Section 17.2.7, " Control of Purchased Material, Equipment, and Services," of the LGS UFSAR.

As a part of this review, the inspectors selected a sampling of purchase orders (PO) for safety related material that had been processed within the last year. The purchase orders and vendor supplied documentation available at the plant site were reviewed to assess the effectiveness of this phase of the procurement process.

b. Observations and fmdinas A representative list of the document packages reviewed and inspector's observations are included in the following paragraphs.

12 l

t b.1 PECO PO LS 605919, dated January 1,1996, to ACCUTECH Division of B&G Manufacturinc Co. Inc.

Item 1 of this PO was for 44, all thread 1-1/4 inch by 10 inches long, ASME SA-193 Grade B7 studs. The PO specified that the material must be manufactured and controlled through manufacturing and supplying (warehouse / delivering) under a quality program which satisfies the requirements of Section NCA 3700/NCA 3800 of Section 111 of the ASME Code.

The documentation package for this materialincluded ACCUTECH's certification that the material had been manufactured in accordance with their ASME Quality Systems Certificate (OSC). The certification included ladie analysis (apparently transcribed from another document), tensile properties, impact properties, and a description of the heat treatment. The documentation package also included ACCUTECH's certificate of compliance with the applicable ASME Code and 10 CFR Part 50, Appendix B requirements and statements that impact specimens were prepared in accordance with the applicable Code requirements, a satisfactory macroetch test had been performed on this material and that ACCUTECH had conducted a satisfactory visualinspection (report attached). The documentation package also included a copy of PECO's receiving inspection report indicating that complete documentation was provided.

The inspector noted that, since this material was procured in accordance with the requirements of NCA 3800 of the ASME Code, and so certified under ACCUTECH's OSC, additional documentation should have been provided by the Material Organization (ACCUTECH) and maintained at the plant s,ite. Specifically, Paragraph NCA 3862.1(b) states that, when the required chemical analysis (including the meliing mill heat analysis), heat treatment, tests, examinations, or repairs are subcontracted, the approved supplier's certification for the operations perforrned shall be furnished as an identified attachment to the certified material test report (CMTR). From the documentation supplied by ACCUTECH, it could not be readily determined which of the operations described on their CMTR were subcontracted, but it appeared that the melting mill, who is responsible for providing the heat analysis for this material, would fallinto that category. ACCUTECH's CMTR did not include the mciting mill certification, including heat analysis, as an identified attachment.

Failure to verify conformance with procurement document rsquirements was identified as an example of Unresolved item 50 352:50-353/96 20101.

b.2 PECO PO LS 607801, dated March 28,1996, to ACCUTECH.

Items 1 and 2 of this PO were for ASME SA 193, threaded studs with diameters of 5/8 inch and 1 inch, respectively. Item 3 was for 1/2 inch heavy hex bolts and item 4 was for ASME SA-194, Grade 2H, heavy hex nuts. The PO specified that the material must be manufactured and controlled under a quality program which satisfies the requirements of Section NCA 3700/3800 of Section ill of the ASME Code.

For items 1 through 3, PECO,s document packages included ACCUTECH's certification that this material had been manufactured in accordance with their 13

y i

I OSC. The certification included ladie and check analyses, apparently transcribed from another document, tensile properties, and a description of the heat treatment.

PECO's documentation also included ACCUTECH's certificate of compliance with the applicable PO, ASME Code and 10 CFR 50, Appendix 8 requirements, a statement that macrootch test results were acceptable, and a statement that ACCUTECH had conducted a satisfactory visualinspection (report attached). The data package for item 4 included similar information, except that proof load and hardness test results were reported instead of the tensile properties. There was no indication that a macroetch test had been performed on this material as required by the specification.

The inspector noted that, as discussed in Section b.1 above, additional documentatien was not provided by the Material Organintir,n and maintained at the plant site. The molting mill certifications, including heat arWyses and, for item 4, macroetch certification were not provided with the matoriol.

Failure to verify conformance with procurement document requirements was identified as an example of Unresolved item 50 352:50 353/96 201 01.

b.3 PECO PO LS 157145, dated November 26,1996, to Energy & Process Corporation (E&P).

i ltem 1 of this PO we for four 3 inch, NPS, buttweld, schedule 40, seamless SA-234, Grade WPB o sws, to be provided in accordance with ASME Code, Section Ill, Class 2.

The document package for this materialincluded E&P Material Certification which certified compliance with the ASME Code and PO requirements and noted that the material was suppiied in accordance with E&P's ASME QSC. Included with this certification was a Material Test Report from Ladish Company, dated March 23, 1994, containing chemical analyses and tensile properties of the elbows and noting that the starting material was seamless pipe. Also included with the certification was a Magnetic Particle inspection Report from Gramlich Inspection Services, item 2 of this PO was for four 3/4 inch NPS, forged stainless steel, SA 182, Grade 316L couplings, to be suppli:,d in accordance with ASME Code, Section lil, Class 1.

In addition to E&P Material Certification, the documentation package included a Test Report Certification from Alloy Stainless Products Co., dated October 7,1992, end a Liquid Penetrant Report from Gramlich inspection Services.

Item 3 of this PO was for two 1 inch NPS, forged SA 105 couplings, to be supplied in accordance with ASME Section Ill, Class 2.

In addition to E&P Material Certification, the documentation package included a Certificate of Analysis from Capitol Manufacturing Co., dated St.ptember 2,1992, containing chemical analyses and tensile and hardness test results.

item 4 of this PO was for two 1 inch NPS, SA-105 tees, to be supplied in accordance with ASME Code, Section lil, Class 2.

14

I In addition to E&P Material Certification, the document package included a CMTR

. from Bonney Forge, dated September 30,1981, containing chemical and tensile

' test results and certifying that the material was supplied in accordance with their ASME QSC, which expires on March 30,1982. The certification stated that the

material property dets was either copied from material records furnished by the

- production mill or obtained from laboratory checks.

}

The inspector noted that, as discussed in Section b.1, above, additional documentation was not provided by E&P and maintained at the plant site. For ltems 1 through 4, the material producing mill certifications, including host analyses

. were not provided with the material.

Failure to verify conformance with procurement document requirements was identified as an example of Unresolved item 50 352;50 353/96 20141.

b.4 PECO PO LS 158648, dated February 7,1997, to Nova Machine Products

Corporation (Nova).

Item 1 of this PO was for tan 1 1/8 inch ASTM A 325, Type 1, heavy hex bolts, to

be supplied in accordance with the applicable provisions of 10 CFR Part 50, i Appendix 8. The PO also required that a CMTR be provided, including the results of all required chemical analyses, tests, and examinations.

! The document package for this materialincluded Nova Certificate of Compliance i which identified the material by heat number, described the heat trastment, and

provided a quality program statement attesting compliance with the PO requirements and referencing Nova QA manual and their 1S0 9001 Certificate.

1 Attach, i to the Nova certification was a CMTM issued by Lake Erie Screw

{ Corporation.' This CMTR identified the steel producing mill and contained chemical

analysis and mechanical properties of the material. It also provided heat identification, described the head markings, and certified that the material was produced in accordance with a QA program that had been surveyed and approved by Nova, b.5 PECO PO LS 155212, dated August 28,1996, to Nova.

Item 1 of this PO was for twenty four 3/8 inch ASTM A 307, Grade A, hex head bolts, to be supplied in'accordance with a QA program meeting the applicable provisions of 10 CFR Part 50, Appendix B. A Certificate of Conformance was required to be provided for this material.

The document package for this materialincluded a Nova Certificate of Compliance describing the material and attesting to compliar ce with the PO and the applicable specifications. It also included a quality program statement, certifying to compliance with the applicable portions of the Nova OA manual and stating that the material has been processed per ISO 9001(94), Certificate # GOC 211.

15

c. Conclusions A review of documentation packages for meterial purchased to the ASME Code

- requirements identified several examples of appefent failure to assure that the Material Organizations supplying this material provided all of the Code required documentation. Specifically, Paragraph NCA 3861(b) of Section ill of the ASME Code requires the Material Organization to transmit all certifications received from other Material Organizations or approved suppliers to the purchaser at the time of shipment. Paragraph NCA 3862.1(b) states that, when the required chemical analyses, tests, examinations, heat treatment, etc., are subcontracted, the approved suppliers certification for the operations performed shall be furnished as an identified attachment to the CMTR. As discussed in the above examples, in ,

severalinstances the approved supplier certifications were not furnished with the CMTR and were not included in the document package. This issue was identified as Unresolved item 50 35:450 353/96 201 01.

Document packages for safety related material purchised for non Code applications were also reviewed. Such material was procured to 10 Ci1 Pan 50, Appendix B requirements and was ordered to different quality levels, apparently determined by the service application or (for replacement material), the original equipment or design specification. Documentation requirements for these orders were defined by specific purchase clauses which are a part of the PO. It was noted that1ome material orders required only the supplier's certificate of conformance, while others required traceable CMTRs including the actual results of all required chemical anelyses and examinations.

l E7.4 Review of Procurement from Amer Industrial Technolonies. Inc. (AITI

a. Insoection Scone The inspectors reviewed the April 26 through 30 and May 5,1993, PECO triennial audit of AIT. The audit was performed using the NUPIC Audit checklist, Revision 4, for review of the supply of piping subassemblies, pressure vessels as nuclear service components and pipe shop fabrication. The PECO audit did not identify any audit findings and concluded that the AIT quality program for the scope of supply was adequate and being effectively implemented. An inspection conducted by the NRC's Special Inspection Branch in January 1996 (Report No. 99901292/96-01) identified numerous inspection findings in several areas of program implementation and came to significantly different conclusion on effectiveness of program than the PECO audit.

The purpose of the inspector's review was to determine why such a difference in the results of the PECO audit and the 1996 NRC inspection existed and to review PECO's evaluation of the NRC inspection findings for relevance to purchased material and what (if any) compensatory measures were taken by PECO in 1996 as a result of this evaluation.

16

p. Observations and findinos The inspectors reviewed the 1993 PECO audit and material supply histcry of AIT with PECO and determined that PECO had not %ntified problems with AIT supplied material received to date. The inspec.s:s determined that AIT was downgraded from approved to bidder status (not qualified for nuclear purchases) on the PECO EVL based upon the negative results of an ASME survey conducted at AIT on June 26 28,1995. When PECO became aware of the NRC inspection findings at AIT, they initiated different actions to determine: (1) how the findings affected delivered items and components such as the Limerick RHR hea:

exchanger, (2) a review of the NRC inspection report and the effectivenass of the PECO supplier audit process, and (3) an engineering evaluation of the NRC. ASME findings for their effect on the operability of the Limerick RHR heat exchanger.

As a result of those actions PECO issued a Nonconformance report and requestec' that engineering perform an operability /reportability determination and a review of the ASME code issues identified in the NRC inspection report for the RHR heat exchangers at Limocick Unit 1. This evaluation determined that the RHR hest exchangers were acceptable for current and continued use for all modes of operation at Umerick Unit 1. This conclusion was based on the fact that PECO did not solely rely on the audit results at AIT for vendor qualification, but also subcontracted with UE & C Nuclear, for 18 source surveillance inspections at AIT between May 1993 and February 1994, during various stages of manufacturing and testing. Also, the NRC inspection report was reviewed to develop a lessons leamed perspective and training sessions were conducted with all auditors on a lessons learned on vendor audit metnodology covering areas such as ASME Code Section til upgrading of material as well as all of the areas where findings were identified with QA program implementation. Finally, on July 31,1996, AIT was removed from the PECO EVL.

The inspectors also determined that previously, in August 1993, Corrective Action Request (CAR) Q 4320/Ouality Evaluation Q0004320 was initiated by PECO to conduct a self assessment to document and resolve problems associated with the NOA Assessment Section's acceptance of supplier audits. As a result of the activities associated with the CAR,13 problem audits were identified. The PECO audit of AIT was one of the problem audits identified in the review. Each of the audits identified required either additional actions or compensatory measures to be taken. However, NQA personr el at the time of the audit, accepted these audits for use without imposing any additional actions or compensatory measures. A root cause analysis was performed utilizing the " events and Causal Factors" methodology and resulted in the identification of 14 causal factors. Several corrective measures were identified and implemented by PECO as a result of this CAR process to correct and improve the supplier audit process. PECO concluded that the corrective measures taken, combined with more rigorous management reviews of work activities, would eliminate all of the causal factors identified.

c. Conclusions Based upon the self initiated CAR supplier audit measures in 1993, the fact that the 1993 PECO audit of AIT was identified as a problem audit that required either additional actions or compensatory measures to be taken, and the actions taken by 17 I

PECO against AIT in 1996 addressing the concerns identified during the ASME survey and the NRC inspection at AIT, the inspectors concluded that PECO initiated appropriate actions to address the supplier audit issues at AIT.

E7.5 Review of Recent 10 CFR Part 21 Activities

a. jnanarlinn._Srnna l

The inspectors reviewed PECO's recent 10 CFR Part 21 activities as related to l

ecent information which B&G Manufacturing Company (B&G) had provided PECO concerning potentially defective dnadequately heat treated) 9tade 87 fasteners.

The fasteners had been originsky supplied to PECO by Cardinal Industrial Products (CIP) prior to B&G's purchase of CIP in July of 1995. After the purchase of CIP, B&G had formed B&G Cardinal and later renamed the company ACCUTECH.

b. Obscrsations and Findinas B&G had informed PECO in a September 12,1995, letter that CIP had provided PECO with 1-1/2" B7 fasteners from a lot which had subsequently been identified to contain defective product. PECO had initiated two Action Requests A09965059 (Limerick) and A0965060 (Peach Bottom) which recommended the removal of the l fasteners from available stock. PECO determined that Limerick had not received any of the material. PECO determined that the fasteners had been issued from the Peach Bottom warehouse, had been used in eight work orders, and installed in . ,

several applications. PECO had discussed the properties of the inadequately heat treated fasteners with ACCUTECH, performed and engineering review of the fastener application, and determined that the fasteners were acceptable for the applications.

B&G had informed PECO in a September 29,1995, letter that Cardinal !ndustrial Products (CIP) had provided PECO with 3" B7 fasteners from a lot which had subsequently been identified to contain defective product. PECO had initiated two Action Requests A09965045 (Limerick) and A0965046 (Peach Bottom) which -

recommended the removal of the fasteners from available stock, PECO determined that Limerick had not received any of the material. PECO determirmd that the fasteners had been issued from the Peach Bottom warehouse and had been installed in a pipe support application. PECO had discussed the properties of the inadequately heat treated fasteners with ACCUTECH, performed and engineering review of the fastener application, and determined that the fasteners were acceptable for the application.

Atwood and Morrill Co. Inc.(A&M), informed PECO in a April 12,1996, letter that potentially defective B7 fasteners, manufactured by CIP, had been used in valves supplied by A&M to PECO. A&M indicated that it did not believe that the fasteners would cause a substantial safety hazard but recommended that the fasteners be replaced at a convenient time. PECO had initiated two Action Requests A1023118 iLimerick) and A1023119 (Peach Bottom). PECO determined that Peach Bottom had not received any of the material. PECO determined that the valves provided by A&M had been installed in a service water application at Limerick.

18

Nonconformance Report (NCI ) LG 96-01975 was initiated, which documented the PECO operability determination that the valves were operat's as is but that the fasteners should be replaced during a system outage. The NCR documented that all fastener replacements had been accomplished by January of 1997,

c. Conclusions The inspectors concluded that PECO had taken adequate actions to review the supplied information concerning potentially defective fasteners, had performed the appropriate review of the suitability for application, had taken adequate corrective actions, and had adequately documented these actions.

V. Manaaement Meetinas X1 Exit Meeting Summary On March 14,1997, the inspection team conducted an exit meeting with members of the PECO staff and management at PECO Nuclear offices. During the exit meeting the team summarized the inspection findings and observations.

PARTIAL LIST OF PERSONS CONTACTED LICENSEE D. Fetters, Vice President Station bupport T. Niessen, Director, Nuclear Quality Assurance J. Cotton, Director, Engineering H. Birch, Manager, Supply Mai.agement W. Texter, Manager, Corporate Nuclear Quality Assurance T. Baxter, Nuclear Quality Assurance K. Borton, Licensing Section, Station Support Department W. Bradley, Nuclear Quality Assurance W. Strickland, Manager, Materials Management C. Kembring, Procurement Supervisor D. Schmidt, Engineer J. Joneja, Engineer HEC G. Cwalina, Chief, Vendor inspection Section, Special Inspection Branch F. Rinaldi, Project Manager J. Bednar, Foreign Assignee 19 APPENDIX A Omar11 tams This report categorizes the it spection findings as unresolved iter,n and inspection follow-up items in accordance with the NRC Inspection Manual, Manual Chapter 0610. An unresolved item (URI) is a matter about which more information is required to determine whether the issue in question is an acceptable item, a deviation, a nonconformance, or a violation. The NRC Region I office willlasue any enforcement action resulting from their review of the identified unresolved items. An inspection follow up item (IFI) is a matter that requires further inspection because of a potential problem, becsase specific licensee or NRC action is pending, or because additional information is needed that was not available at the time of the inspection.

113m Number Findina Titla

.T.rma 50 352:50 353/g6 201-01 URI Failure to meet ASME NCA 3861 and 3862 (Section E7.5 - b.1, b.2, and b.3) i A-1 g * *  %,,

i

[ '*. UNITED STATES j

j f

NUCLEAR REGULATORY COMMISSION

, o WASHINGTON. D C. 20$55 0001 Aucrust 14, 1997 Mr. Jerry Ethridge, Project Manager Tritium Tart,et Qualification Program Battelle Boulevard P.O. Box 999 Richland, Washington 99352

Subject:

NRC INSPECTION REPORT 99900541/97 01

Dear Mr. Ethridge:

On July 18,1997 the U.S. Nuclear Regulatory Commission (NRC) c mpleted an inspection of the Pacific Northwest National Laboratory (PNNL) activities relating to the Tritium Target Qualification Program. The enclosed report presents the resultr of the inspection.

Duitng this inspection, the NRC inspectors found several instances where the implementation of your cuality assurance program failed to meet certain NRC requirements. The specific instances are described in the enclosed report.

However, as described !n the report, the team noted that PNNL too< action to correct the identified deficiencies prior to completion of the inspection. Therefore, a response to this letter is not necessary.

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

Sincerely, s

%%'h' D Stuart A. Richards, Chief SpecialInspection Branch Division of Inspection and Support Programs Office of Nuclear Reactor Regulation Docket No.: 99900541

Enclosure:

Inspection Report 99900541/97-01 i

l

oc:

Max Claus:n County Ex:cutive Office of Commercial Light Water nieigs County Courthouse Reactor Production Decatur, TN 37322 Tritium Project Office U.S. Department of Energy Mr. Michael H. Mobley, Director 1000 Independence Avenue, SW Division of Radiological Health Washington,' DC 20585 3rd Floor, L and C Annex 401 Church Street St: phen M. Sohinki, Director Nashville, TN 372431532 Office of Commercial Light Wales Reactor Production Defense Programs U.S. Department of Energy 1000 Independence Avenue, SW W:shington, DC 20585~

DP-60 Records Management Office of Commercial Light Water s Reactor Production l Tritium Project Office U.S. Department of Energy 1000 Independence Avenue, SW t -Washington, DC 20585 Mr. Oliver D. Kingsley, Jr.

Pr:sident, TVA Nuclear and -

Chief Nuclear Officer Tcnnessee Valley Authority -

6A Lookout Place 1101 Market Street Chattanooga, Tennessee 37402 2801 Mr. Paul L. Pace, Manager Licensing and Industry Affairs Watts Bar Nuclear Plant Tcnnessee Valley Authority P.O. Box 2000 Spring City, Tennessee 37381 R:gional Administrator U.S. Nuclear Regulatory Commission R:gion ll 61 Forsyth Street, SW, Suite 23185 Atlanta, GA 30303-3415 County Executive Rhea County Courthouse Dayton, TN 37321

U.S. NUCLEAR REGULATORY COMMISSION '

i OFFICE OF NUCLEAR REACTOR REGULATION Report No: 99900541/97-01 Project 697 Organization: Pacific Northwest National Laboratory Richland, Washington

Contact:

Jerry Ethridge, Project Manager Tritium Target Qualification Project (509) 372 4991 Nuclear Activity: Fabrication of the Tritium Producing Burnable Absorber Rods for the Tritium Target Qualification Project, Lead Test Assemblies Dates: April 29 - May_2,1997 July 711,1997 July 1418,1997 Inspectors: Steven M, Matthews, Team Leader, VIS/PSlB/ DISP Robert M. Latta Kenneth C. Heck Approved by: Gregory C. Cwalina, Chief Vendor Inspection Section Special Inspection Branch Division of Inspection and Support Programs Robert A. Gramm, Chief Quality Assurance Section Quality Assurance and Maintenance Branch Division of Reactor Controls and Human Factors 1 INSPECTION

SUMMARY

During this inspection, the inspectors reviewed the implementation of selected portions of Pacific Northwest National Laboratory's (PNNL's) quality assurance (QA) progrcm for supplying Tritium Producing Burnable Absorber Rods (TPBARs) for the Tritium Target Qualification Project (TTOP),

Lead Test Assemblies (LTAs). The inspection was focused on the review of PNNL activities related to the design and manufacture of the TPBARs for their subsequent use in Westinghouse designed burnable poison rod assemblies (BPRAs).

The inspection bases were:

e Appendix A, "Geneial Design Criterie for Nuclear Power Plants,"

General Design Criteria (GDC) 10, " Reactor Design," and GDC 12,

" Suppression of Reactor Power Oscillations," to Part 50 of Title 10 of the Code of FederalRegulat/ons (10 CFR Part 50) l e Appendix B, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50 e 10 CFR Part 21, " Reporting of Defects and Noncompliance" e Tennessee Valley Authority (TVA) Nuclear Quality Assurance (NOA)

Plan (TVA NOA PLNB9 A), Revision 6, dated August 31,1995.

e PNNL's TTOP, " Quality Assurance Plan," ETD 003, Revision 4, dated July 1997 During this inspection, several nonconforrnances to NRC requirements were identified. However, PNNL's corrective actions, taken prior to the end of the inspection, resulted in the team closing the nonconformances. No issues remain open at the time of this writing.

2 STATUS OF PREVIOUS INSPECTION FINDINGS This was the first NRC inspection of PNNL activities related to the TPBARs.

3 FINDINGS FROM THIS INSPECTION 3.1 Quality _Ernnram During the weeks of April 29 through May 2,1997, and July 7 through 11, 2-i -

1997, the team evaluated the acceptability of the QA provisions established to control the design and fabrication of the TPBAR LTAs. This evaluation involved the review of TTOP project documents that described the quality and technical requirements imposed on PNNL by the host utility (TVA) for these safety related compcnents. The team also examined the adequacy and implementation of PNNL's quality assurance plan described in EDT-OO3,

" Tritium Target Qualification Project, Quality Assurance Plan," Revision 3, dated July 1997, Sections 3.1.1 through 3.1.6 provide the team's assessment of controls consistent with TPBAR ::omponent safety classification; QA program adequacy; and QA program implementation in the areas of audits, corrective actions, training, and design control.

3.1.1 Rafaty daulficatinn On December 4,1996, the Department of Energy (DOE) submitted for the staff's preliminary review PNNL topical report " Topical Report on the Evaluation of Tritium Producing Burnable Absorber Rod Lead Test Assembly" (PNNL-11419/UC-731), dated November 1996. The purpose of this report was to provide technicalinformation related to the anticipated irradiation of TPBARs in a commerciallight water reactor, in particular, the report provided a description of the TPBAR design and fabrication requirements, as well as general quality provisions and an evaluation of the safety issues associated with the irradiation of these assemblies in a commercial light wMsr reactor.

Based on NRC's review of the report, a request for additionalinformation (RAI) was forwarded to DOE on January 3,1997. DOE responses to the RAI were provided in letters dated January 21 and February 14,1997. Both responses asserted that the TPBARs did not perform a safety-related function and were, therefore, considered to be non-safety related. The topical report, however, indicated that PNNL would " voluntarily" comply with 10 CFR Part 21 provisions and would apply the PNNL QA program to those items which were considered to meet the requirements of 10 CFR 50, Appendix B. By letter dated Februgry 13,1997, the staff conveyed to DOE its position that the TPBARs were part of a basic component and that, as such, were subject to compliance with the provisions of 10 CFR Part 21 and the quality assurance requirements of 10 CFR Part 50, Appendix B.

In response to the staff's position regarding the safety classification of the TPBARs, PNNL forwarded a revised response to the staff's RAI on March 7, 1997, acknowledging that the design and fabrication of TPBARs would be accomplished under a quality assurance program that complies with the requirements of 10 CFR 50, Appendix B. However, the initialinspection of

l PNNL, conducted from April 29 through May 2,1997, found no apparent connection between the safety classifications described in TVA's Nuclear Quality Assurance Plan, TVA NOA PLN89 A, and PNNLs "importance factors" described in procedure TTOP 1-046, Revision O.

Subsequent to a June 4,1997, public meeting between the NRC and TVA, the staff provided further empilfication on the specific safety function of the TPBARs in a letter to Mr. O.D. Kingsley (TVA) from Mr. F.J. Hebdon dated, June 14,1997. This letter underscored the NRC's position that the fuel and control rod assemblies wers, required to be considered basic components subject to 10 CFR Part 21 that, by definition, were required to be designed and manufactured under a quality assurance program that complied with the requirements of 10 CFR Part 50,-Appendix B, and that parts thereof (e.g., I burnable poison rods and TPBARs) were regarded similarly because of their safety function. The letter further stated that the NRC has always considered burnable poison rods in their entirety to be safety related and that, as such, this position includes the TPBARs in their entiraty (and plugs, getter, cladding, plenum spring, etc.).

a. Inanactinn Reana To evaluate the acceptability of PNNL's safety classification process for TPBAR components, the team reviewed PNNL's controlling procedure TTOP-1-046," Tritium Target Qualification Project, TPBAR Component Characteristics and Related importance Factors," Revision l 3,' dated July,1997. The following paragraphs summarize the resuits of this review.

b. Ohmarvatinns and Findinn.

As determined by the team, procedure TTOP-1-046, Revision 3 had been revised to comply with the NFCs position that the TPBAR components were safety related; TPBAR components were listed with corresponding safety

- functions, and controlling critical characteristics. - Specifically, the TPBAR critical characteristics were defined as those important design, material and performance characteristics necessary to provide reasonable assurance that the item will perform its intended safety function. Table 1 of TTOP-1-046 designates those TPBAR components and critical characteristics, as either Category A or B.

As defined in TTOP-1046, Category A characteristics are those that could affect the ability of the lead test assembles (LTAs) to perform their safety function of maintaining the core in a safe condition. Category B 99-

l characteristics are defined as those that could (1) significantly affect the mechanical integrity of the TP8AR, or (2) result in incremental tritium releases and either onsite or offsite d.oses, or (3) result in localized core power peaking. During the review of TTOP-1-046, the team noted that the designated inspection criteria for Category A and B components appeared to be consistent with their relative importance to safety.

C. CODch neinn Based on subsequent reviews related to this area, the team determined that procedure TTOP 1-046 provided an adequate basis for controlling the design, procurement, fabrication, assembly and handling of the TPBAR LTAs and that appropriate provisions for the component safety classification had been implemented.

3.1.2 ouality Accuranca Plan PNNL's project quality assurance program is described in procedure ETD-003, " Tritium Target Qualification Project, Quality Assurance Plan,"

Revision 3. This TTOP project QA plan encompasses all quality activities related to the TP8ARs, including design, procurement, process development, fabrication, inspection, testing, verification and assessment.

a. Innnactinn Renna The team evaluated the adequacy of PNNL's quality assurance program with respect Appendix B to 10 CFR Part 50 requirements,

b. Ohmarvations and Findinne NRC Regulatory Guide 1.28, " Quality Assurance Program Requirements (Design and Construction)," Revision 3, dated August,1985 conditionally endorses ANSI /ASME NOA-1-1983 Edition, " Quality Assurance Program Requirements for Nuclear Facilities," as an adequate basis for complying with the quality assurance requirements of Appendix B to 10 CFR 50. The additional conditions that are imposed by the Reg Guide 1.28, Revision 3, involve qualification of inspection and test personnel, quality assurance records, and audits.

The team reviewed procedure ETD 003, which is based on ANSI /ASME NOA-1-1989, and the referenced documents which implement the TTOP project QA program. The team also examined PNNL's QA procedures with respect to the conditions imposed by Reg Guide 1.28. Additionally, the

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team performed a comparison of the programmatic requirements contained in ANSI /ASME NOA 1-1983 versus those of ANSI /ASME NOA 1 1989 in order to determine if PNNL's quality assurance program appropriately addressed the Regulatory Positions contained in Reg Guide 1.28.

C. Ennilumlnns Based on the team's review of ETD 003, Rev 3, and the implementing QA program procedures, it was determined that PNNL's TTOP project QA program adequately addressed the requirements of 10 CFR Part 21 and Appendix B of 10 CFR Part 50.

3.1.3 Internal Audit Prnnrarn l

a. Inspectinn hana The team reviewed selected internal assessments and surveillances of PNNL activities controlled by the quality assurance program,

b. Obsatvatinns and Findinne l Criterion XVill, " Audits," of Appendix B to 10 CFR 50 requires that a -

comprehensive system of planned and periodic audits be carried out to verify compliance with all aspects of the quality assurance orogram and to determine the effectiveness of the program. Contrary to this requirement, the team determined that not all aspects of the quality assura' ice program had been audited. Although, this deficiency had been previously identified during an internal PNNL assessment conducted in May 1996, the team concluded that adequate corrective actions had not been implemented in that internal auditr, had not been performed during fiscal year 1997 and none had been scheduled until after project completion.

Subsequent to the team's identification of this issue, PNNL promptly initiated corrective action report (CAR)97-010. This CAR addressed this deficiency and, as documented by CAR 97-010, PNNL attributed the root cause to its reliance on internal assessmants, surveillances, and on external audits conducted by Westinghouse and TVA to sufficiently evaluate QA program elements.

The team reviewed PNNL's corrective actions taken to resolve this deficiency and to prevent recurrence. This review included evaluating the following documents.

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l (1) Section 18 " Audits" of the QA Plan (ETD 003) has been revised (Revision 4, July 1997) to include:

- Management participation in audit scheduling, review of audit findings, and corrective actions taken to resolve them.

l

- Responsibility for scheduling annual audits and assuring that they are performed and reported to management has been j assigned to the project lead quality engineer.

(2) Internal audit procedure TTOP 7 048 has been issued which describes the planning, schedaling, preparing, performing, and reporting of internal audits. The method for reporting audit identified deficiencies, follow up action, and re audit of deficient areas is also described in this procedure.

(3) A fiscal year annual audit schedule has been issued, which assures that all aspects of the quality assurance r,rogram are audited.

Additionally, audits of all activities, with :ne exception of Organization (Criterion 1), QA Records (Criterion XVil), and Audits (Criterion XVill) will be conducted prior to shipment of the TPBARs to the Watts Bar nuclear plant. Audits of the three excepted areas are scheduled for August 1997.

(4) An annual audit schedule for fiscal year 1998 has also been issued.

(5) An evaluation of the impact which the lack of formal audits may have had on the project has been performed. As determined by the team, this evaluation takes into consideration the conduct of internal assessments, surveillances, and external audits performed in fiscal year 1997. The evaluation concluded that there were no direct adverse impacts on the project as a result of not having performed internal audits because the subject areas had been alternatively evaluated,

c. canclocinnt Based on review of the above documentation, the team concluded that appropriate corrective actions had been implemented in response to this nonconformance. The team, therefore, closed this nonconformance.

7 i

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I f

3.1.4 t'attactium Actinn Ptna'Am e, inannetian Renna in order to evaluate the adequacy of PNNL's corrective action program, the team reviewed the status of octions taken to correct identified deficiencies, in particular the team examined deficiencies that had been identified during en internal assessment conducted on May 19,1996, deficiencies identified )

by an external audit conducted by Westinghouse on November 18 21, 1998, with a follow s s audit conducted on June 2 5,1997, and otl.or deficiencies identifier :)y an external audit conducted by YVA on March 24-27,1997.

b. Oh=arvattana and Fintilaan The internal assessment report dated May 14,1996, indic9ted that PNNL's corrective action programs were not being used because the tracking system employed by the OA department was cumbersome.

The Westinghouse audit conducted in November 1996 found that identified problems were being tracked by a TTOP project specific action tracking system, but that the tracking system was not described by the OA program.

Additionally, the Westinghouse follow up audit determined th::t the corrective actions taken in response to this audit finding had not been effective.

The TVA audit which was performed in March 1997 also found that the corrective action program was not being effectively implemented. l Nonconformance reports and deviation reports were not being used. "

Consequently, trending, root carse, and extent of condition evaluations were not being performed.

The team reviewed the actions that had been taken to correct the identified conditions and to implement an effective corrective action program. As a result of this review, it was determined that the following project procedures had been issued:

TTOP 7-045, " Corrective Procedure," Revision 0, dated July 1997.

TTOP 7 037, " Corrective Action Processing Procedure for Deficiency Reports and Corrective Action Reports for the Tritium Target Qualification Project," F.evision 2, dated July 1997.

-8

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TTOP 7 040, "Nonconformance Reports

• Revision 0, dated May 1997.

The team reviewed a recently developed TTOP project corrective action tracking system printnut. Based on the review of this printout the team found that the actions responsive to TVA's audit finding on the ct'rective action program were being tracked under item number CAR 97 001.

However, the status of TVA audit findings at the time of the inspection remained open, pending verification activities by TVA, as documented in a letter from the TVA Project Manager to the TTOP Project Manager, dated July 3,1997.

The team reviewed the corr]ctive actions procedures identified above,  !

which had been revised or issued after the TVA audit. These procedures i were determined to contain appropriate requirements, such as provisions for trending, root cause determination and extent of condition evaluation. The team also determined that the procedures prescribed appropriate requirements and responsibilities for identifying, documenting, tracking, evaluating and correcting deviations from established quality assurance requirements and program controls.

Additionally, the team reviewed the corrective action tracking status report and determlhed that deficiencies were being reported for all active project tasks, including material suppliers, des:gn, and fabrication. Based on an examination of the issue dates for reported deficiencies, the team concluded that organizations performing active project tasks were using the system effectively.

c. Cnnelmelnna Based on evaluation of the TTOP corrective actioa program, the team determined that it was generally acceptable. However, the team noted that PNNL was not taking full advantage of the trending program, since a ? rend analysis had not been performed.

3.1.5 Training

a. InenactinnSenne In order to determine the adequacy of the TTOP tra;;dng program, the team reviewed the governing procedure TTOP 7 011. " Training and Qualification ,

Plan for the TTOP 7 011 Project," Revision 1, dated February 1997.

~

+

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

Additionally, the team reviewed a surveillance conducted on training for fabrication personnel and examined selected training records,

b. ObsarntionLand Findings A Westinghouse audit conducted in November 1996 reported weaknesses in maintaining training records for project personnel. To strengthen this area, responsibility for training had been transferred to the TTOP project office.

To verify the accessibility and retrievability of training records, the team requested the training records following issuance of TPOP 7 037 on June 25,1997. The project office provided a listing of personnel and dates on '

which training had been completed. Based on a comparison of this list with '

a listing of 1 TOP personnes for June 1997, the team determined that training records were being adequately maintained.  ;

The team examined surveillance report SR 2 05, which reviewed the training i requirements for personnelin the fabrication facility. The surveillance report found that fabrication personnel had met all training requirements and were qualified to perform associated fabrication activities. The team also noted, during their inspection of the fabilcation facility, that a matrix of training requirements and the current training status for fabrication personnel was posted prominently at the entrance,

c. f' ant lumlnn Team inspection results confirmed that fabrication personnel satisfied the training requirements for job performance ar,d that appropriate training records were maintained and retrievable.

3.1.6 Tank 1 Da=Igo

a. In198r tinn Kr nna in order to confirm the adequacy of methods used by PNNL TVA, and Westinghouse to exchange design information, the team reviewed current Design Interface Agreements, which were determined to appropriately designate single points of contacts for information used to transmit and develop fermal design outputs, a

4 10-

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b. Observatinn_t and Findings The team reviewed the design verification process with the Task 1 Manager and a TVA project representative. Additionally, the team examined pertinent audits and assessments of design activities that had been conducted internally by PNNL (#97 01) and externally by Westinghouse and TVA. These audit and assessment reports were reviewed for content relative to specific findings and to determine the adequacy of the corrective actions related to design activities.

Task 1 procedures reviewed inuiuded:

TTOP-1017, " Design Analysis /Colculation and Associated independent Review", Revision 1, dated June 1997 l -

TTOP 1019, " Design Change Control", Revision 0, dated January 1997 TTOP 1021, " Design Interface Controls Process", Revision 1 dated June 1997 TTOP 1022, " Design Requirements", Revision 1, dated June 1997 TTOP 1058, " Design Change impacts on Technical and Functional Requirements", Revision 0, dated March 1997 PNNL TTOP 1580, " Functional Requirements for the TPBAR",

Revision 2, dated June 1997 The team also examined TVA's letter to PNNL concerning, " Verification of TVA Plant Specific Design Information," dated May 27,1997, and TVA's letter to PNNL concerning, " Westinghouse Changes from Cycle 2 Analysis,"

dated June 19,1997.

Additionally, the team reviewed selected calculations related to the design and procurement of TPBAR LTA components, associated with the TPBAR cladding and end plugs in order to confirm the appropriate implementation of a

design control processes.

~0 The team interviewed representatives from PNNL's Task 1 design organization in order to galn insights into the TPBAR design verification process. As a result of these discussions, it was ascertained that all PNNL design inputs and engineering documents have been reviewed for approval l

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by a Design Review Board (DRB) consisting of representatives from PNNL, TVA, and Westinghouse. The DRB review process, which was completed at the time of the inspection, concluded that an adequate design basis had '

been established for the TPBAR LTAs to be installed in the Cycle 2 core of the Watts Bor Unit 1 nuclear plant.

As determined from interviews, a design report was being prepared and was  ;

scheduled to be reviewed for adequacy by a design review team prior to the Cycle 2 re:oad. Subsequent to the approval of the design report, any future changes will be reviewed for adequacy and impact on the safety and operation of the Cycle 2 core by TVA.

The team reviewed the current status of the observations and findings that resulted from internal assessments and external audits of the deign control program. As determined by the team, all corrective actior.s had been '

appropriately tracked and no significant design related corrective actions-remained open.

The team examined eleven calculation files selected from a list of engineering calculations related to the design and fabrication of TPBAR cladding and and plugs. As a result of this review effort,-it was determined that all design input information had been properly verified and the team did not identify any deviations from either the administrative design requirements or the technical / functional requirements.-

C. Ennel"*lant Based on the team's reviews of design control documents, audits and ,

assessments, and Interviews with engineering management personnel, it was determined that an adequate design control process had been '

established for the design of the TPBAR LTAs.

3.2 Prneuramant Activittaa a.: Inanactinn Renna i The team evaluated procurement activities for selected critical component parts of the TPBAR lead test assemblies (LTAs) to determine whether applicable regulations were imposed, material specifications were met, sind procedures followed.

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b. Ohsatvations and Andings PNNL's procurament process is defined in Section 4.0, ' Procurement," of its Commercial Light Water Heactor (CLWR), Tritium Target Qualification Project (TTOP), Quality Assurance (QA) Plan, documented in ETD 003. To evaluate the acceptability of a component part, the team used PNNL's critical characteristics as defined in procedure TTOP 1046, Revision 3.

In order to evaluate the procurement activities, the team reviewed PNNL's acquisition of the stainless steel material for the TPBAR cladding tubes and end plugs, and the Lithium Aluminate (LIAIO,) pellets. The following paragraphe summarize the results of this review, b.1 316 Stainless Steel Material PNNL procured the stainless steel bar stock material from Westinghouse Hanford, that had originally procured the material for use in DOE's Fast Flux Test Facility (FFTF). The material was procured to material specification TTOP-1-003, " Specification for 316 Stainless Steel Seamless Cladding Tubes," Revision 1, dated May 1996. TTOP 1003 complied with ASTM Standard A 77188, " Standard Specification for Austenitic Stainless Steel Tubing for Breeder Reactor Core Components," that, according to PNNL, reflects the fabrication and technical data gained over two decades of cladding development and procurement for the FFTF.

The team determined that both ASTM A 771 and TiOP 1003 require double vacuum melted feed stock, and chemistry and inclusion limits on the produr:t. The products to be produced from the 316 bar stock procured from Westinghouse Hanferd for PNNL fabrication was TPBAR clad tubing and end caps.

PNNL verified the adequacy of the stainless steel materialir accordance with TTOP 2 001, " Material Verification Procedure for the Tritium Target Qualification Project," Revision 0, dated May 1996. However, on the basis of the team's review of the certification of the starting bar stock material, the team determined that the Material Reverification Record, signed and certified by PNNL, was not complete and that procedure TTOP 2 001 failed to adequately establish requirements for the completion and certification of that document. The team identified this issue as a nonconformance.

Subsequently, PNNL issued TTOP Deficiency Rcrort (DR) 07 066, dated May 1,1997. The team reviewed PNNL's corrective actions that included revising procedure TTOP 2 001 to address the weaknesses identified and 1

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correctly completing the Material Reverification Record. The team found these corrective actions adequate and closed the nonconformance.

• During this review the team also detetrnined that TTOP 1046 failed to adequately describe the actual sample size PNNL used to confirm the chemistry and inclusions of the stain'ers steel cladding bar stock or the actual characteristics verified during PNNL's reverification process. In addition the team determined that material specification TTOP 1004,

" Specification for Target Rod End Cap Bar Stock Material," failed to agree with the importance factor sampling plans specified in TTOP 1046 and that TTOP 1046 did not adequately address the use of the ASTM standard sampling frequencies, where applicable, that PNNL actually used during its i reverification plan.

Subsequently, PNNL issued DR 97 068, dated May 7,1997. That DR defined the deficiency as follows:

The 316 stainless steel bar stock dedication activity which was -._

performed in the fall of 1996 is inconsistent with the sampling requirements specified in TTOP 1046. In addition, both the sampling requirements and nomenclature used to describe characteristics in TTOP 1004, Revision 1, " Specification for Target Rod End Cap Bar Stock Material," are inconsistent with '

the sampling requirements and nomenclature used in TTOP 1-046.

On the generic basis of this nonconformance, PNNL required all of its material specifications to be compared to TTOP 1046 to assure agreement in all cases.

The team reviewed the effectiveness of this corrective action and determiried that the actions ta'in by PNNL were adequate to address the specific item issue and the generic implications "f this issue. The team therefore, closed this nonconformarce, b.2 Lithium Aluminate Pellets PNNL procured the Lithium Aluminate pellets from ICI Advanced Ceramics.

The pellets were procured in accordance with TTOP 1-009, " Specification for Enriched, Annular LiAIO, Pellets," Revision 3, dated April 1997. That specification provided that the seller shall be capable of showing with 95%

confidence, at least 95%,90%, and 75% of the pellets in a lot meet the specifications for the characteristics defined in Table 3, " Classification of

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Pellet Characteristics for Sampling Plans." That table listed the confidence / inspection level (95:95, 95:90, or 95:75) for each characteristic of the pellet to be verified, in addition, the specification required ICI to subm!t to PNNL the sampling plan for review and approval prior to use.

In a memorandum to PNNL dated May 1,1997,=lCl specified the sample sizes associated with the inspection levels as follows:

inanar tInn t avalet Ramnia RIsa!

95:95 15 pellets 95:90 -7 pellets 95:75 3 pellets -

According to ICl, the sample sizes were verified by the OC curves foiand in ANSl/ASOC Z1.91993, " Sampling Procedures and Tables for Inspection by j Variables for Percent Nonconforming." The team determined that ASOC Z1.9 sample sizes are based on the assumption that the data represents a i normal distribution. Therefore, in order for PNNL to use this standard to determine sampling sizes, PNNL would have to show a documentation of ICI's past performance of complying with the critical characteristics as the basis to support PNNL's assumption that the data represents a normal distribution.

On the basis of its review of PNNL's procurement, receiving inspection, and acceptance of the Lithium Aluminate pellets, the team identified the following concerns that constitute a nonconformance: .

(a) PNNL failed to document its basis for the assumption that ICl's data represents a normal distribution and therefore the appropriateness of using the small sample sizes to verify critical characteristics of the l

pellets.

(b) PNNL failed to include these sampling sl:es in the Inspection / Test instructions (ITls) used by the QC inspectors to verify the adequacy and acceptance of the pellets for use in the TPBARs.

PNNL responded to the team's concerns by taking the following corrective actions: .

(a) in a memorandum dated July 18,1997, PNNL adequately '

documented its previous procurements from ICI and established its basis for using the small sample sizes to accept the pellets.

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l (b) The Itis for receipt inspection for all pellets for LTA lots received were revised to address the sampling matrix and confidence / inspection levels specified in TTOP 1009.

The team reviewed the effectiveness of these corrective actions and determined that PNNL's actions taken were adeo.uste to address the issues. ]

The team, therefore, closed this nonconformance. '

C. f'anclutlant The team identified concerns with the procurements of the stainless steel material for the TPBAR cladding tubes and end plugs, and the Lithium Aluminate (LIAIO3 ) pellets. These concerns constituted nonconformances.

However, PNNL responded with corrective actions that adequatc!/

addressed the team's concerns and resulted in the team determining that the nonconformances were closed.

3.3 IRBAR Fahrleatinn Actlwitles

8. Inanactinn Renna During this portion of the inspection, the team evaluated the material control for the TPBAR fabrication, hand;ing and storage of cladding tubes, welding of the upper end caps, and pencil assembly activities for the TPBAR LTAs to determine whether adequate quality assurance provisions were established i and procedures followed. PNNL's fabrication process and quellty plan is defined in TTOP 2 013. " Manufacturing and Quality Plan for Tritium Producing Burnable Absorber Rods for the Tennessee Valley Authority Lead Test Assemblies," Revision 1, dated June 1997, and in TTOP 2 014,

" Tritium Target Rod Fabricating Process Plan," Revision 1, dated June 1997.

b. Ohaarvatinn and Findinne To evaluate the acceptability of the fabrication process, the team initially reviewed the established procedure and QA plan for the process and then observed PNNL's activities in performing the process activity. The following sections describe the results of this review.

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b.1 Material Control PNNL's material control prot,ess is described in the following:

TTOP 2101, " General Receiving for the Tritium Target Fabrication Facility (TTFF)," Revision 0, dated February 1997 TTOP 2102, " Handling of Miscellaneous Components." Revision 0, dated February 1997 TTOP 2105, " Control and inventory for Received Storage,"

Revision 0, dated January 1997 TTOP 2106, " Control and inventory for Accepted Storage," i Revision 0, dated January 1997 TTOP 2-109, " Control and i.wentory for Rejected Storage,"

Revision 0, Ated January 1997 The team reviewed the established measures for niatorial control contained '

in TTOP procedures and interim change notice (ICN) 2-101 01, completed

-July 8,1997, and determined that the measures established were adequate.

The team reviewed the Component inventory Ledger and the Transaction

- Log Sheet for each of the material control cages (received, accepted, and rejected storage cages) and found that adequate controls were in place and that all entries reviewed matched with existing inventory. The team also '

verified that the colored tags found on many items in the TTFF were appropriately documented and controlled in accordance with estab!!shed  ;

procedures. t b.2 Cladding Tubes PNNL's process for handling cladding tubes were documented in TTOP 2-103, " Handling of Empty Cladding Tubes," Revision 0, dated February 1997, TTOP 2104, " Handling of Loaded Cladding Tubes," Revision 0, dated February 1997, and TTOP 2 211, "incpection/rest Instructions for inspecting Cladding Tubes," Revision 0, dated March 1997. The team reviewed the established measures for the control of cladding tubes in TTOP 2103 and 104 and determined that the measures established were 1 - adequate.

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The team reviewed the Transaction Log Sheet and tags for the empty cladding found in the TTFF and determined that adequate controls were in place and that all documents and tags reviewed matched with existing recoids, b.3 Welding i PNNL's process for welding qualification, performance, and inspection were documented in the following:

TTOP 2 018, " Qualification Plan for End Plug Welding," Revision 1, dated June 1997 TTOP 2 024, " Radiography inspection Procedure Qualification Plan,"

Revision 0, dated June 1997 TTOP 2117, " Top a.:d Bottom End Plug Welding," Revision 0, dated June 1997 TTOP 2 303, " Weld Visual Inspection," Revision 0, dated May 1997 i

- TTOP 2 310 " Radiography inspection," Revision 0, dated June 1997 The team reviewed the established measures for welding qualification, performance, and inspection and determined that the measures established were adequate.

l The team witnessed the real time radiography (RTR) of the 12 end cap welds used to qualify the end cap welding process and determined that tae process was adequately controlled and performed in accordance with established procedures. The team also review 9d the radiographs taken of the 12 end cap welds which were used by Pl4NL as additionalinformation regarding the adequacy of the welding process. The radiographs were produced by X raying the welds in accordance with Westinghouse Hanford Nondestructive Examination Procedure Manual WHC CM 4 38, Section NDT.

RT 4000, " General Radiographic Examination Procedure," Revision 2, dated January 15,1994. Specifically, Appendix C, " Capsule, Fuel, and Absorber Pin Radiography," Revision 1, dated July 16,1994, was followed to achieve the radiog aphs of the 12 end cap welds using the beam filtered tangential ,

radiography technique. The tcsm found the radiographs and the processes l used to produce them to be adequate. i

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On July 17,1997, the team reviewed the ccmpleted welding qualification report for the qualification of the end plug welding qualification. That report was documented in TTOP 2 023, " Qualification Report for End Plug Welding," Revision 0, dated July 1997. The team found the report to be well documented and very thorough. Overall, the team found PNNL's welding qualification package to be excellent.

On July 18,1997, the team witnessed actual production welding of the qcond set of c top and cap welds. The activity was well controlled and the welding processes were performed in accordance with the established procedures.

b.4 Pencil Assembly PNNL's process for performing the pencil assemblies were documented in TTOP 2125, " Pencil Assembly Loading," Revision 0, dated May 1997, and TTOP 2 225, " Pencil Assembly inspection," Revision 0, dated May 1997.

The team reviewed the established measures for the control of pencil assembly anti inspection and determined that iie measures established were adequate.

The team witnessed the pencil assembly process for several pencils and determined that adequate controls were in place and that all documents and tags reviewed matched with existing records. The team noted and commented on the extreme care and precision exhibited by PNNL's staff performing the pencil assembly and inspection activities.

c. Onnelitelnrts For the fabrication activities reviewed, the team found that PNNL performed those activates with great care and attention to detail and that the personnelinvolved followed established procedures. No adverse findings were identified by the team.

4 ENTRANCE AND EXIT MEETINGS in the entrance meeting on April 28,1997, the NRC inspectors discussed the scope of the inspection, outlined the areas to be inspected, and established interfaces with PNNL rnanagement, in the exit meetings, on July 11 and July 18, 1997, the inspectors discussed their findings and concerns.

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i PARTIAL LIST OF PERSONS CONTACTED PNNL i J. Ethridge G. Sorensen i R. Latorre l S. Engl!sh l

-D. Sensor C. Painter L. Erickson S. Bales D.' Rittenhouse C. Thornhill R. Guenther IVA J. Chardos DOE M. Clausen l

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greaCICp UNITED STATES l/g (1.3 i j

! t NUCLEAR REGULATORY COMMISSION WACWNGTON, D.C. M564:m sy*****/

August 10, 190?

i Mr. Lew Goetz l ' President and CEO i SOR, Inc.

14685 West 105th Street l Lenexa, Kansas 60215-5904

SUBJECT:

NRC INSPECTION REPORT 99900824/97-01 AND NOTICE OF NONCONFORMANCE

Dear Mr. Goetz:

On June 19, 1997, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the SOR, Inc. (50R) facility. The enclosed report presents the results of that inspection.

The inspection was conducted to essess the adequacy of the correctivo actions SOR took to correct manufacturing defects associated with certain safety-related pressure, vecuum, and temperature switches made by SOR and sold to licensees of nuclear plants. The inspector assessed specific attributes of the SOR quality assurance program ar.d reporting of defects under 10 CFR Part 21.

He also assessed licensee monitoring of the quality of SOR switches.

During this inspection, the inspector determined that in 1993 and 1994, SOR applied a heavy coating of epoxy on the insulated lead wires of switches it sold to licensees for safety-related applications. The epoxy hardened the insulation causing it to crack when it was bent during installation. In September 1994, Nebraska Public Power District (for the Cooper Power Station) and Connecticut Yankee Atomic Power Company (for the Haddam Neck Plant) reported cracked insulated switch wires in their plants. SOR issued a Part 21 response and repaired or replaced the affected switches. The NRC inspector concluded that SOR had failed to prescribe procedures or instructions to prevent the epoxy from being applied on the insulation. On this basis, the inspector concluded that 50R's quality assurance program had not met certain NRC requirements imposed upon it by NRC licensees.

This issue is cited in the enclosed Notice of Nonconformance (NON), and the circumstances surrounding it are described in detail in the enclosed report.

In a June 19, 1997 letter, 50R reported steps it took to correct the nonconformance and prevent recurrence. No further response is required.

In addition, the inspector observed that during the January 1995 Nuclear Utilities Procurement Issues Committee audit, lic ' sees did not evaluate SOR's action to cr.trect two manufacturing defects that .>R reported to its

, customers.

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f L. Goetz '

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, A

Stuart A. Richards, Chief Special Inspection Branch Division of Inspection and Support Programs Office of Nuclear Reactor Regulation Docket No. 99900824 Docket No. 99900912 Er. closures: 1. hotice of Nonconformance

2. Inspection Report 99900824/97-01

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NOTICE OF NONCONFORMANCE SOR, Inc. Docket No.: 99900824 Lenex4, KS l

Based on the results of an NRC inspection conducted on June 16 through 19, '

1997, it appears that certain of your activities were not conducted in accordance with NRC requirements.

Criterion V of Appendix 8 to 10 CFR Phrt 50, " Instructions, Procedures, and Drawings," requires, in part, that activities affecting quality shall be prescribed by documented instructions and procedures of a type appropriate to the circumstances and shall include appropriate quantitative or qualitative acceptancecriterlafordeterminingthatimportantactivitieshavebeen satisfactorily accompitsbed.

SOR Nuclear Quality Assurance Nanual 8303-100, Section 5.2, " Instructions, Procedures, and Drawings," dated April 14, 1993, requires, in part, that .

instructions, procedures, and draeings must contain acceptance criteria to ensure compliance with customer q;ality requirements.

Contrary to these requirement., SOR did not prescribe instructions or procedures to ensure that epoxy was not apolied on insulated lead wires of safety-reltled pressure, vacuum, and temperature switches, or that quality inspectors examine the switches properly. As a result, a heavy coating of epoxy was applied on the insulated wires during manufacture. The epoxy hardened the insulation causing it to crack when it was bent. In September 1994,= Nebraska Public Power District (for the Cooper Power Station) and _

Connecticut Yankee Atomic Power Company (for the Haddam Neck Plant) identified cracked insulated lead wires 'of switches installed in their plants. The

,: racked insulation had the potential to reduce the wire insulation resistance or cause_a short to ground. On October 14, 1994 SOR issued a 10 CFR Part 21 to inform the NRC and applicable customers of this defect and pertinent corrective action. SOR replaced ~or repaired the' safety-related switches and'

- took appropriate steps to prevent epoxy from being applied on the insulation of lead wires.

!n a June 19, 1997, letter to the NRC, S0R reported the steps it took (in 1994) to correct the problem and prevent recurrence. Steps comprised alerting customers about the cracked insulation, repairing defective switches installed ,

in plants, preparing instructions and procedures, and implementing new quality control inspections to prevent the use of epoxy on the insulation. No further response is required (99900824/97-01-01).

Dated at Rockville, Maryland

.this loth day of August 1997 Encloture 1 l

.)

.)

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

Report No: 99900824/97-01 Organization: SOR, Inc. (SOR)

Lenexa, Kansas

Contact:

Colbert O. Turney Vice President, Quality Assurance 913/888-2630 Nuclear 1.sdustry Pressure, vacuum, and temperature switches Activity:

Dates: June 16-19, 1997 Inspector: Anil S. Gautam,. Senior Engineer i

Approved by: 3regory C. Cwalina, Chief Vendor Inspection Section Special Inspection Branch Division of Inspection and Support Programs Enclo'sure 2

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1 INSPECTION SUNNARY During this inspection, the NRC inspector assessed the adequacy of the actions

taken by SOR to correct manufacturing defects associated with certain safety-related pressure, vacuum, and temperature switches (hereafter referred to as switches). The defects included (1) cracked insulation of lead wires for switches, (2) leakage of 0-ring seals in switches exposed to radiation and elevated temperatures, and (3) leakage of epoxy seals in switches. The- ,

inspector assessed specific attributes of SOR's quality assurance program and reporting of defects under 10 CFR Part 21, and licensees' monitoring of 50R's control of quality.

The inspection bases were as follows:

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

• 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)

• SOR Nuclear Quality' Assurance Nanual (QAN) 8303-100, Revision 9, dated  ;

April 14, 1993, and associated implementing procedures l l'

During this inspection, the inspector noted one instance in which-SOR failed to conform to NRC requirements imposed upon it by NRC licensees. This nonconformance is discussed in Section 3.1 of this report. In addition, the inspector observed that during the January 1995 Nuclear Utilities Procurement Issues Committee-(NUPIC) audit, licensees did not evaluate SOR's corrective actions regarding certain manufacturing defects repotted by SOR in information notices to customers. Licensees' monitoring of 50R's control of quality is discussad in Section 3.2 of this report, t STATUS OF PREVIOUS INSPECTION FINDINGS Ocen Item 99900912/93-01-06 (Closed)

-During_a June 1993.NRC inspection of National Technical Systems _(NTS), Inc.,

in Acton, Massachusetts, the inspector assessed qualification testing of SOR pressure switches and observed (1) a pressure leak in one sample during a i high-energy-line break (HELB) test, and (2) excessive leakage current in another sample during a dielectric withstand test. The inspector found no-documented evaluation by NTS or SOR of the root cause of the test failure nor pertinent corrective action. The inspector considered this an open item.

Following the NTS inspection, SOR gave the NRC documentation regarding the test failures. On the basis of the documents, the inspector determined that the pressure leak was due to a. leakage path provided by unsealed mounting bracket screws for the microswitch (switching element) mounted in the switch housing. SOR believed that the screws had not been resealed after the microswitch was readjusted during factory calibration. Failure to reseal the screws allowed the switch diaphragm (seal) to be over)ressurized during the test and caused it to leak. SOR stated that since otier switches with the 2

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same type of housings did not suffer a similar test failure, the test failure i was attributable to a random occurrence, not to an inherent weakness in the <

. design. 50R's corrective action consisted of (1). resealing the microswitch mounting screw threads-if the microswitch was readjusted during factory i calibration, and (2) applying a primer to the microswitch bracket screws to i improve the curing of tse-theead_ sealant in stainless steel housings. SOR j -confirmed that applicable safety-related pressure switches installed in plants

!- were not compromised because the corrective measures had been instituted

before production. This issue is closed. '

Regarding excessive leakage current in one sample: the inspector observed that the SOR pressure switch test specimens passed the dielectric withstand test at

- 1500 Vac for 1 minute, except for one sample that experienced about 2 L milliamps (mA)_ of leakage current at 900 Vac. SOR could not find a root cause

and believed that the 2 m4 leakage current was a random' anomaly and not

=

indicative of a common failure mode. 50R's basis was that the other specimens (1) passed the test at 1500.Vac,

! Vde, and (3) had sufficient marginfor (2)service had adequate conditions insulation because resistance the switchesat 500

! were. rated for 250 Vac and typically energized for !?0 Vac or 125 Vdc

applications. The inspector determined that on the basis of information '

provided by SOR, and because leakage current from moisture intrusion would have been higher than 2 mA, the anomaly was satisfactorily addressed. This i

issue is closed.

) 3 INSPECTION FINDINGS AND OTHER COMMENTS 3.1 Ouality Assurance Proaram 4

a. Insnection Scone The inspector examined the adequacy of SOR's Part 21 evaluations, L corrective actions, conformance to procurement documents, and.self-i assessment of performance.

! b, Observations-and Find'Q21 The inspector observed that $0R's QA program was-based on the policies  !

! and criteria of 10 CFR Part 50, Appendix B. .The QA program staff was J

comprised of the Quality Assurance Vice President-(QAVP) and 2 quality i control-(QC) inspectors. The QAVP reported directly to the 50R's i President /

CEO. The QC inspectors were authorized to stop production of

a nonconforming item until the nonconforming conditions were corrected.

' The inspector obse ved that SOR had posted sections of the Federal i Aeofster, dated September 19, 1995, concerning the latest changes to

! 10 CFR Part 21' bet had not posted the complete Part 21 regulation, as is required by 10 CFR 21.6. During the inspection .the QAVP posted copies-4 of the complete regulation in appropriate locations. No'further concerns j were identified.

E i

! 3 f.

-l2 $*

The inspector assessed SOR's Part 21 reports and corrective actions for manufacturing defects associated with SOR se.tches during the past 5 years. Dcfects included cracked lead wire insulation, leaking 0-ring seals, and leaking epoxy seals in the sr 5 conduit seal. The inspector's review is sumarized belov (1) Cracked Insulated Lead Wires Insulated lead wires for the switch enter and exit an epo:y seal in the conduit adapter cf the switch housing. The conduit auapter is potted (sealed) with epoxy to keep moisture from entering the switch housing.

B In September 1994, Nebraska Public Power District (for the Cooper Power Statien) and Connecticut Yankee Atomic Power Company (for the Haddam Neck Plant) notified SOR cf eight defective switches that had cracks in the insulated lead wires. SOR determined that the cracks had been caused by SOR's misapplication of the epoxy on the insulation, subsequent hardening of the insulation, and cracking and tearing of the insulation when it was bent.

On October 14, 1994 SOR sent a 10 CFR Part 21 report to the NRC and customers about the cracking of the lead wire insulation in SOR's nuclear-qualified switches (the NRC also issued event notification 27902 on October 14, 1994, to inform licensees that the switches posed a potential risk of failure of safety-related equipment). Subsequently, approximately 11 licensees returned their switches to SOR for repair.

During this inspection, SOR provided a written response (Attachment 1) to the inspector, dated June 19, 1997. SOR stated that it inadvertently applied the epoxy on the insulation "due to poor workmanship" and that "the condition was undetected because SOR quality inspectors did not notice the coating of the epoxy on the (insulated) wires." SOR also told the inspector that it had not prepared instructions to ensure that epoxy was not applied to the wira insulation. The inspector concluded that SOR's failure to prescribe instructions or procedures to ensure that epoxy was not applied on the insulated lead wires of the switches, or that quality inspecto " examined the switches properly, as required by Criterion V, Instructions Procedures, and Drawings," of Appendix B to 10 CFR Part 50, constituted Nonconformance 99900824/97-01-01.

The inspector observed that, in 1994, SOR revised its work procedures to preclude application of epoxy on insulated wires, and to reject any insulated wire that may have been covered with epoxy. SOR added shrink tubing to the insulated wires where they entered the conduit adapter epoxy seal to protect the wire insulation during shipping and handling. SOR also recommended not exceeding a minimum bend radius for the insulated wires. The inspector determined that SOR's actions to correct the misapplication of epoxy and prevent recurrence were adequate. No further response is required.

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l (2) Leakage of 0 !Gng Seal The vacuum switch included an 0-ring installed in a triangular gland to seal the process air or fluid between the vacuum screw and the piston, and between the vacuum piston and the primary diashragm. In 1992 and 1993, SOR implemented a new program to qualify tie vacuum switches (SOR test report 9058-102), and discovered that the 0-ring ses1 in the vacuum switch was not capable of retaining the required maxirma pressure after exposure to high radiation, high temperature, and hydrostatic pressure greater than 150 psi. SOR concluded that leakage between the vacuum screw and the 0-ring had occurred during testing.

On April 1 19> SOR issued "Information Notice Concerning Vacuum 0-RingSealin50sNuclear-QualifiedVacuumSwitches"toapplicable custon.ers "egardir g potential leakage in vacuum switches, and suggested to c.Womers that all switches be replaced if exposed to pressures greater than 150 psi. The affected switches were those designated by a 54N6, 54TA, 52N6, or 52TA in the first section of the i model number and JJTTX6, JJTTX7, JJTTX13, or JJTTX14 at the end of the model number.

During this ins)ection, on June 19, 1997, SOR provided a written response (Attac1 ment 2) to the inspector. SOR stated that it had not discovered the condition described above earlier "because of inadequate engineering testing and analysis of the vacuum switch." On May 20, 1993 SOR took corrective action to elimir. ate the leak path by (1) welding the vacuum screw to the vacuum piston and (2) replacing the triangular 0-ring seal with a face seal of the same material. The face seal was qualified by analysis (SOR test report 9058-102, Section 14, Appendix 4, Analysis 8923-219) to retain a pressure of 750 psi after exposure to radiation and elevated temperatures. SOR reported that it replaced applicable switches sold to licensees. No further concerns were identified.

(3) Leakage of Conduit Seals The switch lead wires > ass through the outer nipple of the conduit seal cannector, throug1 the epoxy seal potted in the nipple, and through a glass seal which is soldered inside the nipple. In May 1994, during routine testing, SOR discovered leakage of pressure through the conduit epoxy seal of "NQ" switches. Da June 10, 1994, SOR issued "Information Notice Concerning Conduit Seals in SOR Nuclear-Qualified Pressure, Vacuum, and Temperature Switches" to inform the NRC and customers about tne ptential leak. In the notice, SOR stated that the leak could lead to reduced insulation resistance or loss of function of the switch during or after a HELB. SOR suggested to customers that all swit:hes be returned to the SOR fa(tory for inspection if they were subject to HELB conditions during or after an event, or if subject to conditions in which condensate may form inside the conduit, or if subject to any other conditions in which moisture could penetrate the conduit seal.

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During the inspection, on June 19, 1997, SOR provided a written response (Attachment 3) to the inspector. SOR stated that the problem 4 went undetected because (1) the leakage was a random problem, (2) 50R's inspection. steps were not adequate to identify the faulty 1 condition seals, and (3) there was a manufacturing error in the heat cure of the epoxy because manufacturing personnel had not followed  ;

procedures. 50R also determined that the work format was not adequate  !

because it did not require manufacturing personnel to record the actual heat cure temperature and the cure time for each batch.

In June 1994, SOR took measures to prevent recurrence by implementing ]

more stringentresistance an insulation testing ontest all conduit seals, for conduit including seals, (2) a requiring (k1) 100 psi lea test for the completed conduit seal assembly, (3) a housing leak test ,

for the conduit seal after completing all assembly steps and all thermal testing, and (4) test results to be approved by manufacturing and QA personnel for every order of switches. In addition, SOR took measures to record the cure tempe;ature and time for the epoxy to ensure that the correct heat cure was used. This activity is required to be approved by manufacturing and QA personnel for every order of '

conduit seals. No further concerns were identified.

The inspector observed that SOR did not ask customers to identify any chemicals that the switch components would be exposed to during installation or operation to ensure that the switch was not compromised in the performance of its function. Chemicals in the process (e.g.,

ammonia) could degrade switch components (e.g., seals). SOR indicated that they assessed any process chemicals if identified by the licensee.

The inspector assessed 50R's implementation of licensee purchase order requirements in 50R's design documents (50R assembly drawings 8520-264 Revision 2, 8520-506 Revision 1, and 8215-659 Revision 2). No concerns were identified.

The inspector observed that 50R's General Instructions did not address the protection of switch components during handling (e.g., debris l entering the switch housing, damage to lead wires) which could affect the operation of the microswitch. The QAVP added a cautionary statement to the General Instructions.

The inspector assessed SOR's internal audit report 7701-128, revision 4, dated December 30, 1996. . The audit, in part, assessed the results of -

nonconformance reports and correctiv actions. No concerns were identified.

c. Conclusions In general, $0R's QA manual and its implementation were in compliance with the requirements of Appendix B to 10 CFR Part 50, except for the nonconformance described herein. SOR took adequate corrective actions

-and steps to prevent recurrence of identified manufacturing defects.

6

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1

{

3.2 Review of Licensee Monitorina of SOR

a. Insoection Scone The inspector evaluated Itcensee monitoring of 50R's control of quality for-safety-related items purchased by licensees, including Part 21 reports and associated corrective actions,

b. Observations and Findinas In January 1995, NUPIC - represented by Southern California Edison (SCE),

Baltimore G" & Electric Coerany, and Yankee Atomic Electric Company -

audited 50R's QA program. Part of the scope of the audit was to verify whether SOR had established and effectively implemented a QA program in compliance with the requirements of 10 CFR Part 50, Appendix B and other industry standards. Th.i NUPIC audit team identified SOR deficiencies in the areas of (1) control of corrective actions, and (4) control purchased materials.

of measuring and(2) testtest control, equipt.. ant. (3)

NUPIC considered the findings to be " administrative" and believed there was no adverse impact on the quality of SOR's completed products. NUPIC accepted 50R's corrective actions for the above findings and closed the findings on April ll, 1995. NUPIC concluded that 50R's QA program was adequate and that implementation was satisfactory. -

The inspector observed that during its audit, NUPIC reviewed 50R's 10 CFR Part 21 report, dated October 14, 1994, regarding cracking of the insulation of switch lead wires. NUPIC verified SOR's corrective actions by observing in-process assembly of pertinent switches and associated documentation. The inspector noted that durir.g the NUPIC audit, licensees did not evaluate E3R's 1993-1994 corrective actions regarding leakage of 0-ring seals in switches exposed to radiation and elevated temperatures, and leakage of epoxy seals in switches. These manufacturing defects were reported by SOR in information notices to customers (see section 3.lb of this report). After a telephone discussion with SCE's procurement-quality staff, the inspector confirmed that NUPIC had not included these issues in the scope of the audit.

In February 1997. NUPIC, represented by Omaha Public Power District, examined the application of SOR's QA program to all phases of the design and manufacture of SOR switches. NUPIC noted that SOR had issued a 10 CFR 21 report.in 1993 (no details are noted in the NUPlc report).

NUPIC recommended, in part, that 50R should clearly document its methods of verification of critical characteristics, and develop a checklist of specific inspection criteria for items purchased. No findings were identified.

c. Conclusions In general, licensee monitoring of SOR's quality was in accordance with proper criteria, procedures, and checklists. NUPIC did not evaluate-SOR's corrective action for two manufacturing defects reported by SOR to its customers.

7

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3.3 Entrance and Exit Meetinos 1997 At the entrance scope of the inspection, meeting outlined on Junet 16,he are,as to be inspected, andthe NRC inspecto established interactions with SOR management. In the exit meeting on June 19, 1997, the inspector discussed his finding and observations.

4 -PARTIAL LIST OF PERSONNEL CONTACTED 103 Lew Gcetz, President and CEO Colbert Turney, Vice President (VP), Quality Lind Coutts, Coordinator, Nuclear Engineering Joseph Modig. Engineer. Wuclear Engineering Landen Tuggle, Director, Manufacturing Harold Moddy, VP Sales Charisse Smith, VP Finance Tim Ceillesen, Product Manager Richard Johnson, QC Engineer Southern California Edison Jeff Larson, Supervisor, Procurement Quality ITEMS OPENED, CLOSED, AND DISCUSSED Ooened 999008?d/97-01-01 Para 3.1 b NON inadequate instructions and procedures Closed 99900912/93-01-06 Para 3.10 of Open item inadequate information inspection report regarding test anomalics 99900912/93-01 s

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OEB E i.e., w ,osis so.., . t.e,.. x. m eemo us' Te' G13 888 2630

• r.: 913 8e8 0767 ,, , ,

9,0,0,1 Ntr. Anil S. Gautam. NRC

Subject:

10Cl R2l dated October 14.1994

Dear hir. Gautam:

In Jul.v of 1993 SOR began to manufacture pressure. Vacuum, and temperature switches that were qualified by SOR test report 9058102. In September of 1994 SOR was notified of a manufacturing defect by Nebraska Public Power iRef. RGA 2125. ses en defectise unitsi and Connecticut Yankee (Ref. RGA 2117. one defectise unit). In aeJition SOR assembly personnel Fad identified the same defect iRef. NIRR 1479. one defectis e unit t The defect was identified as a crack in the lead wire insulation 'hi3 prompted SOR to issue a 10CFR21 and insestigate the cause of the defect. The cause o' the cracked insulation was a heas) coating of epoxy on the wires outside of the potted area and was due to poor workmanship. This condition went undetected because SOR quality inspectors did not notice the coating of epoxy on the wires.

As noted in the Pan 21 Notification, the following correctis e action was taken in October 19 4:

1. The Work Order famiats for the conduit seals were revised to include specific instructions not to allow epoxy on the wires. In addition. there is an inspection step at the end of the Work Order that instructs the inspector to examine the wire and reject any that have epoxy on the wire. Each of these steps must be signed off on the Woik Order by N1anufacturing and QA personnel for each order of conduit seals.

2. Shrink tubing was added to the lead wires where they exit the conduit seal. The purpose of this tubing is to protect the wire insulation during shipping and handlim.

This step is signed off on the assembly procedure by Alanufacturing and it is reviewed by QA per<onnel for esery switch.

3. The wire manufacturers re mmended minimum bend radius was added is. the SOR General Instructions that are provided to the customer with each switch.

The c.onduit seals are manufactured as a sub assembly in a separate environmentally controlled room. Therefore. there is no danger of epoxy contamination on any other parts of the switches.

Reference Correctise Action Report 0357.

Regards, adsk Colben Turney,y V.P. Quality 0

Q.*

Joseph G. Modig. Engineer

'Af7

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fm l <

MIIQfD G 2 L_ J 14685 W.105th street

• Leness Kansas 6621s-5964 U$A lei 913 868 2630

• Fas 913 688 0767 ,,t ,, l 1

hfr. Anil S. Gautam, NRC l

Subject:

Information Notice of April 1.1993

Dear hit. Gautam:

Prior to April 1.1993 SOR had been manufacturing nuclear qualified vacuur.) switches for approximately 10 years. These switches were qualified by a combination ofiesting and analysis as listed below:

. AETC Test Repon 17344 82N D. Rev.1 e AETC Test Report 18441 83N. Rev. I e AETC Test Repon 17344 82N C, Rev. 3 e AETC Test Repon 18577 83N. Rev.1

• AETC Test Report 18878 84N 2. Rev. 2

. SOR Analysis 8215-959 in 1992 and 1993 SOR underwent a new qualification program (SOR Test Report 9058-102) and discovered that the o ring gland design on the vacuum piston was not capable of retaining maximum operating pressure after exposure to radiation, aging, and cycling.

SOR informed the NRC and the utilities of this condition on April 1,1993. The qualification test specimens were left in the test program with no modifications and continued to function properly and passed all tests with the exception of the hydrostatic test at the conclusion of the HELB and LOCA. SOR redesigned this seal .a meet hydrostatic requirements and qualified it by analysis (Ref. SOR Test Report 9038 102, Section 14. Appendix 4 Analysis 8923 219).

This condition v.as not discovered earlier because ofinadequate engineering testir; and analysis of the vacuum switch.

As noted in analysis 8923 219, examination of the test specimens revealed that the o ring was still sealing between the vacuum piston and the diaphragm, but leakage was occurring between the vacuum screw and the o ring. This is attributed to the trianguiar gland design and a combination of compression set, volumetric swell, and shrinkage which occurs from exposure to elevated temperatures and irradiation. All of these factors contributed to the loss of the line of contact between the vacuum screw and the o ring, and the resultant leakage at high hydrostatic pressures.

The redesign. which was released by an Engineering Order on May 20,1993, climina'tes the leak path mentioned above because the vacuum piston is now welded to the vacuum

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screw. In addoiun, the triangular o ring gland was changed to a face seal configuration.

A face seal is utilized on the pressure port o ring of the vacuum switch and has successfully retained 750 PSI hydrostatic pressure after exposure to radiation and thermal aging. The pressure pon o ring and the vacuum screw o ring are made of the exact same material (Parker compound E740 for option

• M9": Parker compound V709 for option "M4") and differ only in siu. The o ring gland dimensions are in accordance with the Parker 0 Ring llandbook for a static face sea s '*nd.

Regards.

cuyn- '$ b Y SO 7 Colbert Turney, \(P. Quality Joseph G. Modig. Engir.x i

l l

l l

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MtM 3 1468s W 105th Street

• Lenena Kansas 6621s-s964 usa Tel 913 898 2630 Fan 913 888 0767 ee i, i g ,,,e n Mr. Anil S. Gautam. NRC

Subject:

Information Notice ofJune 10.1994

Dear Mr. Gautam:

In July of 1993 SOR began to manufacture pressure. vacuum, and temperature switches that were qualified by SOR test repatt 9058-102. In May of 1994 SOR discovered a potential leakage problem in the conduit seals of these switches during routine testing and reported this discovery to the NRC and the afilected utilities on the subject Information Notice. This problem wer,t undetected by SOR for three reasons:

1. The leakage was a random problem.

2. Inspection steps were not adequate to identify faulty conduit seals.

3. There was a manufacturing error in the heat cure of the epoxy. This error was due to manufacturing personnel not following procedures. In addition, the Work Order format was not adequate because it did not require manufacturing personnel to record the actual heat cure temperature and cure time for each batch.

A As noted in the Information Notice, more stringent testing was instituted immediately (June,1994). This included the following steps:

1. An insulation resistance test was added to the Work Order format for conduit seals.

This step is signed off by the manufacturing personnel and reviewed and tigned off by QA personnel for every order of switches.

2. The insulation resistance test procedure was changed to include testing of wire to wire (all combinations) in addition to the standard wires to case test.

3. A 100 PSI leak test was added to the Work Order format for the completed conduit seal assembly. This step is sigr-d off by the manufacturing personnel and reviewed and signed off by QA personnel for every order of conduit seals.

4. A housing leak test was added to the assembly procederes in order to test the conduit seal after all assembly steps and all thermal testing is complete. The test pressure is equivalent to the HELB or LOCA pressure as applicable. This step is signed off on the assembly wrocedure by tee manufacturing personnel and reviewed and signed off by QA personnel for every order of switches.

_ in addition to the above steps. the following corrective action was taken in August 1994:

1

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l 1. The Work Order formats for the conduit seals were changed to require manufacturing to record the cure temperature and time for the epoxy. This will insure that the concet heat cure is used. This step is signed off by the manufacturing personnel and reviewed and signed off by QA personnel for every order of conduit seals.

Reference Conective Action Repon 0338.

Regards, o

Mu w?r Colben Tu,rney. V.P. Quality v 3. k f % ff 7 Joseph G. Modig. Engineer

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88c y' 4 UNITED STATES j ') W j NUCLEAR REGULATORY COMMISSION

$ ff WASHINoTON, D.C. 20566 4001 p.sv / ,

• ..* July 22, 1997 Mr. Richard P. Bender Vice President Yuasa Exide, Inc.

2366 Bernville Road, Post Office Box 14145, Reading, Pennsylvania 19612

SUBJECT:

HRC INSPECTION REPORTS 99900358/97-01 AND 99900359/97-01, NOTICE OF VIOLATION, AND NOTICE OF NONCONFORMANCE

Dear Mr. Bender:

On March 4-7, 1997, and on April 28-May 2,1997, the U.S. Nuclear Regulatory Commission (NRC) conducted an inspection at the Yuasa-Exide, Inc. (YEI),

facilities at Reading, Pennsylvania, and Richmond, Kentucky, respectively.

The enclosed report presents the results of those inspections.

During the inspections, the NRC inspectors found that certain of your activitias appeared to be in violation of NRC requirements. Specifically, the inspectors determined that contrary to Section 21.21 of Part 21 of Title 10 of the Code of Federal Regulatfons (10 CFR 21.21), YEI failed to inform all applicable licensees that certain YEI GN type battery cells that were manufactured between October 1992 and December 1992 could potentially have less than the manufacturer's publicized rated capacity of 8-hours. YEI sent letters to Southern California Edison (San Onofre Nuclear Generating Station) and Cleveland Electric Illuminating Company (Perry Nuclear Power Plant) informing these affected licensees of this deviation in GN type battery cells pursuant to 10 CFR 21.21(b), but not to the Washington Public Power Supply System (Washington Nuclear Plant, Unit-2).

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 follow the instructions specified in the er. closed NOV when preparing your response. The NRC will use your response, in part, to determine whether further enforcement action is necessary to ensure compliance with regulatory requirements.

In addittor., the NRC found that the implementation of the YEI quality assurance program failed to mest certain NRC requirements imposed on you by your customers. YEI did not comply with its Quality Assurance Manual requirements regarding documenting nonconforming material upon the failure of 2GN-15 cells for the San Onofre Nuclear Generating Station. Also, the measures YEI established for review of suitability of application of purchased parts and materials to be used in Class 1E batteries and verification that r

those purchased parts and materials met the procurement specifications were inadequate. The specific findings and references to the pertinent requirements are identified in the enclosures of this letter.

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Mr. Bender The failure of the San Onofre cells has potentially generic implications. The apparent susceptioility of the 2GN cell to premature lon of capacity at elevated temperatures may be indicative of an inherent weakness in some aspect of the design or manufacturing process. Should this be the case, it could result in the inability of station batteries to maintain required voltage for the reouired time under certain design basis conditions. In particular, elevated ambient temperatures due to loss of air conditioning during design basis events such as station blackout may impact the station batteries' ability to perform their safety function. The NRC believes that this pctential deviation from YEI's published product performance claims should be thoroughly investigated. If it is found to be a deviation, all affected licensees or purchasers should be informed in accordance with 10 CFR 21.21(b).

! Please provide us within 30 days from the date of this letter a written statement ,a accordance with the instructions specified in the enclosed Notice of Nonconformance. We will consider extending the response time if you can show good cause for us to do so.

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 (PDR). )

Sincerely, I Qfd Stuart A. Richards, Chief Special Inspection Branch Division of Inspection and Support Programs Office of Nuclear Reactor Regulation Docket Nos. 99900358, 99900359

Enclosures:

1. Notice of Violation

2. Notice of Nonconformance

3. Inspection Report 99900358,359/97-01 8

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NOTICE OF VIOLATION Yuasa Exide, Inc. Docket Nos.: 99900358/99900359 Reading, Pennsylvania / Richmond, Kentucky Report No.: 97-01 During NRC inspections conducted at ysu. facilities March 4-7, 1997, and April 28-May 2,1997, 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:

Section 21.21, " Notification of failure to comply or existence of a defect and its evaluation," of Part 21 of Title 10 of the Code of federal Regulations (10 CFR 21.21) became effective October 29, 1991. Section 21.21(b), states that if a deviation or failure to comply is discovered by a supplier of basic components, or services associated with basic components, and the supplier deternines that it does not have the capability to perform the evaluation to

' determine if a defect exists, or if the failure to comply is arso" ated with a substantial safety hazard, then the supplier must inform the purchasers or affected licensaes within five working days of this determination so that the purchasers or affected licensees may evaluate the deviation or failure to comply, pursuant to 10 CFR 21.21(a).

Contrary to the above, YE; failed to inform all affected licensees that certain YEl GN type battery cells that were manufar.ured between October 1992 and December 1992 could potentially have less than the manufacturer's publicized rated capacity of 8-hours. YEI sent letters to Southern California Edison (San Onofre Nuclear Generating Station) and Cleveland Electric Illuminating Company (Perry Nuclear Station) informing them of this deviation in the affected GN type battery cells, but not to Washington Public Power Supply System (Washington Nuclear Plant, Unit 2), which had received eight Gt!

type safety-related battery cells that were manufactured during the same time period. Violation 99900358,359/97-01-01.

This is a Severity Level IV violation (10 CFR Part 2, Appendix C, Supplement VII).

Pursuant to the provisions of la CFR 2.201, Yuasa Exide Inc., 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 bc clccrly ::: rked as a " Reply to a Notice of Violation" and should include for 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. Your response may reference or include previous docketed correspondence, if the correspondence adequately addresses the required response. Where good cause is shown, consideration will be given to extending the response time.

Dated at ,Rockville, Maryland this F day of July, 1997 Enclosure 1 l

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NOTICE OF NONCONFORMANCE Yuasa-Exide, Inc. (YEI) Docket Nos.: 99900358/99900359 Reading, Pennsylvania / Richmond, Kentuck/ Report No.: 97-01 Based on the results of an NRC inspection conducted at YEI facilities in l Reading, Pennsylvania, and Richmond, e,ert..cxy, on March 4-7, 1997, and on i April 28-May 2,1997, respectively, it appears that certain of your activitiet were not conducted in accordance with NRC requirements as follows:

l A. Criterion XV, " Control of Nonconforming Material," of Appendix B,

" Quality Assurance Requirements for Nuclear Power Generation Facilities," to Part 50 of Title 10 of the Code of Federal Regulations

. (10 CFR Part 50, Appendix B), requires in part that nonconforming items  ;

l shall be reviewed and accepted, rejected, repaired or reworked in l accardance with documented procedores.

Section 15, "honconforming Material, Parts, or Components," of the YEl Quality Assurance (QA) Manual required that QA personnel prepare material review reports in cases of nonconforming material and disposition the nonconformances.

Cortrary to the above YEI did not document and disposition four test failures of the SCE batteries in accordance with its 10 CFR Part 50, Appendix B quality assurance program in July 1996.

, Nonconformance 99900359/97-01-02.

B. Criterion III, " Design Control," of 10 CFR Part 50, Appendix B, requires a review for suitability of application for safety-related structures, systems, and components.

Contrary to these requirements, the measures established by YEI-Richmond, Kentucky, for review for suitability of application (commercial grade dedication procedures prescribed in QAP 70.0 and individual technical evaluations) of purchased parts and materials to be used in the manufacture of Class IE station batteries for nuclear power plants were not adequate as follows:

Not all critical characteristics for certain items were identified, e.g., 0-ring material and cure date, Not all verification methods or ecceptance criteria were appropriate or correct, or consistent with design documents (drawings or bills of materials), or purchase specifications (which themselves were not always consistent with design documents), or expressed in technically correct terms.

Not all engineering design drawing changes were incorporated into purchase specifications, technical evaluation or acceptance process attachments to QAP-70, or into incoming inspection report forms.

Nonconformance 99900358,359/97-01-03 Enclosure 2

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~

-C.- Criterion _ VII, " Control of Purchased Material, Equipment, and Services,"

of 10_ CFR Part 50, Appendix B, requires that measures be established to i

- verify, by review of suppliers and supplier documentation, and examination of products upon delivery, that the purchased material,-

equipment, or services, meet the procurement specifications.

Contrary to these requirements, the measures established by YEl-Richmond, Kentucky, for verification that purchased parts and materials to be used in the manufacture of Class IE- station batteries for nuclear power plants met the procurement specifications-were not adequate-as

.follows:

• For those critical characteristics that were identified, 'not all were adequately verified, e.g... material specified for. intercell connector fasteners was not adequately verified and the wrong-material was specified on incoming inspection report forms for post-seal c.aps.

• Commercial grade supplier surveys used to support verification of one or more critical characteristics for various items were broad-based programmatic reviews (not performance based) and without.

adequate specificity to verify that the supplier controls the critical characteristic of interest. Certificates of conformance were accepted from distributors whose ability to provide valid certificates was not verified.

Nonconformance 99900359/97-01-04 -

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, Offi_ce 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) the reason for the nonconformance, or if contested, the basis for disputing the nonconformance, (2)'the corrective steps that have been taken and Lne results achieved, (3) the corrective step:

that will be taken to avoid further noncompliances, and (4) the date when your corrective actions will be completed. Where good cause is shown, consideration will be given to extending the response time.

s Dated at Rockville, Maryland this22_?dayofJuly,1997 Enclosure 2

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

ORGANI7ATION

Yuasa Exide, Inc. (YEI) 2366 Bernville Road, i Post Office Box 14145, Reading, Pennsylvania 19612 l DOCKET / REPORT N0(s).- 99900358/97-01 (Reading, Pennsylvania) 99900359/97-01 (Richmond, Kentucky)

ORGANIZATION CONTACT: Richard P. Bender, Vice President, Quality Assurance and Purchasing (610)208-1972 Lana West, Quality Assurance Manager (606)6;.,-7346 NUCLEAR INDUFTRY: Manufactures and supplies stationary batteries and ACTIVITY battery racks.

INSPECTION Reading, Pennsylvania, March 4-7, 1997 l CONDUCTED: Richmond, Kentucky, April 28-May 2, 1997 TEAM MEMBERS: K.R. Naidu, Team Leader, NRR J.J. Petrosino, NRR S.D. Alexander, NRR S.N. Saba, NRR Approved By: G.C. Cwalina, Chief Vendor Inspection Section Special Inspection Branch Division of Inspection and Support Programs Office of Nuclear Reactor Regulation Enclosure 3

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1 INSPECTION

SUMMARY

The purpose of this inspection was to evaluate the quality assurance (QA) program and its implementation in the design, qualification and manufacture of rectangular, vertical plate, lead-acid battery cells (Type "GN") by Yuasa-Exide, Inc. (YEI), used in Class IE/ vital.st; tion batteries at nuclear power plants.

Conducted at YEI's engineering facility in Reading, Pennsylvania, and its

-factery in Richmond, Kentucky, the inspection focused on: (1) the implementation of the manufacturer's process controls, (2) procedural adequacy (including consistency with established requirements) and procedural compliance, (3) procurement and acceptance of purchased parts and materials used in battery manufacture (including commercial-grade dedication of components and parts for resale as basic components, e.g., battery racks and replacement parts), and (4) purchase orders (P0s) from NRC licensees and certificates of conformance (COCs) and associated docurrents provided to .;RC licensees.

Inspection bases were:

Appendix B, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Processing Plants," to Part 50 of Title 10 of the Code of Federal Reoulations (10 CFR Part 50, Appendix B) 10 CFR Part 21, " Reporting Defects and Noncompliknce" The inspectors identified two minor violations of 10 CFR Part 21 (521.21(a) ,

and 521.6) (See Section 3.1); one Level IV violation of 10 CFR 21.21(b) (See -

Section 3.2); one nonconformance with respect to Criterion XV of 10 CFR Part 50, Appendix B (See Section 3.5}; nonconformances with respect to Criterion 111 of 10 CFR Part 50, Appendix B (See Section 3.7); nonconformances with respect to Criterion VII of 10 CFR Part 50, Appendix B (Section 3.7); and one inspector followup item (Section 3.6).

2 STATUS OF PREVIOUS INSPECTION FINDINGS There have been no NRC inspections conducted since Exide became Yuasa-Exide, Inc. (YEI) and after the company reorganized under new management.-

3 INSPECTION FINDINGS AND OTHER COMMENTS 3.1 10 CFR Part 21 Procedure and Postina

a. Es.q,g.g The inspectors reviewed YEI's quality assurance procedure (QAP) for reporting in accordance with 10 CFR Part 21 (Part 21), QAP 80.0, "10CFR21 - Procedure for Reporting Non-Conforming Material," dated November 2, 1994. QAP B0.0 was developed by YEI to address Part 21 requirements at the two YEI facilities 2

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which perform activities reinting to " basic components." The inspectors also observed and reviewed the document that was posted at both facilities to l comply with 6 21.6 of,Part 21 " Posting requirements."  ;

b. Observations and Findinos The inspectors _ determined that QAP 80.0 included certain provisions of 10 CFR

-21.21(c) (NRC notification procedures), which are not required by the-. current revision to 10 CFR Part 21 to be included in procedures adopted pursuant to

. the regulation. However.-QAP 80.0 did not contain any of the required L , provisions of $21.21(a) (evaluation of deviations'and failures to comply,

interim reports,-'and notification of directors or responsible officers).

The inspectors also determined that the Part 21 posting at YEI's Reading

-facility was not in accordance with the requirements of 10 CFR 21.6. The inspectors-observed that YEl-Reading, apparently opting for a'521.6(b) posting, had postea only a .otice that indicated that a copy of 10 CFR Part 21 was available for review in its administrstive office. However, the notice

-lacked the other information required by 121.6(b), i.e., a description of the regulation and the Part 21 procedures, the location where the procedures (as well as the_ regulation itself) .may be viewed, and the name of the person to whom-reports should be made. In. addition, YEl-Reading did not post Section 206 of the Energy Reorganization.Act of 1974 as required by both $21.6(a) and 521.6(b). At its Richmond, Kentucky, facility, YEI had complied with the

- posting requirements of $21.6(a).

The_ inspectors discussed with the YEl staff the provisions of $21.21(b), which require that. deviations or. failures to comply, discovered by a supplier of basic components, for which the supplier determines that it does not have the capability to perform the evaluation of $21.21(a)(1) to determine if a defect exists, must be reported to the purchasers or affected licensees within five working days of this determination so that the purchasers or affected licensees may evaluate the deviation or failure to comply. The inspectors explained that although 521.21(b) is not specifically required to be' included in the procedures adopted pursuant to 10 CFR part.21, it is perhaps the most 4 important provision for a particular vendor / supplier's disposition of '

deviations or failures to comply, because most vendors or suppliers do not have the capability to perform a $21.21(a)(1) evaluation. As also defined in_.

Section 21.3, the 521.21(a)(1) evaluation is the process of determining whether a particular deviation constitutes a defect, i.e. whether it could create a. substantial hazard or lead to exceeding a license technical specification safety limit, or determining whether a failure to comply (with the Atomic-Energy Act of 1954, as amended, or any rule, regulation, order or license of the NRC) is associated with a substantial safety hazard. YEI agreed that this proce=s would normally be beyond its capability because, although YEI can advise a licensee or purchaser of. the effect of a particular deviation on tha performance of the battery, it cannot determine the ultimate effect on plant operation, reliability or safety.

The inspectors also explained that nothing in the regulation should be construed as prohibiting a report to the NRC by anyone who is concerned that a deviation may be a defect or that a failure to comply may be associated with a 3

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substantial safety hazard, even if they are not capable of performing a

$21.21(a)(1) evaluation. However, a vendor who is not qualified to perform I the 521.21(a)(1) evaluation should not perform the evaluation (in lieu of informing affected licensees or purchasers in accordance with 521.21(b)) and determine that a report to the NRC is not required because the deviation decs not appear to the unqualified vendor to be dsfect or because the failure to comply does not appear to be associated with a substantial hazard.

After discussing 10 CFR Part 21 responsibilities in detail with the YEI QA staff, the inspectors informed YE! that the failure to establish an adequate.

procedure and having an inadequate posting at its Reading facility were violations of 10 CFR Part 21. However, these failures constituted violations of minor significance and will be treated as a Non-Cited Violations, consistent with Section IV of the NRC Enforcement Policy. Subsequent to the Reading inspection, the YEI corporate staff drafted its. revision 'o QAP 80.0 and the draft document was discussed during the Richmond inspection,

c. Conclusion The inspectors concluded that YEI had not developed an adequate procedure to comply with the requirements of 10 CFR Part 21, and had not complied with 521.6 posting requirements at its Richmond facility. However, YEI had complied with the 521.6 posting requirements at its Richmond facility and has adequately revised QAP 80.0 to address the procedural requirements of 10 CFR 21.21(a) as well as including provisions to ensure compliance with 521.21(b).

3.2 10 CFR part 21 - Informino Affected licensees

a. ScoDe The inspectors reviewed an April 19, 1996, YEI letter that was sent to Southern California Edison (San Onofre Nuclear Generating Station) (SONGS) and Cleveland Electric Illuminating Company (Perry Nuclear Power Plant) (Percy) regarding GN type battery cells that were supplied. The YEI author also provided a copy of the letter to the NRC staff for information. The inspectors reviewed the letter to determine the adequacy of informing its customers of deviations or failures to comply,

b. Observations and Findinos The YEI letter indicated that certain YEI GN type cattery cells that were manufactured between October 1992 and December 1992 could potentially have less than YEI's publicized rated capacity of 8-hours. The letter stated to the two licensees and NRC staff that only Perry and SONGS received GN type batteries manufactured in the suspect time period.

During the Richmond facility inspection, the inspectors reviewed the letter and associated documents including YEl-Richmond's 1992 " custom order status log" (status log) for the suspect time period. The status log is an internal YEl-Richmond document that is used by YEI-Richmond quality control (QC) personnel and order entry personnel to maintain the status of all commercial and nuclear incoming battery cell orders that have special requirements l

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imposed. The inspectors noted that the status log reflected that both lice, sees that received the YEl letter had GH type cells manufactured in the subject time period.

However,-the inspectors determined that there was a third licensee that received cells manufactured during this. tine period. It was noted that Washington Public Power Supply System (WPPSS) had also received eight GN type safety-related battery cells. Further review of YEI's records showed that the WPPSS Purchase Order (PO) 221020, dated October 1992, did in fact order eighl GN batt.-ry cells and impossd nuclear safety-related requirements on YEl. YEl ,

committed to informing WPPSS of the deviation by July 25, 1997.

The inspectors determined that YEI had failed to inform one of the affected licensees of the deviation in affected GN type battery cells within the time limit required by 10 CFR 21.21(b). This has been identified as Violation i 99900359/97-Cl-vl.

As discussed in Section 3.3 below, the inspectors determined that the GC and GN cells are similar in design and performance and their manufacturing process contr ols are nearly identical. Therefore, the inspectors noted that GC cells 1 may be susceptible to the deviation discussed above. Although YE! has l qualified only the GN cells for safety-related service, some licensees buy the l commercial-grade GC cells and dedicate them. The inspectors found that Florida Power Corporation (FPC) and New York Power Authority (NYPA) purchased GC-Type cells manufactured during the period in question. FPC purchased 30 2GC-9 cells, assembled October 2,1992 (FPC P0 A730l!66), and NYPA purchased four GC-33 cells (PO number undetermined), assembled November 23, 1992, and tested December 11, 1992.

c. Conclusion The inspectors concluded that YEI did not perform an adequate review of its manufacturing records for the time period in question. Consequently, it did '

not identify the one other nuclear customer that received affected battery cells intended for safety-related applications. As a result, WPPSS was not informed of the deviation, a potential substantial safety hazard.

3.3 Battery Manufacturino Process 5- SS&D.t The inspectors observed processes in the various stages of manufacturing of YEI GN type battery cells. The inspectors examined items in production, reviewed logs and other process records and interviewed technicians to determine the adequacy and availability of the manufacturing instructions at the work stations and compliance with those instructions. The inspectors also reviewed production QA activities.

b.. Observations Ti.e inspectors observed various phases of cell manufacturing, including oxide mill operation, grid casting and assembly, flat plate manufacturing, paste mixing, machine pasting, positive and negative plate processing (paste curing, 5

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trimming, etc.), cell assembly and sealing, leak testing (15 GC-25 cells, of which two cells htd post seal-to-cover leaks and were put aside for repair and retesting) final formation of the cell; (electrolyte filling and a specified sequence of initial charging, discharging, recharging), and capacity discharge testing. In general, manufacturing instructions were available and being followed. Cperators were knowledgeable n. concerned with ensuring quality.

No concerns were identified in this area.

The inspectors observed QC personnel performing various prescribed routir-checks during the manufacturing process, including inspecting the plates Jr hairline cracks, plate thicknesses, loose / missing pellets, etc. No concerns were identified in this area.

The inspectors noted that while the GN (nuclear qualified Class IE) and GC (commercial) cells are generally similar in design and comparable in performance, tliare are some differences, including the following:

• The container material c' GN is a type of polycarbonate. The GC container material is a type of styrene.

The GN cover is polycarbonate; the GC cover is polyvinylchloride.

• The positive pastes of GN and GC cells have different amounts of lead.

The outside negative plates of the GN are thinner than the negative GC plates.

The GC cells have not been qualified by YEI for seismic requirements.

• The GN and GC plate separators are of different materials.

c. Conclusions The inspectors concluded that the YEI manufacturing process controls observed were effectively implemented.

3.4 Oualification of YEI GN-Seri-i Batteries for Class 1E Service

a. Scope The inspector reviewed Wyle Laboratories Report 45001-1, Revision A, dated January 15, 1989, the environmental and seismic qualification report for YEI's l "GN"-Series batteries and battery racks for Class IE service in nuclear power l plants and interviewed YEI's Manager of Engineering Support, Large Stationary l Batteries, who had witnessed the Wyle testing. The inspector also reviewed I the report prepared by Flight Dynamics, Inc., in which it documented the seismic qualification analysis for the YEI Series GN battery design change.

In the design change, the number of terminal posts was changed from two positive and two negative posts per cell to one positive and one negative post per cell to conform to the current design of YEI's similar cells for non-nuclear applications.

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b. Observations Although Class 1E batteries are typically not subject to exposure to the harsh environment of a design basis event such as loss-of-coolant accident or high energy line break (therefore not required to be qualified in accordance with 10 CFR 50.49), the Wyle test progr r vis intended to qualify them for the expected extremes of normal service conditions (in accordance with the applicable general design criteria of 10 CFR Part 50, Appendix A) and subject the test specimen batteries to seismic testing at an end-of-life condition.

Accordingly, the various test specimen batteries underwent accelerated thermal and radiation aging to the equivalents of 10,15 and 20 years of nuclear plant service (at Wyle Huntsville) and the order of 10,000 rads Co-60 radiation exposure (at The Georgia Institute of Technology) before undergoing seismic testing at Wyle. No deficiencies were noted in this report.

GN-Series batteries originally had two positive and two negative posts per cell (a total of eight posts in the usual 2GN or 2-cell-per-jar configuration) for ampacity reasons. However, the non-nuclear line had larger single posts to save weight and cost without sacrificing ampacity. In order to qualify the improved design, YEI contracted Flight Dynamics, Inc., who used finite element analysis to show that (1) the stresses during design basis seismic excitation were actually lower in the larger single posts and (2) the single-post design was stronger than the 2-post design that had been qualified by the Wyle tests cited above. No deficiencies were noted in this report.

c. Conclusions The original qualification test pr; gram and subsequent design-change reconciliation analysis for the GN-Series batteries and racks appeared to have been conducted in accordance with applicable requirements and guidelines in effect at the time including General Design Criterion 2 of Appendix A to 10 CFR Part 50, Regulatory Guide 1.100, Institute of Electrical and Electronic Engineers (IEEE) Standard 323-1974 (Class IE equipment qualification), IEEE Std 344-1975 (seismic qualification of Class IE equipment), IEEE Std 450-1987 (Class IE battery qualification), and IEEE Std 535-1987 (Class 1E battery seismic qualification). The inspectors had no concerns in this area.

3.5 Purchase Orders

a. Scope At the YEI Reading and Richmond facilities, the inspectors reviewed selected licensee purchase orders (P0s) for YEI GN type batteries and associated records to determine whether NRC licensees imposed the necessary and appropriate technical and quality requirements on YEI for the procurement of basic components,

b. Qbservations and Findinas The inspectors noted that each of the licensee P0s reviewed imposed the requirements of 10 CFR Part 50, Appendix B, and stated that 10 CFR Part 21 was applicable. Each of the YEI customer ff'es (called P0 packages) contained 7

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documents such as test and manufacturing records, C0Cs, and Lab Test Assignment (LTA) sheets. The LTA sheet is a form that is used as a manufacturing, test and quality assurance function traveler. The LTA for each particular job contairs all of the customer's specific manufacturing and test requirements. It is generated by Richmond QC personnel, in conjunction with the order entry department. The inspector: determined that all of the P0 packages reviewed contained appropriate quality and test requirements from the licensees except for one package discussed in Section 3.6 below.

c. Conclusion The inspectors concluded that YEI's customer order packages were well maintained, retrievable and reflected the required test results. With the exception of the test failures discussed in Section 3.6, no other anomalies were noted in this area.

3.6 Failure of Capacity Test Discharae by San Onofre Cells

a. Scope Ti.e inspectors reviewed Southern California Edison (SCE) P0 6L225004, dated March 1, 1995 (with subsequent revisions in October 1995 and August 1996), and associated documentation for ten 2GN-15 replacement calls for the San Onofre Nuclear Generating Station (SONGS). Also reviewed were test data for several other batteries with similar cells (Type 2GC) that had been capacity discharge tested at the 8-hour rate,

b. Observations and Findinas The test data indicated that ten 2GN-15 battery cells tested in July 1996 failed to meet SCE's original testing requirements on four occasions. The cells finally passed with acceptance criteria modified by SCE.

Although the test failures were largely attributable to some weak cells (e.g.,

50001), YEI explained that they felt that the elevated temperatures (as high as 79'F) were a significant contributing factor to the poor performance of the weakest cells. YEI explained that they had experienced difficulty in maintaining lower temperatures due to inadequate air conditioning in the t u t room and the July heat. This was YEI's justification for conducting the test three more times after the initial failure, all with similar unsatisfactory results.

Originally, the 10 cells were to be discharged at the temperature-corrected 8-hour rate, maintaining ambient temperature as close to 77'F as possible, to an average end voltage of 1.75 volts per cell (VPC) with a minimum of 90%

capacity per IEEE 450-1987. The temperatures during the discharges conducted by YEI on July 1, 4, 11, and 17, 1996, varied between 74*-79'F. However, in each test, the average cell voltage reached 1.75 VPC before 90% of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (7 4

hours, 12 minutes). The worst results were from the July 17th test conducted at 79'F. When the test was terminated at the 7-hour reading, the individual cell voltages ranged between 0.64-1,67 volts. In this test at the highest temperatures, most of the cells suffered degra'ied performance; although, cell 8

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50001 had exhibited the poorest performance consistently in the previous three test discharges. Cell 50002 was also weak, but not as bad and the other eight cells had performed consistently better during the previous tests run at lower temperatures. Thus, the data suggested that one or two cells were weak, but all the cells exhibited a sensitivity to high temperature.

Section 15, " Nonconforming Material, Parts, or Components," and Section 16,

" Corrective Action," of YEl's QA Manual required that quality personnel prepare a material review report and disposition the nonconformances in accordance with the QA program. However, YEI did not generate any material I review reports to document the deviation from the licensee's PO requirements and its disposition, as required by its QA procedures. Instead, according to

! YEl, after the fourth test failure, the YEI Richmond facility QC staff contacted YEI corporate engineering, and YEl engineering contacted the SCE engineering staff. After consultation between the YEI and SCE engineering staffs, SCE changed its testing requirements to 95% capacity at the 4-hour discharge rate. The cells rat this requirement during the fifth test anu were shipped to SONGS.

c. Conclusi,gn The inspactors concluded that YEl had established measures to assure that cor.ditions adverse to quality, such as test failures, and des.ations are promptly identified and corrected, and that YE! had procedures to assure that l nonconforming items would be reviewed and accepted, rejected, repaired or reworked in acccrdance with documented procedures. However, YEl failed to document and disposition the four test failures of the SONGS cells in accordance with its 10 CFR Part 50, Appendix B quality assurance program.

Accordingly, Nonconformance 99900359/97-01-02 with respect to Criterion XV of 10 CFR Part 50, Appendix B, was identified in this area.

3.7 Safety Imolications of Failure of SONGS Cells

a. Scope The failure of the San Onofre cells is of technical concern with potentially generic implications. In addition to concerns about the requirements for SONGS batteries, the inspectors were concerned that perhaps YEI's advertised 8-hour capability for the 2GN cells in its product literature was not always achievable, particularly under elevated ambient and cell temperature conditions. To try to resolve concerns raised by the failure of the SONGS cells to pass their 8-hour capacity test discharge, the inspectors reviewed test data for several other groups -f similar cells for other plants to determine whether there was any inherent difficulty in YEI 2GN cells meeting the manufacturer's published 8-hour discharge capability,

b. Observations The 8-hour-rate capacity test data for several other batteries with sir .lar cell types (2GC) w'.th no anomalous cells and no elevated temperatures showed that all the cells exhibited greater than 90-percent capacity at the 8-hour rate. Some, at the lowest temperatures, were above 100 percent. However, 9

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there were no data readily available for other 8-hour discharges at temperatures above 77'F. Therefore, the inspectors could not rule out generic susceptibility to premature loss of capacity at elevated temperaturss.

c. Conclusion Based on the cell test data reviewed, the ins'ectors concluded that YE! cells of this type should be able to meet an 8-hour rate, 90-percent capacity t requirement within the bounds of expected ranges of normal service conditions in nuclear plant battery rooms. The inspectors further concluded that a significant factor contributing to the failures of the group of 10 cells for l SONGS was the poor performance of Cells 50001 and 50002. However, the apparent susceptibility of the 2GN cell to premature loss of capacity at elevated temperatures may be indicative of an inherent weakness in some aspect j of the design or manufacturing process. For example, according to YEI, the '

2GN cell is of a relatively low electrolyte volume design. This feature causes the cell to exhibit a capacity-versus-discharge-rate profile typical of lead-acid cells up o about the 4-hour rate. However, for longer discharges /  ;

lower rates, the cell appears to suffer from electrolyte depletion and starts  !

to exhibit reduced capacity in very long discharges. The inspectors were l concerned that should this become limiting, particularly at high temperatures,  ;

it could result in the unexpected inability of station batteries to maintain i required voltage for the required time under certain design basis conditions. I In particular, elevated ambient temperatures due to loss of air conditioning )

during design basis events such as station blackout may impact the station batteries' ability to perform their safety function. Accordingly, the inspectors strongly recommended that this be investigated and that should it be determined to be a deviation, affected licensees and purchasers should be informed in accordance with 10 CFR 21.21(b). This issue was identified as Inspector Followup Item 99900358,359/97-01-05.

Instead of crdering a replacement for the weak cells, SCE revised the test acceptance criteria. The inspectors were not able to determine at YEI what the basis for SCE's original specification was, whether the criterion of 95 percent at the 4-hour rate was appropriate, or whether the weak cells would adversely impact the performance of one of SONGS's Class IE station batteries.

3.8 YEI Commercial Grade Dedication Proaram GN-Series batteries are designed and manufactured under YEI's 10 CFR Part 50, Appendix B, QA program and accordingly, are supplied to NRC licensees as basic components as defined in 10 CFR 21.3. YEI uses the provisions of its commer-cial grade dedication program as a systematic means to verify that purchased material and components used in the manufacture of Class IE GN-type battery cells are suitable for safety-related service.

In addition, the seismically qualified battery rack systems, manufactured to YEI specifications by the KIH Company, are purchased by YEI as co'mercial grade items, dedicated by YEI and suppliu. to NR: licensees as basic components. An adequate and effectively implemented comercial grade dedication program, being an activity affecting quality, functions under the applicable controls of the vendor's 10 CFR Part

! 50, Appendix B, QA program. However, it must meet, in particular, the re-l quirements of Criterion III, " Design Control," and Criterion VII, " Control of l

Purchased Material, Equipment, and Services," of 10 CFR Part 50, Appendix B.

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a. ScoDe a.1 Procedures and Technical Evaluation / Review for Suitability In order to evaluate the effectiveness of the YEI QA program and its implementation regarding controls applicable to the review for suitability of application and design verification in accordance with Criterion III, the inspector reviewed YEI Procedure QAP-70.0, " Dedication of Commercial Grade Items (CGI) for Nuclear Safety Related Application:," revision dated April 70, 1994. The review included the attachments to QAP-70.0 which comprise the component technical evaluations and associated acceptance process sheets. For reference, component part manufacturers' technical information, YEI design documents (principally drawings), and YEI procurement specifications, were also reviewed, a.2 Acceptance /Procurament Specification Compliance To evaluate the effectiveness of the YE! QA program and its implementatiur, regarding controls applicable to supplier selection and qualification, review of supplier documentation and examination of purchased material and components in accordance with Criterion VII for verification of satisfaction of l procurement specification requirements, the inspectors reviewed acceptance process sheets, incoming inspection reports (IIRs), supplier audits and l commercial grade surveys, independent laboratory material analysis reports, l

procurement documents and supplier certificates of conformance. The-inspectors also interviewed technicians and QA/QC personnel and examined

clected purchased component parts and materials,

b. Observations and Findinos b.1 Procedures and Technical Evaluation / Review for Suitability The inspector determined that 0AP 70.0 was not fully consistent with the requirements of Criteria III and VII of 10 CFR Part 50, Appendix B, and the provisions of Electric Power Research Institute (EPRI) Report NP-5652,

" Guideline for the Use of Commercial Grade Items in Nuclear Safety Related Applicat P7s(NCIG-07)." The procedures were also not consistent with certain provisions of NRC Generic Letter 89-02, " Actions to Improve the Dedication of Counterfeit and Fraudulently Marketed Products," issueo March 21, 1989 and NRC GL 91-05, " Licensee; Commercial-Grade Procurement and Dedication Programs,"

issued April 9,1991, in which the NRC promulgated clarifications of staff positions tan key issues, later incorporated into the revision of 10 CFR Part 21 that became effective November 1995.

QAP 70.0 did not contain the restrictions from NRC Leneric Letter 89-02 on the use of Acceptance Methods 2 (commercial grade surveys) and 4 (product and supplier performance history) alone of EPRI P,eport NP-5652. However, the inspector noted that QAP 70.0 and its attachments, which comprise the technical evaluation and acceptance worksheets for dedication of individual GN component parts and materials and for the battery rack components, did prescribe multiple acceptance methods for many items and relied predo,ninantly on Method 1 (special tests and inspections) when a single method was employed.

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QAP-70.0 defined critical characteristics as in EPRI NP-5652, i.e., measurable quantities that when verified provide reasonable assurance that the item received is the item specified. It then defined critical characteristics for design and acceptance as in EPRI NP-64C6, " Technical Evaluation of Replacement items (NCIG-ll)." However, these definitions and the stated position (as in NP-6406) that critical characteristics far acceptance are a subset of critical characteristics for dedgn are not cun.,istent with the intent of GL 89-02, nor the explicit definition of and NRC staff position on critical characteristics as promulgated in GL 91-05, nor the definition of critical characteristics contained in the revision of 10 CFR Part 21 that became effective in 1995.

The NRC position is that critical characterictics when verified provide reasonable assurance that the item will perform its safety functions (not necessarily all design functions) and not fail in a manner detrimental to safety under all design basis conditions.

During the inspector's review of selected QAP 70.0 technical evaluation and acceptance precess worksheets for the battery racks and various batte y components and materials, the lists of critical characteristics for acceptance and their associated verification methods and acceptance criteria were not all complete and consistent with each other or with plant application requirements. The technical basis or rationale for the selection of characteristics, verification methods and acceptance criteria was not apparent l in some cases. For example, QAP 70.0, No. 100-Series Attachments, the l technical evaluation and acceptance process worksheets for the battery rack

' and components did not address the integrity of the welds in the fabricated rack components, nor the integrity of the bolted joints (e.g., stiffness, bolt torque, etc.) of installed racks. In another instance, terminal post 0-ring material and cure date were omitted. Sir.ce the battery racks must remain structurally sound, and the cell connections remain tight during a dasign basis earthquake in order to ensure the operability of the safety-related station batteries, weakness in the welds (or loose / broken connections) could lead to rack failure and battery failure and thus prevent the batteries from performing their safety functions.

In addition, the inspectors found that not all acceptarce criteria were appropriate or correct, or consistent with design do ments (drawings or bills of materials), or perchase specifications (which .emselves were not always consistent with design documer.ts), or expressed in technically correct terms.

One reason for this was because not all engineering design drawing changes were incorporated into purchase specifications, technical evaluations, or acceptance process sheet attachments to QAP-70, or into IIR forms. The inspectors' sampling review indicated that for post seal caps, the wrong material, i.e., not in accordance with the latest revision of the design drawing, was given on the incoming inspection report form. In another example, the incoming inspection frm specified the wrong durometer hardness value for post-seal 0-rings. Although this had been corrected by a pen-and-ink change on the IIR forms, the QAP-70 acceptance process sheet had a number from the previous drawing revision (D) and the P-spec had yet another number.

Having incorrect acceptance criteria may fail to datect components that are not capable of performing their safety functions under all application design basis service conditions.

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b.2 Acceptance / Procurement Specification Compliance

-For -those critical characteristics that were identified, not all were correctly specified or adequately verified. The inspectors identified

-material as an example of inappropriate verification method and acceptance criteria that was seen througnout t h 4 % ted job files reviewed. Material was specified on most of-the acceptance process sheets, but-the verification method -visual inspection-upon receipt, did not actually verify material. The

-IIR forms (for which there was no-procedure for preparation or for use and- l l little training beyond 0JT) where material was identified as a= critical '

characteristic, would simply say " material" under one of-the critical l

characteristic-column headings. For example, !!Rs for intercell connector fasteners,-specifically,-bolts,-specified "55-316," for the material and the l -blocks for each sample ~ specimen would simply say "yes." In effect, entering a yes just reflected the belief that the material was correct based on markings and other factors. IIRs for post seal caps, showed that the material was similarly inappropriately verified.

The inspectors followed up on the question of intercell connector fastener

-- material in more detail.- The material specifieo for intercell connector- i fasteners, expressed as "SS-316," was supposed to be ASTM Type 316 stainless steel. YEI had Singleton Laboratories perform annual chemistry analyses on

. samples of various materials including these fasteners to confirm proper material. However, YEI's implementation audits / commercial grade surveys of the suppliers of these fasteners-(who were distributors -not the-manufacturers) Threaded Screw Products-(TSP), Inc., and PM Fasteners, Inc.,

were not performance based or item and critical characteristic specific. They did not document objective quality evidence that the suppliers obtained valid, lot-traceable information (e.g., CMTRs) on material and fabrication from the fastener manufacturers, nor did they document objective quality evidence of-the' suppliers' commercial quality controls to ensure that substandard or fraudulent material was not commingled. Therefore, the audits were not usable as a-basis for acceptance of the fasteners from lots other than the ones from which actual analysis samples were taken. YEI also did not maintain lot traceability on the fasteners. YEI.did not know, for example, whether their suppliers obtained the fasteners from the same manufacturers (the ~ markings to be verified were inadequately described) or whether the manufacturers had adequate material contro', nor had YEI documented the history of the consistency of this attribute (material). Finally, the audits were not consistent with the restrictions:on the use of EPRI Method 2 in NRC GL 89-02 (nor was this mentioned in QAP-70.0). .Therefore, in view of these deficiencies, the significant instances of substandard or fraudulent fasteners on the market, many from Asia,.in view of the Asian origin of the received caterial, and-in view of the non-standard markings on the fasteners, the inspectors concluded that the yearly sample analysis was not adequate to

. ensure consistent material suitability. The inspectors also noted that the COCs from TSP were not in accordance with the P0 requirements (B and N of QAP-500). Although the signature blocks had the typed name of a person presumably-in-authority, it appeared that none of the COCs were signed by the named person, but rather by two different subordinates who signed the name of the designated person in authority's name instead of signing their own name with the annotation "for" the named person. It was also not known whether the 13

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actual signers had proper authority or whether the named person had ever reviewed or approved the C0Cs. The fasteners were confirmed to be non-magnetic, consistent with an austenitic stainless steel such as 316. In addition, YEI milled a bolt taken from stock to check for indication of irregularities such as welded-on bolt heads. No such indications were apparent by visual examination.

An example of a technically incorrect expression of a specification was the electrical resistance specification in the purchase specification for Amerace ACE-SIL plate separators. Inconsistent with the separator manufacturer's technical information, the YEI purchase specification, acceptance process sheet, and the IIR form expressed this parameter in terms of ohms / square inch / mil of separator web thickness. Separator resistance (as used in battery parlance) is actually conduction path length-specific resistr.nce (resistivity normalized on conduction path cross sectional area) and is properly expressed in units of ohm-inch / mil of separator web thickness. Although the separator resistance was specified using incorrect units, the inspector's calculation confirmed that the numerical values specified were consistent with the manufacturer's specifications.

c. fantlusions n

The deficiencies in the YEI commercial grade dedication program description documents is ansidered a weakness in the YEI 07. Program with respect to conformance to Criteria III and VII of 10 CFR Part 50, Appendix B. In addition, the YEI definition of the term critical characteristic in 0AP 70.0 was inconsistent with 10 CTR Part 21.

The inspectors concluded that contrary to these requirements, the measures established by YEl-Reading, and implemented by YEI Richmond, Kentucky, for reviev for suitability of application (CGI dedication procedures prescribed in QAP 70.0 and individual technical evaluations, acceptance process sheets, purchase specifications and prepared IIR forms) for purchased parts and materials to be used in the manufacture of Class IE station batteries for nuclear power plants did not meet the requirements of Criterion III of 10 CFR Part 50, Appendix B. Accordingly, Nonconformance 99900358,359/97-01-03 was identified.

In addition, YEI's measures for verification that these purchased parts and materials met the procurement specifications did not meet the requirements of Criterion VII of 10 CFR Fart 50, Appendix B. Accordingly, this was identified as part of Nonconformance 99900359/97-01-04.

3.9 Sucolier Quality Audits

a. $_qngg The inspectors reviewed YEI audit and commercial grade survey procedures and checklists and also reviewed audits performed by YEI on its vendors. The inspectors evaluated them to determine if YEI auditors verified critical characteristics of the items furnished by those suppliers, and if the audits were performance based.

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b. Observations and Findinas According to the audit reports reviewed. YEI used checklists to perform the audit / surveys of Cobra Wire and Cable, Inc., Amerace-Microporous Products, Inc., I.E. Dupont De Nemours & Co, Threaded Screw Products, Inc., and KIM Engineering Company (KIM), who-supplied cable, separators, acid, bolts & nuts, and battery racks respectively. YE! documer.ted adverse conditions in Audit Corrective Action Requests (ACARs) and requested suppliers to return the ACARs with the proposed actions to correct them. YEI then closed the ACARs if the proposed corrective actions were acceptable. l 1

The inspectors determined that the attributes in the audit checkli*t e common to all suppliers and covered a broad programmatic overview v s supplier rather than focusing on the specific item being supplied.

Furthermore, the audit did not attempt to verify the control of the quality of j the critical charactaristics of the specific item being manufactured.

For example, the audit of KIM failed to specifically verify the qualitications i of the welders who welded the steel components of the rack and the quality control welding inspectors (QCWIs) who inspected the welds. The audit did not reveal that the QCWIs were using a welding inspection checklist that contained all the attributes of an acceptable weld (such as, size, undercut, porosity, and length).

QAP 30.9, " Rack Welding requirements and Welder Qualifications," dated October 15, 1982, stated, in part, " Responsibility for the qualification and reexamination rests with Exide working with their suppliers as to requalification and certification of the welders." The YEI auditors did not verify this attribute. YEI management acknowledged this weakness and committed to take adequate corrective action by developing supplemental item- ,

specific checklists. I Similar concerns regarding the audit / survey of Threaded Screw Products is .

discussed in Section 3.8 above.  !

c. Ccaelusion The inspector's review of YEl ., supplier quality audit and commercial grade survey procedures and checklists and selected supplier audit / survey reports revealed that they were a broad-brush, programmatic review, not performance based. . Such surveys / audits could provide a reasonable basis for preliminary qualification of a commercial grade supplier (e.g., placing the supplier on an approved commercial grade suppliers list), and may be iseful in managing supplier quality resources, but they were not critical characteristic-specific and item-s M fic, and inconsistent with NRC Gl. 89-02. The audits / surveys did not adequately verify that the suppliers' comercial quality programs were effectively implemented, and did not cover distributors' programs where applicable. The inspectors concluded that YEI's qualification of certain suppliers and examination of supplier documentation did not meet the requirements of Criterion VII of 10 CFR Part 50, Appendix B. Accordingly, this concern is identified as part of Nonconformance 99900358,359/97-01-04.

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1 3.10 Seismic Racks

a. ELQDR To evaluate the requirements that YEI established to ensure that the seismic racks are manufactured to meet or exceed the requirements of those racks that met the seismic qualification tests, the inspectors reviewed QAP 30.8,

" Quality Assurance Control-$eismic Racks," dated November 16, 1994, QAP 30.9,

" Rack Welding requirements and Welder Qualifications," dated October 15, 1982, and YEI procurement specification P-011, Section 07, " Welding Steel Racks,"

dated October 1, 1982.

b. Observations and Findinas QAP 30.8 referenced QAP 30.9 and P-Oll, Section 07. QAP 30.9 referenced American Welding Society (AWS) Standard 01.1, as the applicable standard for welding the racks ar.d specified the following:

• Filler material to be AWS A5.18 Z 70S-3 or equivalent i

• Welded material to be ASTM A-36 of specified minimum yield strength

• Welding to be done in horizontal position

• Weld profiles-to be in accordance with AWS D1.1, Sections 2.7 and 3.6

• Weld Procedure Specification and Welder Qualification to be in accordance with AWS D1.1, Parts 8 and C

• Weld quality as required by AWS DI.1, Paragraphs 5.10.3 and 5.11.2

• Visual inspection in accordance with AWS D1.1, Paragraph 5.6.3

• -Weld size and location as specified on Drawing MC-83860, " Frame - Steel

- 2 Step-(G)," Revision E, dated October 3, 1985 The inspectors noted that Drawing MC-83860 had no weld dimension tolerar ces and observed that.the engineering specifications were scattered among various documents and not consolidated in the drawing to facilitate QC inspection of the welds to the applicable requirements. YEI committed to revise the drawing which is frequently used by welders as well as inspectors to ensure that all welds meet or exceed the quality of the qualified specimen.

c. Conclusion i The engineering specifications for the weld and assembly of the racks were adequate, but were found to be scattered in various places instead of being consolidated in one place. The lack of colocated specifications and the audits or surveys of the KIM Company not verifying welder qualifications were weaknesses in YEI's QA control of special processes which YEI cont.itted to strengthen.

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3.11 Internal Audits

a. - iggtg in order to determine if QC performed internal audits to verify the effectiveness of the quality assurance progr3m, the inspectors reviewed Quality Procedures Manual-(QPM) 4.17 "Ir.iersial Quality Audits," dated October 7,1996, as well as selected internal audit reports.

b. Observations and Findinas-According to the records, the internal addit performed by QA on April 3,1997, to verify compliance with-QPM 4.3.1, " Contract Review" dated July 10, 1995,

..-identified 19 nonconformances. The audit conducted March 31, 1997, to verify compliance with QPM 4.18. " Training," dated March 13, 1997, identified several j nonconformances; three of them repeated. The inspectors noted that corrective l action for the repeat nonconformances was being implemented, but full com% innee had not yet been achieved. The inspectors determined that-coristive action requests (CARS) were written to identify nonconformances in all-instances and were sent to the cognizant manager to determine the root-cause of the problem, and> document the corrective action.taken or recommend.

the proposed-corrective action, c,- Conclusion The inspectors concluded that plant QA auditors performed internal- audits in -

accordance with procedures to verify that quality activities comply with the planned arrangements and to determine the effectiveness of the quality system, audits were_. legible and retrievable, and corrective action being taken was being. monitored.- No concerns were identified in this area.

3.11 Trainino

a. Ltn.g I The inspectors reviewed selected training records of craftsmen to determine.if they had -aceived training for the activities they were performing.

b. Observations and Findinos The inspectors reviewed the training folders of five craftsmen working in the pasting, oxide, burning, and gluing areas. The folders contained training record sheets that documented the date.and subject of the training and the name of the instructor. -The sheets were signed by attendees, acknowledging -

the training received. The training records were legible, retrievable, and complete.

[ c. Conclusion The inspectors determined that personnel were trained in the areas in which they were working.

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4 PERSONS CONTACTED 1.C. Baeringer, Vice President, Engineering R.P. Bender, Vice President, Quality Asstrance and Procurement C. Claypool, Quality Control Technician J. Hall, Quality Control E. Simpscn, Quality Control L. Rickman, Plant Manager B.P. Lightner, Manager, Supplier Quality M.A. Patel, Manager, Engineering Support, Large Stationary Batteries S.J. Weik, Manager, Design Engineering and Document Control L.R. West, Manager, Quality Assurance, Richmond ITEMS OPENED. CLOSED. AND DISCUSSED Opened 99900358,359/97-01-01 VIO Informing Affected Licensees of Deviation 99900358,359/97-01-02 NON Documenting /Dispositioning Test Failures 99900358,359/97-01-03 NON Review for Suitability of Application 99900358,359/97-01-04 NON Product Verification / Supplier Auditt/ Surveys Closed None. No prior open items.

Discussed 99900358,359/97-01-05 IFI Elevated temperature effect on GC and GN cells 18

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Selected Generic Correspondence on the Adequacy of Vendor Audits and the Quality of Vendor Products identifier Tda Information Notice 97-45 Environmental Qualification Deficiency for Cables and Containment Penetration Pigtails information Notice 97-53 Circuit breakers Left Racked Out in Non-Seismically Qualified '

Positions Information Notice 97-59 Fire Endurance Test Results of Versawrap Fire Barriers

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NRC FORM 335 U.S. Nuct 8 AR REoVLATORY commission 1. REPORT NUMBEA p.egy (Assigned by Netc, Add VA, supp., Rev.,

NRC.J 1102.

and Addendum Numbers, N any.)

m :2m BIBI,10 GRAPHIC DATA Si4EET (see merrucsonm N ->

NUREG-0040

2. mLE AND SUBTITLE Vol. 21, No.3 Ucensee Contractor and Vendor inspection Status Report 3 DATE REPORT PUBLISHED Quarterly Report -uowfi~ YEAR July September 1997 November 1997

4. FIN OR GRANT NUMBER 5 AUTHOR (S) 6 TYPE OF REPORT Quarterly

7. PERIOD COVERED (Inciumve Defes)

July - September 1997 8 PE , FORMING ORGANtZATION . NAME AND ADORESS (r NRC, powoe owson, omco or Regen, u 3 Nucsser Roguewy Comasace, eno medag eness. #contracks, comoe name ano meane eneu >

Division of Reactor Controls and Human Factors Offica of Nuclear Reactor Regulation U S. Nuclear Regulatory Commission W:shington, DC 20555-0001 9 SPONSORING ORGANtZATION NAME AND ADDRESS tit NRC. type *Same es enove*, # contractor. Powoe NRC Dwoon. omcw or Regen. v s Nucseer Reguderwy comasson, and mesng a:k>ss )

Sime cs above 10 SUPPLEMENTARY NOTES 11 /;BSTRAn T (200 mortis or dess)

This periodical covers the results of inspections that were performed by the NRC's Special inspection Branch, Vendor inspection Section, and that were distributed to the inspected organizations during the period from July through September 1997

12. KEY WOROSloESCRIPTORS (L,st ment orpareses inet ar asest researctwa n eceang N recort) '3 ^V^*'un stATEmNr V;ndor inspection Unlimited 14 SECURITY CLASSIFICATION (This Pepe)

Unclassified TrGGer)

Unclassified 15 NUMBER OF PAGES 16 price NRC FORM 335 t?49)

Ttus form wee em:trorucasty prochced by Ehte Federei Forme. Wu

Printed on recycled paper Federal Recycling Program

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