NUREG-0040 

Vol. 2130.3 

Licensee Contractor 

and Vendor Inspection 

Status Report 

Quarterly Report 

July - September 1997 

U.S. Nuclear Regulatory Commission 

Office of Nuclear Reactor Regulation 
AVAlLABl LITY N OTlCE 

Availability of Reference Materials Cited in NRC Publications 

Most documents cited in NRC publications will be available from one of the following sources: 

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

20555-0001 

2. The Superintendent of Documents, U.S. Government Printing Office, P. 0. Box 37082, 

Washington, DC 20402-9328 

3. 

Although the listing that follows represents the majority of documents cited in NRC publications, it is not intended to be exhaustive. 

The National Technical Information Service, Springfield, VA 221 61 -0002 

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 documents and correspondence. 

The following documents in the NUREG series are available for purchase from the Government 

Printing Office: formal NRC staff and contractor reports, NRC-sponsored conference proceedings, international agreement reports, grantee reports, and NRC booklets and brochures. Also available are regulatory guides, NRC regulations in the Code of Federal Regulations, and Nuclear Regulatory Commission Issuances. 

Documents available from the National Technical Information Service include NUREG-series 

reports and technical reports prepared by other Federal agencies and reports prepared by the 

Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission. 

Documents available from public and special technical libraries include all open literature 

items, such as books, journal articles, and transactions. Federal Regisfer notices, Federal 

and State legislation, and congressional reports can usually be obtained from these libraries. 

Documents such as theses, dissertations, foreign reports and translations, and non-NRC conference proceedings are available for purchase from the organization sponsoring the publication cited. 

Single copies of NRC draft reports are available free, to the extent of supply, upon written 

request to the Office of Administration, Distribution and Mail Services Section, U.S. Nuclear 

Regulatory Commission, Washington, DC 20555-0001 . 

Copies of industry codes and standards used in a substantive manner in the NRC regulatory 

process are maintained at the NRC Library, Two White Flint North,f 1545 Rockville Pike, Rockville, MD 20852-2738, for use by the public. Codes and standards are usually copyrighted 

and may be purchased from the originating Organization or, if they are American National 

Standards, from the American National Standards Institute, 1430 Broadway, New York, NY 

10018-3308. 

A Year's SubscriDtion of this reDort consists of four auarferlv issues. 
NUREG0040 

Vol. 21, No. 3 

Licensee Contractor 

and Vendor Inspection 

Status Report 

Quarterly Report 

July - September 1997 

Manuscript Completed: November 1997 

Date Published: November 1997 

Division of Reactor Controls and Human Factors 

Oflice of Nuclear Reactor Regulation 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 
~ ~~ 

NUREG-0040, Vol. 21, No. 3 has been 

reproduced from the best available copy. 
DISCLAIMER 

Portions of this document may be illegible 

electronic image products. Images are 

produced from the best available original 

document, 
DISCLAIMER 

This report was prepared as an account of work sponsored by an agency of the 

United States Government. Neither the United States Government nor any agency 

thereof, nor any of their employees, makes any warranty, express or implied, or 

assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents 

that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or impiy its endorsement, recommendation, or favoring by the United States Government or any agency thereof. 

The views and opinions of authors expressed herein do not necessarily state or 

reflect those of the United States Government or any agency thereof. 
ABSTRACT 

This periodical covers the results of inspections performed between July 1997 and 

September 1997 by the NRCs Special Inspection Branch, Vendor Inspection Section, that have 

been distributed to the inspected organizations. 

iii 

CONTENTS 

PAGE 

Abstract ....................................................................................................................... i ii 

Introduction ................................................................................................................. vii 

Inspection Reports ...................................................................................................... 1 

ABB Services, Inc. 

Columbia, MD 

(99901281/97-01) ................................ 2 

Hartford Steam Boiler (99900601/97-01) ............................... 19 

Inspection and Insurance Company 

Hartford, CT 

NUS Instruments, Inc. 

Idaho Falls, ID 

Pacific Gas & Electric Company 

Diablo Canyon Power Plant 

San Francisco, CA 

PECO Nuclear 

Wayne, PA 

(99901320/97-01) ................................ 40 

(50-275/97-201, 50-323197-201) ........... 52 

(50-352/96-201, 50-353/96-201) ........... 68 

Tritium Target Qualification Program (99900541/97-01) ................................ 94 

Pacific Northwest National Laboratory 

Richland, WA 

SOR, Inc. 

Lenexa, KS 

(99900824/97-01) ............................... 116 

Yuasa Exide, Inc. 

Reading, PA 

(99900358/97-01 , 99900359/97-01) ... 132 

Select Generic Correspondence on the Adequacy of Vendor .................................... 155 

Audits and the Quality of Vendor Products 

V 
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. 

viii 
INSPECTION REPORTS 

-1 - 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-MP31 

July 23, 1997 

Mr. John J. Connolly, Vice President 

Business Development 

ABB Service Incorporated 

9050-A Red Branch Road 

Col umbi a , Mary1 and 2 1045 

SUBJECT: NRC INSPECTION REPORT NO. 99901281/97-01 

Dear Mr. 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 discussion 

conducted 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 several 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. 

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 

action 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 January 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-re1 ated K-Line 

breakers. Similarly, ABB Service did not apprise their refurbishment 

facilities or licensees of the weakness in testing. 

apprised licensees of the testing weakness, a report to the NRC would likely 

have been generated. 

Please note 

If ABB Service had 

-2- 
Mr. John J. Connolly -2- 

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. 

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. 

The staff is also 

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 assembled 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). 

Docket 99901281/97-01 

Encl osures : 1. 

2. 

3. 

Sincerely, 

Stuart A. Richards, Chief 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Notice of Violation 

Notice of Nonconformance 

Inspection Report 99901281/97-01 

cc: Mr. Joseph M. Tate 

General Manager 

AB6 Service Inc. 

Regional Service Center 

5311 Comnerce Street 

Cleveland, OH 44130 

-3- 
Enclosure 1 

NOTICE OF VIOLATION 

ABB Service 

Cleveland, Ohio 

During an NRC inspection conducted on May 14 

between July 14-22, 1997, a violation of NRC 

accordance with the "General Statement of Po 

Enforcement Actions," NUREG-1600, the violat 

Docket No.: 99901281 

1997, and discussions conducted 

requirements was identified. In 

icy and Procedure for NRC 

on 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 5 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 

signi f icant nonconformance in nucl ear-safety-re1 ated equipment shall be 

evaluated to determine if a 10 CFR Part 21 report needs to be filed. 

AB8 Service QAP 15.2, "Reporting of Defects and Noncompliance in Accordance 

with 10 CFR 21," Revision 2, approved 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. 

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-600s low voltage breaker (Serial 

Number 518170-211135), and identified the problem to ABB Service on April 28, 

PNPP identified three refurbished 

1997. (99901281/97-01-01) 

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

-4- 
Notice of Violation -2- 

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 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 af 

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 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. 

docketed correspondence, if the correspondence adequately addresses the 

required response. 

extending the response time. 

Your response may reference or include previous 

Where good cause is shown, consideration will be given to 

Dated at Rockville, Maryland 

this L. day of July 1997 

-5- 
Enclosure 2 

NOTICE OF NONCONFORMANCE 

ABB Service Docket No.: 99901281 

C1 eve1 and, Ohio 

Based on the results of an inspection conducted on May 14, 1997, it appears 

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

requirements. 

Criterion X of Appendix B 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) 

0 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 51817D-211135. 

Please provi 

Regulatory C 

with a copy 

and Support 

the date of 

should be cl 

de a written statement or explanation to the U.S. Nuclear 

:ommission, ATTN: Document Control Desk, Washington, D.C. 20555, 

to the Chief, Special Inspection Branch, Division of Inspection 

Programs, Office of Nuclear Reactor Regulation, within 30 days of 

the letter transmitting this Notice of Nonconformance. This reply 

early marked as a "Reply to a Notice of Nonconformance" and should 

include for each nonconformance: (1) a descriDtion of steps that have been or 

on of steps that have 

the dates your corrective 

eted. 

will be taken to correct these items; (2) a descript 

been or will be taken to prevent recurrence; and (3) 

actions and preventive measures were or will be comp 

Dated at _. Rockvi 11 e, Mary1 and 

this A 2 .&day of July 1997 

-6- 
Enclosure 3 

INSPECTION REPORT 

ORGANIZATION: 

REPORT NO.: 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY : 

INSPECTION CONDUCTED: 

U. S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF INSPECTION AND SUPPORT PROGRAMS 

SUBSEQUENT DISCUSSIONS: 

INSPECTORS: 

APPROVED BY: 

ABB Service Incorporated 

Cleveland Service Center 

5311 Commerce Parkway West 

Cleveland, Ohio 44130 

99901281/97-01 

Mr. D.E. Leckey 

Quality Assurance (QA) Manager 

(216) 267 2882 

Can provide new ABB switchgear, procurement of 

replacement parts, maintenance, refurbishment 

and on-si te swi tchgear services. 

May 14, 1997 

June 5 and July 14-22, 1997 

Kamalakar R. Naidu, NRR 

Virgil L. Beaston, NRR 

Joseph J. Petrosino, NRR 

Gregory C. Cwalina, Chief 

Vendor Inspection Section 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

-7- 
1 INSPECTION SUUUARY 

ABB Service Incorporated ABB Service), is headquartered at its Columbia, 

Maryland, Service Center. 

17 service center facilit es throughout the United States that could have 

potentially performed eng neering, consulting services, or other work on, or 

related to, safety-related circuit breakers for NRC licensees. However, due 

to quality concerns, ABB Service modified that policy and currently allows 

only four of its service centers to process safety-related circuit breaker 

work. They are: Cleveland, Ohio Service Center; Columbia, Maryland Service 

Center; Houston, Texas Service Center; and Charlotte, North Carolina Service 

Center. 

Prior to 1996, ABB Service had approximately 

The ABB Service Center located in Cleveland, Ohio, has been performing 

refurbishment services on metal-clad, 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 

1.T.E.- Gould, Gould-Brown Boveri, Brown Boveri Electric and finally ASEA 

Brown Boveri . AB6 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-re1 ated 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: 

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

Fuel Reprocessing Plants," to Part 50 of Title 10 of the Code of Federal 

Requlations (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. 

Service has a current organization chart which depicts the organization. 

The team determined that ABB 

2 

-8- 

I 
Unresolved Item 94-01-02 (Closed). 

job descriptions and qual ifications of persons performing safety-related 

activities were not available. The team identified that ABB Service has 

identified and delineated job descriptions and qualifications of persons 

performing safety-related activities. 

The previous inspection identified that 

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 94-01-038 (Closed). 

not all contact resistances were delineated for purchaser use. 

determined that maximum acceptable contact resistance values are now furnished 

in the ABB Service test instructions. 

The previous inspection determined that 

The team 

3 INSPECTION DETAILS 

3.1 K-line Breaker Refurbishment 

a. ScoDe 

The inspectors reviewed circumstances that led to incorrectly assembled ABB 

Service K-Line breakers being delivered to PNPP. 

inspection at A66 Service’s Cleveland facility, the inspectors: 

During and subsequent to the 

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

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

K-Line breakers, 

e examined the documentation related to the work performed by ABB 

Service on PNPP breakers to determine if it met the 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. 

3 

-9- 
I 

b. 

b.1 K-Line Breaker failure at PNPP 

On October 4, 1996, while PNPP was operating at 100 percent power, a new ABB 

K600S K-Line Breaker (Serial Number 934277-031295) supplying 480-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 two lead wires from one of the breaker'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. 

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. 

Crossing the phase sensor leads changed the 

Six doughnut-shaped current transformers (CTs) are mounted on the lower 

molding current transformer assembly of the K600S 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 

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. 

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. 

ends of the lower terminal blocks are connected to the appropriate terminals 

of the Power Shield trip unit using a multiconductor cable. 

Two 

The other 

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. 

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, 51817C264135 and 518170-211135). All three of these incorrectly assembled 

K-Line breakers had been refurbished and tested by ABB Service. 

The errant signal caused the trip unit to sense an abnormally high 

4 

-10- 
The inspectors determined that current and phase sensors (sensors) wiring and 

assembly errors may occur in any of the following manners: 

e 

e 

e 

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 the wires have been lifted to conduct K-Line Breaker tests. 

e Installing a sensor upside down or incorrectly wiring a sensor when the 

lower molding current transformer assembly of the K-Line Breaker is 

di sassembl ed (to rep1 ace a sensor). 

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 KLine breakers and that single phase testing methods typically used by the 

licensees and ABB Service organization would not identify the problem. The 

problem could affect safety and non-safety-related K-Line breakers. However, 

ABB did not take action to assure the circumstances of the problem were 

affect i vel y d i ssemi nated . 

b.2 Procurement Documents 

Purchase Order (P.O.) S137920, Rev. 3, dated June 14, 1996, controlled the 

work performed on safety related K-Line breakers including S/N 518170-211135. 

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, hand1 ing, 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. S137920 

required, in part, the following documentation to be submitted, as applicable: 

e 

ABB Service Certi f i cate of Compl i ance (CoC) . 

Procedural Checklist for Safety-Related Nuclear Switchgear (ABB 

Service Form QA2). 

Final Inspection/Acceptance Criteria Check1 ist (ABB Service Form QA6). 

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

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

part that: 

5 

-1 1- 
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 uti 1 izing factory approved design documents.. . Any observed 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 K3000S breaker. 

PNPP sent this breaker to ABB Service for an inspection, lubrication 

evaluation, 10-year refurbishment, and repair in accordance with P.O. No. 

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 Manager. 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 

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 of the breaker, and therefore there was no impact to plant 

operability. The line item "Potential Part 21 Evaluation Required?" on the 

NCR was checked "No" and this response was signed by both the Service Center 

Manager and the Service Center QA Manager. 

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 Check1 i st for Safety Re1 ated Swi tchgear," 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. 

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. 

dated November 14, 1996. 

The inspectors verified that these corrective 

Both revisions were 

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. 

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

6 

-12- 
b.4 K-Line Breaker S/N 51817C-264135 

K-Line Breaker S/N 51817C-264135 is a non-safety-re1 ated type ABB K600S K-Li ne 

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. 

ABB Job # 43-02902-3. In this NCR also, the line item "Potential Part 21 

Evaluation Required?" was checked "No" and this response was signed by the 

Service Center Manager. 

ABB Service Forms QA2 and QA6 were being utilized to prevent recurrence of 

future wiring errors. 

Based on Perry's information, ABB Service generated NCR 

The NCR stated that PNPP was assured that the revised 

From the available documentation, the inspectors could not determine who 

caused the wiring error. 

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 transformers were not rep1 aced for this breaker. 

The leads were not reversed at the Power Shield, as 

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-re1 ated 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 K600S 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. 

Required?" was checked "NO" and this response was signed by the Service Center 

QA Manager. 

In the NCR, the line item "Potential Part 21 Evaluation 

7 

-1 3- 
The inspectors determined from the available documentation that this 

nonconforming condition was caused by ABB Service during replacement of the 

lower molding current transformer assembly. The K-Line Breaker repair 

worksheet (ABB Service Form CBRW, Rev. 4., Dated March 10, 1995) for breaker 

S/N 51817D-211135 indicated that on April 2, 1997, ABB Service replaced the 

phase-C lower assembly which contained the upside-down power sensor. 

Service Form CBRW for breaker S/N 51817D-211135 was signed and approved by the 

Service Center QA Manager on April 15, 1997. 

The AB6 

On April 16, 1997, ABB Service issued a CoC for Breaker S/N 518170-211135, 

signed by the QA Manager, 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 51817D-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 AB6 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 

second example for Nonconformance 99901281/97-01-02}. 

c. Conclusions 

Based on the above, the inspectors determined that AB6 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. ScoDe 

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 Findinqs 

b. 1 Procedure 

The team reviewed QAP 15.2, Revision 2, to assess whether it effectively 

implemented the provisions of 10 CFR Part 21 and 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 

-14- 
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. 

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. A1 though the 

clarifications appeared 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 

Part 21 verbiage except for written interpretation by the NRC General Counsel. 

The team noted that several of the 

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

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 definitions are not in 

accordance with the current revision of 10 CFR Part 21. 

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. 

a "defect" or "failure to comply." 

Identification of the basic component or activity which contains 

8. 

activity which contains the "defect" or "failure to comply." 

Identification of the Company supplying the basic component or 

C. Nature of the "defect" or "failure to comply" and the safety 

hazard which is or could be created. 

D. Number and location of all affected components: (include 

identification of all purchasers to whom component has been 

suppl ied) 

E. Corrective action, identification of party responsible or 

action, and schedule for action. 

The team determined that these requirements may not be fully understood by the 

average employee/technician. 

have somewhat of a chilling effect. 

not typically have the expertise to identify whether a "defect" or "failure to 

comply" exists. 

not complete the form since the employee would be unable to provide the 

required information. 

identifying problems or deviations to their supervision to ensure that all 

potential defects are identified and dispositioned in accordance with 10 CFR 

Part 21. 

personnel, may have prevented deviations from being reported in the past. 

Additionally, the requirements could tend to 

The team believes that an employee would 

Therefore, the team felt it possible that an employee would 

The employee's responsibility should be limited to 

This form, if not understood completely by the ABB Service 

9 
After discussing the 10 CFR Part 21 regulation, intent and QAP 15.2, with ABB 

Service staff, the team informed ABB Service that it had determined that it 

had not established an adequate procedure to appropriately implement 

Part 21. 

and will be treated as a Non-Cited Violation, consistent with Section IV of 

the NRC Enforcement Policy. 

resol vi ng the NRC concerns. 

10 CFR 

This failure is characterized as a violation of minor significance 

The ABB Service Director of Quality committed to 

b. 2 PNPP Eva1 uat i ons 

The team reviewed licensee and ABB information regarding identified problems , 

with the K-Line low voltage circuit breakers. On three separate occasions, as 

discussed above, PNPP reported that ABB Service refurbished and 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 Required?" 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 phase calibration testing does not 

adequately verify the overcurrent trip set 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 

1 icensee 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 

518170-213135, 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 safetyrelated 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 

-16- 
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 unidentifiable 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 improper current sensor polarity wiring exists. 

0 It appears that the error in polarity may have existed prior to 

shipment from the AB8 manufacturing factory .... 

0 

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 sDecific. This would 

cause premature tripping .... 

The comparison of’ the field measurements and the circuit breaker 

limitations, with one phase reversed, clearly indicate that having 

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 should be 

identified and have the polarities of the current sensors checked 

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 Service 

employees into believing that they must perform a review of the circumstances 

11 

- 1 7- 
I 

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 Eva1 uat i ons 

The 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 

The 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 

potential problems that could have ensued at operating nuclear power plants as 

a result of the potentially generic K-Line series matter. 

4. PERSONS CONTACTED 

ABB Service CornDanv 

3.3. 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.O. Webb, Director of Quality 

Centerior Enerqv ComDany 

M.R. Fournier, Quality Engineer, PNPP Nuclear Assurance 

ITEMS OPENED, CLOSED, AND DISCUSSED 

99901281/97-01-01 NOV inadequate evaluation/failure to inform 

99901281/97-01-02 NON inadequate inspections 

C1 osed 

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 avai 1 ab1 e 

12 

-18- 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

August 29, 1997 

WASHINGTON, D.C. 20555-ooO1 

Mr. Wilfred C. LaRochelle, Manager, 

Quality Assurance 

Hartford Steam Boiler Inspection and Insurance Company 

One State Street 

P.O. Box 5024 

Hartford CT 061 03-31 02 

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 identified that the qualification files for two lead auditor 

candidates did not contain adequate documentation to support the point scores 

assigned 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 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. 

-19- 
2 

In accordance wi:h 10 CFR 2.790 of the NRC’s “F ules of Practice,” a copy of this letter 

and its enclosures will be placed in the NRC‘s Public Document Room. 

Since re1 y , 

ORIGINAL SIGNED BY GREGORY CWALINA FOR: 

Stuart A. Richards, Chief 

Special Inspection Branch 

Division of Inspection and Support Frograms 

Office of Nuclear Reactor Regulation 

Docket No:. 99900601 

Enclosures: 1. Notice of Nonconformance 

2. Inspection Report 99900601/97-01 

-20- 
NOTICE OF NONCONFORMANCE 

Hartford Steam Boiler Inspection 

and Insurance Company 

Hartford, CT 

Docket No.: 99900601 

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 

Acc red ita t ion. 

A. Criterion XVIII, “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 

verify all aspects of your quality assurance program and to determine the 

effectiveness of the program. 

Engineering Services Manual (ESM) Chapter 4400, “Audits,” states, in part, in 

Section 4421, that the Quality Assurance 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 assigned that responsibility. 

2. HSB did not perform the annual audits of RMES activities for San Francisco 

(1 996), Atlanta (1 9959, and NortheasVPhiladelphia (1994 and 1995). 

(Nonconformance 99900601 197-01-01) 

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 431 0, that 

this Chapter outlines the requirements for the identification, documentation, and 

disposition of nonconformances to the ESM, supporting procedures or 

in st ructions. 

Enclosure 1 

-21- 
Section 4340, states, in part, that the RMESIQAM 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 (1 996 Atlanta 

regional office audit) where the RMES used a nonconformity report to 

document and disposition the audit findings that were identified during the 

QAM's annual audit of RMES activities. (Nonconformance 99900601/97- 

0 1 -02) 

C. Sections 1-1 , "The Authorized Inspection Agency," Subsection 1-1.2, "Duties," 

Paragraph 1-1.2.4 of ASME QAI-1-1995, requires, in part, that the agency shall 

establish and implement an internal program which shall provide assurance that 

those of its employees holding the positions of supervisor or authorized nuclear 

inservice inspector (ANI I) 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 perform the annual audits of RMES activities, did not include any 

provisions for reviewing the disposition and corrective actions implemented 

for findings identified during past audits of RMES activities. 

2 

-22- 
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. 

National Board forms NB-71 and NB-178, "Audit Verification Record,'' are 

referenced in Sections 4471 and 4475 of the ESM and used by HSB as theA 

method to notify the National Board of completion of required audits, but are 

not included in Section HI "Forms," of the ESM. 

7 4. 

t 

(Nonconformance 

99900601/97-01-03) 

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

"Duties," Paragraphs 1-2.2.6 & 1-2.2.7 of ASME QAI-1-1995, require, in part, 

that the ANllS shall audit the performance of each ANI1 under his supervision on 

a planned and periodic basis. Each ANI1 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, 

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 ANII. 

2. The audits conducted by the Atlanta region RMES of the assigned ANI1 

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 Q*fi+h 

these inst ructions. 

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 NQA-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 
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. Similar provisions are contained in NQA-1. 

Contrary to the above, two lead auditor candidates were credited the maximum 

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-O1-05) 

Plesse provide a wtitten statemem or explanation to the U.S. Nuclear Regulatory 

Commission, ATTN: Document Control Desk Washington D.C. 20555, with a copy to 

the Chief, Special Inspection Branch, Division of Inspection and Support Programs, 

Office of Nuclear Reactor Regulation, within 30 days of the date of the letter 

transmitting this Notice of Nonconformance. This reply should be clearly marked as a 

"Reply to a Notice of Nonconformance" and should include for each nonconformance: 

(1) a description of steps that have been or will be taken to correct these items; (2) a 

description of steps that have been or will be taken to prevent recurrence; and (3) the 

dates your corrective actions and preventative measures were or will be completed. 

Dated at Rockville M ryland 

thisgbay of&&& 

4 

-24- 
Report No: 

Organization: 

Contact: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

9990060 1 /97-0 1 

Hartford Steam Boiler Inspection and Insurance Company 

One State Street 

Hartford, CT 06103-31 02 

Viiifred C. LaRochelle, Manager 

Corporate Quality Assurance 

Nuclear industry 

Activity: Authorized Nuclear Inspection Agency 

Dates: July 14 - 18, 1997 

Inspectors: 

Approved by: 

Uldis Potapovs, Senior Reactor Engineer 

Richard P. Mclntyre, Senior Reactor Engineer 

Larry L. Campbell, Reactor Engineer 

Gregory C. Cwalina, Chief 

Vendor lnsoection Section 

Special inspection Branch 

Division of Inspection and Support Programs 

Enclosure 2 

-25- 
1 INSPECTION SUMMARY 

During this inspection, the NRC inspectors reviewed the implementation of selected 

portions of Hartford Steam Boiler Inspection and Insurance Company's (HSB) quality 

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, 

CT and included a review of the corporate organization structure, 10 CFR Part 21 

implementation program, and HO responsibilities for audits, training and qualification, 

and nonconformity control. The second part of the inspection was conducted at the 

HSB regional office in Atlanta, GA and focused on 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 111, "Rules for Construction of Nuclear Power Plant Components" of the 

ASME Boiler and Pressure Vessel Code (Code) 

ASME QAI-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 

licensees. These nonconformances are discussed in Sections 3.3.2 and 3.3.3 of this 

report. 

2 STATUS OF PREL"3US 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 DescriDtion 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 

-26- 
ASME Section 111, Divisions 1 and 2, ASME Section XI, and ASME Sections I, IV, 

Vlll, 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, 

at 44 plant sites. 

3.2 10 CFR Part 21 Procram and its lmdementation 

HSB's procedures for implementing 10 CFR Part 21 regulations are described in 

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 supervi+m 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 (AN Is), 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 reportabie 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 the procedure were not consistent with the current revision of 

the regulation. 

(b) Although Appendix D defined a deviation 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 

-27- 
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 obligations (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, generated as a 

result of regional office or home office operations, had not been 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 shall 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 

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 significance 

and are treated as Non-Cited-Vi&tions, 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 Program and its Imdementation 

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 QAI-1 

series standards. Although the requirements of 10 CFR Part 50, Appendix B, 

are invoked through purchase orders (PO) of several licensees, the ESM does 

4 

-28- 
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 implementation 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 Inspection Responsibilities 

a. Inspection Scope 

The NRC inspectors reviewed ESM Chapter 4500, "ASME Inspection 

Responsibilities" and the authorized nuclear inservice inspector (ANII) bound 

diaries, inspection logs, and records of QA monitoring activities for selected 

nuclear plants 9 assess the program controls and their implementation. 

b. Observations and Findings 

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 ANlls and audits of their performance is the 

responsibility of the authorized nuclear inservice inspector supervisors (ANIS). 

Chapter 4500 also describes the duties and responsibilities of various 

categories of authorized inspectors consistent with the provisions of 

ASME -QAI-1. 

The implementation of ASME inspection activities was evaluated by reviewing 

;?:avant documentation (ANII bound diaries, monitoring schedules and reports, 

qualification records, etc.) for selected nuclear plants and, in some cases, 

discussing specific issues with the assigned ANI1 by telephone. The NRC 

inspectors reviewed the ANI1 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 

-29- 
The ANI1 bound diaries, in general, were found to contain appropriate entries, 

consistent with the applicable ASME QAI-1 requirements. Records of monitoring 

activities indicated that these activities were performed in accordance with the 

schedules established by the ANlls, using supplementary checklists (developed 

by the ANII), 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 ANll's bound diary and documented on HSB Form 939, "ES 

Record for Monitoring QNQC 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 

indicated that these activities were properly documented, and that identified 

concerns and their resolution were documented in the ANll's diary. 

A general observation was that activities described above were performed 2nd 

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 Monitoring QNQC Programs" 

was used for all identified concerns, while, at another plant site, the ANI1 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 QAI-1 requirements and that the ANI1 records of these 

activities were generally well documented and complete. One observation in 

this 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 Audits 

a. ScoDe 

The inspectors reviewed Chapter 4400, 'Audits,' of the ESM, which described 

the audit and survey requirements for internal audits, external audits, pre-review/ 

6 

-30- 
survey audits, and ASME Code required audits, and also described the 

accompanying requirements for audit report documentation. The following 

Sections of Chapter 4400 of the ESM were reviewed: 

0 Section 4420, "Internal Audits" 

0 Section 4430, "External Audits" 

0 Section 4470, "Nuclear Audits" 

b. Observations and Findinas 

b.1 Internal Audits - Home Office Encrineerina Services 

The Internal Audit Department (IAD) is responsible for conducting annual audits 

of HO Engineering Services (ES) department's ASME activities. The Quality 

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 QAI-1 standards, as applicable. 

The inspectors reviewed implementation of the above internal audit processes 

for compliance to the ESM. While attempting to review IAD's audits of HO ES 

activities, the inspectors determined that only one audit (November 1995) had 

been performed and documented by IAD 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 IAD 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 

RMESQAM 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 Internal Audits - Reaional Manaaer. Enaineerina Services 

All internal audits of the RMES activities are conducted by the QAM. Currently 

there are four regional offices that conduct ASME Section 111 and Section XI 

7 

-31- 
nuclear inspection activities. The inspectors reviewed the schedule for audits of 

RMES activities for the last four years to select a sample of audit reports to 

review. When reviewing 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 located in the HSB HO or Regional files for Atlanta, 

NortheastlPhiladelphia, and San Francisco in the 1993 to 1995 time frame. 

During the inspection the QAM committed to issue a Nonconformance Report to 

address this issue. The failure to audit all RMES activities on an annual basis 

as required by the ESM was identified as another example of Nonconformance 

99900601 197-01 -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 documented by either the QAM or by 

the RMES. The inspectors determined there was only one instance (1 996 

Atlanta regional oifice audit) where the RMES documented and dispositioned the 

audit findings on nonconformance reports. This issue was identified as another 

example of Nonconformance 99900601 /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 and was not 

referenced in ESM Section 4420 or in Appendix H, "Forms." During the 

inspection HSB committed to approve and control the checklist under the ESM 

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 indepth guidance and detail for conducting and documenting internal audits and 

accompanying audit findings and that there was limited documented QA program 

requirements for accomplishing this quality activity. No implementing procedure 

existed that addressed the internal audit process. This issue was identified as 

another example of Nonconformance 99900601 /97-01-03. 

8 

-32- 
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 findings 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 

Section 4470 of the ESM documents HSB requirements for the conduct of audits 

to comply with ASME QAI-1, 1995, ASME Section 111, Division 1 , and ASME 

Section XI. This included ANI and ANI1 audits of nuclear Section 111 and Section 

XI work activities, and ANI and ANI1 performance audits conducted by the 

authorized nuclear inspector and inservice inspector supervisors (AN IS and 

ANIIS). Chapter 4500 of the ESM, "ASME Inspection Responsibilities," 

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

and the ANIIS. 

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

"National ANI1 Assignments" listing for ASME Section 111 shops in the Atlanta 

region and Section XI plant sites (including Atlanta region). The ASME Section 

111 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 ANI1 performance 

audits are required to be performed twice a year and are conducted using the 

ES lnservice 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 ANI1 by the ANIS 

and the ANIIS. These forms are not included in Section H, "Forms," of the ESM. 

This issue was identified as another example of Nonconformance 99900601 1-7- 

01 -03. 

The inspectors reviewed a sample of semiannual inspector's performance audits 

conducted by the ANIS at Section 111 nuclear shops and by the ANllS at Section 

XI sites. The audit reports included the appropriate review 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 QAI-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 21 63 checklist described above, 

9 

-33- 
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 197-01 -04. 

During the above review, the inspectors also identified that the ESM does not 

include any requirement for the ANIS and the ANllS 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-O1-04. 

c. Conclusions 

Based on the above, the inspectors concluded that Hartford had failed to 

implement an adequate audit program. 

3.3.3 Qualificaticn of Lead Auditors 

a. ScoDe 

The NRC inspectors reviewed five HSB lead auditor qualification files to 

determine if the provisions contained in ESM Appendix C, "Qualification of Lead 

Auditors" were being 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 NQA-1- 

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 

-34- 

10 
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 

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 proces-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 

weakness in the HSB lead auditor qualification program. 

The NRC inspectors' review of the five HSB lead auditor qualification files 

identified 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 evidei ice that a 

minimum of 10 points under the score system provided in Appendix C have been 

accumulated. 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 NQA-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 

-35- 
C. 

3.4 

a. 

b. 

auditing disciplines. HSB presented a resume of one of the three lead auditors, 

HSB Identification Number 02562, which indicated for the period 1972-1 977, the 

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). 

Conclusions 

The NRC inspectors determined that the lead auditor files for HSB employees 

Identification Numbers 02026 and 02664, initially qualified as lead auditors on 

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

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

objective evidence that the provisions contained in Section 2.3.1.2 of ANSI 

N45.2.23 or Appendix 2A-3 of NQA-1 had been met. This issue was identified 

as Nonconformance 99900601 /97-01-05. 

Service Contract Provisions and Applicabilitv of 10 CFR Part 21 to HSB 

Activities 

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. 

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 BKI 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 UFI No. PTC00004, "Contract No. 

XT00000026 between Carolina Power & Light Company and the Hartford 

Steam Boiler Inspection and Insurance Co. ,I' hand dated April 5, 1990, 

12 

-36- 
stated that the work to be performed by HSB had been determined to be 

non-nuclear safety related. In Section 27, "1 OCFR21 ,'I the CP&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. 

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 and supervisory personnel to meet 

the requirements of ANSI N18.1 , and the AIA to meet the requirements of 

ASME 626.1-1 982. 

Duke Power Co. PO MN 12553, dated March 12, 1996 (for McGuire 

Nuclear Station) stated that the services to be supplied are safety related 

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(l)(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 

ths 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 21 10, "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) assigned to perform IWA 21 10 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 

-3 7- 
discussed that because the Inspector's performance of certain verifications and 

reviews 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 inspectors also determined that licenses 

were not performing QA program implementation audits on HSB as a provider of 

safety-related services. 

3.5 Entrance and exit meetinas 

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 their findings and concerns. 

14 

-3 8- 
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 

ITEMS OPENED 

ODened 

99900601 /97-01-01 NON Failure to perform required audits 

99900601 /97-01-02 NON Failure to issue nonconformity reports 

99900601 /97-01-03 NON Inadequate procedures 

99900601/97-01-04 NON Inadequate control of audits 

99900601 /97-01-05 NON Inadequate documentation of auditor qualification 

15 

-39- 
33NWQIOJN03NON 

A0 33ILON W I O-L6IOZf I0666 LXOd3tf N01133dSNI 3W :L33f8nS 
Ms. WiKieS -2- 

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 i ic’ude any personal privacy, proprietary, or 

safeguards information so that it can be placed in the P3R without redaction. However, if you 

find it necessary to include such information, you should clearly indicate the specific infomation 

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, 

Stuart A. Richards, Chief 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Docket No. 99901320 

Enclosures: 1. Notice of Nonconformance 

2. Inspection Report 9990 1320/97-0 1 

-41- 
NOTICE OF NONCONFORMANCE 

NUS Instruments, Inc. 

Idaho Fall;, Idaho 

Docket No. 9990 1320 

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 Intnunents, 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 of inadequate 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 of less than acceptable work quality in the wea of Quality Control 

inspection in July of 1996, for significant conditions adverse to quality. 

(Nonconformance 9990 1320/97-0 1-0 1) 

Please provide a written statement or expLation 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 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 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 hrther noncompliances, and (4) the date when your corrective action will be 

completed. Where good cause is shown, consideration will be given to extending the response 

time. 

-42- 

Dated at Rockville, Maryland 

this 263 day of Lptember 1997 n Enclosure I 
U. S. NUCLEAR REGULATORV COMMISSION 

Docket No: 

Report No: 

Facility: 

Location: 

Dates: 

Inspectors : 

9990 1320 

97-0 1 

hX'S Instruments, Inc. 

440 W. Broadway 

Idaho Falls, Idaho 83402 

August 19-21, 1997 

B. Rogers, Team Leader 

R. Pettis, Senior Reactor Engineer 

Approved by: G. Cwalina, Section Chief 

Vendor Inspection Section 

Special Inspection Branch 

Enclosure 2 

-43- 
1 

1.1 

1.2 

2 

3 

3.1 

a. 

Inspection Summary 

kkground 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: 

0 10 CFR Part 50, Appendix B 

0 10 CFR Part 21 

0 American National Standards Institute/American Society of Mechanical 

Engineers Standard N45.2.6- 1978, “Qualification of Inspection, Examination, 

and Testing Personnel for Nuclear Power Plants” as committed to in the NUS 

Quality Assurance Manual 

Nonconformances 

Nonconformance 9990 1320/97-0 1-0 1 was identified and is discussed in Sections 3.1 and 

3.3 of this report 

Status of Previous Inspection Findings 

No previous inspections have been conducted at this facility. 

Inspection Findings and Other Comments 

10 CFR Part 21 Program and Corrective Actions 

ScoDe 

The inspectors reviewed the NUS 10 CFR Part 21 implementing procedure, 10 CFR 

Part 2 1 Posting, and the NUS records related to 10 CFR Part 2 1 evaluations 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 

-44- 
b. Observations and Findings 

On August 8, 1996, Public Service Electric & Gas Co. (PSE&G) notified the NRC that 

numerous NUS Model OCA801 Signal Isolator Modules and Model MTHSO 1 Signal 

Summ:.for Modules, used in the Salem Nu( ‘ear Plant, Reactor Control and Protection 

System (RPS), were found have to unsoldert d or insuficiently soldered internal 

electrical connections. 

The documentation provided by PSE&G indicated that the isolator module function is to 

provide electrical separation between the Reactor Protection System (Class 1E) and the 

Process Control System (non-Class 1E) portions of the RPS instrument loops, and the 

summator module function is to algebraically combine analog signals. The reactor 

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 modu!e:’ ability to 

function during a design basis earthquake and could result in a loss of I .d~,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 ActiodRoot Cause Evaluation,” which was completed in 

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 person review team contracted by Scientech (NUS’S parent company) . 

IS Review 

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 selfverification practices, and differences in the soldering workmanship criteria of NUS and 

PSE&G. Corrective actions were specified to include (1) removal of the dual 

assignment of the manufacturing lead person, (2) increasing manufacturing supervision, 

(3) hiring an additional QC Inspector, and (4) providing supervisory skills training to 

supervisors. The performance of the NUS review had occurred just prior to the 

Scientech prchase of NUS (October 1996). Scientech and NUS management indicated 

in a telephone conversation with the NRC inspection team on August 26, 1997, 

subsequent to the completion of the inspection, that 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 usefbl 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 

-45- 
partial inspections were being performed in accordance with an agreement, between the 

manufacturing and inspection staff, to modify the operation process sheet flow. The 

Manufacturing Supervisor stated that there. had been some indication during the period 

of the PSE&G module production (approximately July 1995 to July 1996) that there had 

been sc me modification of activities occ' ir .ing at the staff level but that reorganization of 

the work activities and reassignment of staff 'lad 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 

activities and did not agree with this conclusion of the NUS review. The inspectors 

noted that modification of the operation process sheet flow, to alter the specified 

inspections, would be, as defined by the NUS Quality Assurance Manual (QAM), a 

Level 11 inspection functioti and that the QC Inspector performing the work at the time 

was certiiied to Lwel 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. Further, NUS had taken several covective 

actions which would affect this area to prevent any potential for reoccurrence on 

modification of activities that might have occurred at the staff level. 

Scientech 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 overview, personnel qualification, and work station adequacy. The 

recommendations made in the Scientech review included evaluation of client action 

requests, evaluation of work station ergonomics, increasing the frequency of trend 

analysis, improving the coordination of training activities, refining the wave soldering 

process, and improving -1ipervisor-y skills. Discussion with Scientech and NUS 

management 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 NUS to address the PSE&G 

soldering iswe. 

e Corrective Actiqg~ 

The inspectors reviewed the corrective actions taken by NUS, following the PSE&G 

notification, to address the inadequate soldering of the PSE&G Modules. The NUS 

Manufacturing Supervisor indicated that when PSE&G notified NUS in July of 1996 of 

the inadequate soldering, NUS had reviewed the manufacturing and inspection process 

and had taken numerous, immediate corrective actions. 
The Manufacturing Supervisor had reviewed the soldering process and had determined 

that the majority of the manufacturing line (referred to as “assemblers” by NUS) were 

soldering at the “preferred” level while &o assemblers were soldering at the 

“acceptable” level (a lesser quality level than “preferred”). While NUS had determined 

that the “acceptable” level was adequate fr: functicaing electrical connectior,; i: .!:! nct 

meet the Zustomer’s (PSE&G) expectatio ,. -411 of the assemblers were retrah-c’ in +he 

soldering technique by being provided physical examples of “preferred” soldering, 

studying the applicable procedures, and undergoing a supervisory assessment of each 

person’s soldering ability. In addition, the Manufacturing Supervisor assumed the 

manufacturing lead person’s responsibilities (that person was reassigned), 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 

retrained and the other was removed from the manufacturing activities and within one 

month of the PaE&C. notification the entire manufacturing line was solderrng at the 

“preferred” level. Tka inspectors noted that the requirement to solder at chhe “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 

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 than the PSWG 

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 inadequate soldering discovered 

in the PSE&G modules in accordance with the requirements of the NUS QA program 

and had not documented the immediate corrective actions previously discussed. 

-1 Corrective Act’ I-> 

In addition to the immediate corrective actions taken 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, txining the manufacturing supervisory and lead personnel, and 

upgrading the QC Inspector position to Level Il. 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 corrective actions had not yet been proceduralized at 

the time of the inspection. 

5 

-47- 
C. Conclusions 

The inspectors noted the indication that modification of the operation process sheets, to 

alter specified inspections, had occured in the July 1995 to July 1996 period, that such a 

rncdification v.rov’? have been i~ Lr- i, 11 mpection function, and that the QC Inspector 

perf3rrning the work at the time was ( ei ified to Level I. Indication of this modification 

highlighted a potential weakness in the NUS Quality Assurance program. 

The inspectors concluded that NUS had taken reasonable corrective actions in response 

to *e identification of inadequate soldering in the Salem modules. However, NUS had 

not adequately documented the corrective actions which NUS indicated were taken upon 

identification of the inadequate soldering. This was identified as an example of 

Nonconformance 9990 1320/97-0 1-0 1. 

3.2 Manufacturing Procedure and Implementation of Soldering Techninue 

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 and Findings 

The inspectors reviewed the training records for several assemblers to determine their 

qualifications to perform soldering. Training was documented on a “Certification of 

Training” which indicated that assembler was qualified 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 certificate 

documented that the assembler had received the required instruction, had passed a 

written examination, and corn; ‘eted a practical demonstration of soldering ability. 

The inspectors observed assemblers installing components, wrapping wire, and 

performing hand soldering. The inspectors also observed a demonstration by the QC 

Inspector on performance of an inspection of solder connections and discussed the levels 

of classification 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 “preferred” 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 

-48- 

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 such as verifying component placement, 

ver’fication of soldering, and final visual in pections. However, the inspectors cautioned 

NUS management that allowing manufactur ig and inspection staff to modify the 

operation process sheet flow without management approval, as discussed in detail in 

Section 3.1.2, would be a Level I1 inspection function and inappropriate if the QC 

Inspector performing the work at the time was certified to Level I or if such operation 

process modifications were disallowed by the NUS QA program. 

During review of the training files to verify QC Inspector qualifications, two documents 

were id-qtified which indicated a quality control inspection activity weakn .ss. Two 

memoranda from the QX Supervisor identified weaknesses in the QC Inspector’s 

performance ad one recommended management action. These memoranda, ir.:tiaLed in 

January and July of 1996, and were currently in the QC Inspector’s training file. 

Discussion with the current NUS and 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. However, during the period of time from 

July 1996 until December 1996, NUS had been made aware of the inadequate 

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, discussed their performance with the QA 

Supervisor and the Manufacturing Supervisor, reviewed work documentation, and 

observed work performance of the current QC Inspector. The inspectors did not observe 

any indication of inadequate performance in the documentation or work observations of 

the current QC Inspector. 

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 Inspector’s training file, NUS had not taken 

any correc:ive action for this situation adverse to quality until December of 1996. The 

failure to take corrective action to correct a significant condition adverse to quality was 

identified as an example of nonconformance 99901 329/97-0 1-0 1. 

8 

-50- 
4 Personnel Contacted 

Dama Wirries, President, NUS 

John McGimpsey, General Manager, NUS 

Cheryl Allen, QA Supervisor, NUS 

Shauna Royack, Manufacturing Supervisor, VUS 

Heath Buckland, Testing Supervisor, NUS 

Ron Todd, Quality Control Lead, NUS 

Paul Sturm, Principal Engineer, Scientech, Inc. 

9 

-51- 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. -1 

September 9, 1997 

Mr. Gregory M. Rueger 

Pacific Gas & Electric Conpany 

NPG-Mail Code AlOD 

PO Box 770000 

San Francisco, CA 94177 

Dear Mr. Rueger: 

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 Inspection Branch of the U.S. 

Nuclear Regulatcry Commission’s (NRC’s) Office of Nuclear Reactor Regulation 

(NRR) performed an inspection of Pacific Gas & 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 primary purpose of the inspection was to determine if design verification 

testi ng of the modi fi ed breaker was accompl i shed i n accordance with appl i cab1 e 

requirements of American National Standards Institute (ANSIVInstitute of 

Electrical and Electronic Engineers (IEEE) Standards. A secondary purpose of 

the i nspecti on was to exami ne several related issues involving production 

testing, modifications, and post installation testing in-service failures. 

The results of this inspection are contained in the enclosed inspection 

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 qf 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 

Commi ttee (Attachment 2 to the enclosed report), are discussed in the report, 

but in summary, the PG&E approach is considered consistent with the intent of 

the standards. A1 though there i s no regulatory guide endorsing the standards 

in question, PG&E has 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. With regard to the engineering evaluations necessary to 

support the PG&E approach, the inspectors could not, on the basis of 

docurnentati on avai 1 ab1 e for review at DCPP, conclusively determi ne the 

adequacy of a1 1 the justi fi cati ons for not reperforming certain desi gn tests. 

-52- 

I 
Mr. Gregory M. Rueger -2- 

Therefore. the inspectors identified the need for further review at National 

Technical Systems (NTS) , Inc. , Acton, Massachusetts : and possibly a1 so Power 

Distribution Services (PDS), Inc., Cincinrati , Ohio. 

In accordance with 10 CFR 2.790 of the Commission’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, 

Stuart A. Richards, Chief 

Special Inspection Branch 

Di vi si on of Inspecti on 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 

-53- 
Mr . Gregory M. Rueger -3- 

cc: 

NRC Resident Inspector 

Di ab1 o Canyor Nuclear Power P1 ant 

c/o U. S. Nuci ear Regulatory Commi ssi on 

P.O. Box 369 

Avi 1 a Beach, Cal i forni a 93424 

Dr . Richard Ferguson, Energy Chai r 

Sierra C1 ub Cal i forni a 

1100 11th Street, Suite 311 

Sacramento, Cal i forni a 95814 

Ms. Nancy Culver 

San Luis Obispc 

P.O. Box 164 

Pismo Beach, Cal i forni a 93448 

Mot hers ' for Peace 

Chai rman 

San Luis Obispo County Board of 

Room 370 P.O. Box 112 

County Government Center 

San Luis Obispo. California 93408 

Supervisors 

Mr. Truman Burns 

Mr. Robert Kinosian 

Cal i forni a Pub1 i c Uti 1 i ties Commission 

505 Van Ness, Room 4102 

San Franci sco, Cal i forni a 94102 

Mr. Steve Hsu 

Radi 01 ogi c Health Branch 

State Department of Health Services 

Post Office Box 942732 

Sacramento, Cali forni a 94232 

Diablo Canyon Independent Safety 

ATTN: Robert R. Wellington, Esq. 

Legal Counsel 

857 Cass Street, Suite D 

Monterey , Cal i forni a 93940 

Commi ttee 

-54- 

Regional Administrator, Region IV 

U. S. Plucl ear Regulatory Commi ssi on 

Harris Tower & Pavilion 

611 Ryan Plaza Drive, Suite 400 

Arl ington, Texas 76011-8064 

Christopher J. Warner, Esq. Pacific Gas & Electric Company 

Post Office Box 7442 

San Francisco, Cal i forni a 94120 

Mr. Robert P. Powers 

Vice President and Plant Manager 

Di ab1 o Canyon Nucl ear Power P1 ant 

P.O. Box 56 

Avi 1 a Beach, Cal i forni a 93424 

Tel egram-Tri bune 

ATTN: Managing Editor 

1321 Johnson Avenue 

San Luis Obispo. CA 93406 
Mr. Gregory M. Rueger -4- 

cc w/encl: 

NRC Resident Inspector 

Di ab1 o Canyor Nuclear Power P1 ant 

c/o U.S. Nuclear Regulatory Commission 

P.O. Box 369 

Avila Beach, California 93424 

Dr. Richard Ferguson, Energy Chai r 

Sierra Club California 

1100 11th Street, Suite 311 

Sacramento, Cal i forni a 95814 

Ms. Nancy Culver 

San Luis Obispo 

P.O. Box 164 

Pismo Beach, Cal i forni a 93448 

Mothers for Peace 

Chai rman 

San Luis Obispo County Board of 

Room 370 

County Government Center 

San Luis Obispo, California 93408 

Mr. Truman Burns 

Mr. Robert Kinosian 

California Pub1 ic Uti 1 i ties Commission 

505 Van Ness, Room 4102 

San Francisco, Cal i forni a 94102 

Supervisors 

Mr. Steve Hsu 

Radi 01 ogi c Health Branch 

State Department of Health Services 

Post Office Box 942732 

Sacramento, Cal i forni a 94232 

Di ab1 o Canyon Independent Safety 

ATTN: Robert R. Wellington, Esq. 

Legal Counsel 

857 Cass Street, Suite D 

Monterey, California 93940 

Committee 

Regional Administrator, Regi on IV 

U , S. F.!ucl ear Regulatory Commi ssi on 

Harri s Tower & Pavi 1 ion 

611 Ryan Plaza Drive, Suite 400 

Arlington, Texas 76011-8064 

Christopher J . Warner, Esq. 

Pacific Gas & Electric Company 

Post Office Box 7442 

San Francisco, Cal i forni a 94120 

Mr. Robert P. Powers 

Vice President and P1ar.r F?an,ger 

Di ab1 o Canyon Nuclear Power P1 ant 

P.O. Box 56 

Avi 1 a Beach, Cal i forni a 93424 

Chief, Fuel Cycle and Decommiss 

Branch, Region IV 

U.S. Nuclear Regulatory Commiss 

611 Ryan Plaza Drive, Suite 400 

Arl i ngton, Texas 76011-8064 

Mr. Peter H. Kaufman 

Deputy Attorney General 

State of Cal i fornia 

110 West A Street, Suite 700 

San Di ego, Cal i forni a 92101 

Mr. Thomas A. Moulia 

Humboldt Bay Plant Manager 

Humbol dt Bay Nuclear Power P1 ant 

1000 King Salmon Avenue 

Eureka. Cal i forni a 95503 

Chai rman, Humboldt County Board 

of Supervisors 

County Courthouse 

825 Fifth Street 

Eureka, Cal i forni a 95501 

oni ng 

on 

Redwood A1 1 i ance 

P.O. Box 293 

Arcata , Cal i forni a 95521 

-55- 
Docket Nos. : 

License Nos. : 

Report No. : 

Li censee : 

Faci 1 i ty : 

Locat i on : 

Dates : 

Inspectors : 

Approved by: 

U. S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

50-275, 50-323 

DPR-80, DPR-82 

50-275/97201, 50-323/97201 

Pacific Gas & Electric Company 

Diablo Canyon Power Plant, Units 1 and 2 

Avi 1 a Beach, Cal i forni a 

June 9-12, 1997 

Stephen D. Alexander, Team Leader, NRR 

Billy H. Rogers, Reactor Engineer, NRR 

Gregory C. Cwalina, Section Chief 

Vendor Inspection Section 

Speci a1 Inspecti on Branch 

Di vi si on of Inspecti on 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 

inspection of Pacific Gas & 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 Poww Plant (DCPP), Units 1 and 2. 

The inspectors reviewed engineering and qual i ty assurance documentati on, 

interviewed cognizant staff and examined equipment in order to evaluate PG&E’s 

dedication and modification of 4-Kv Yaskawa circuit breakers for use in safety 

related applications at DCPP. The primary purpose of the inspection was to 

determine if prot,otype 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 in 

question, PG&E had committed to them in the Final Safety Analysis Report 

(FSAR) for DCT?. and they are therefore relevant to the design and :icensing 

basis of the plant. 

A secondary purpose of the inspection was to examine several related issues 

i nvol vi ng product i on testi ng , modi f i cations , post i nstal 1 at i on test i ng , and 

in-service failures. The review included all documentation pertinent to the 

design, design verification (including prototype testing), conversion/modi - fication, fabrication of adapting hardware. production testing. procurement 

and dedication done by PG&E, its principal contractor, National Technical 

Systems (NTS) , Inc. , of Acton, Massachusetts, and the NTS subcontractor who 

actually performed the conversions, Power Distribution Services (PDS), Inc. , 

of West Chester, Ohio. 

The inspectors determined that PG&E’s approach to design verification was 

generally consistent with the applicable NRC regulations, primarily Criterion 

111, ”Design Control,” of 10 CFR Par 50, Appendix B, in that PG&E undertook to 

verify, by engineering analysis and a testing program, that DCPP’s safetyrelated 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 approach to the conversion and its 

design verification process with ab; icable 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 verification 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 ci rcui t current interrupti on, load switching and 

capacitance switching tests, need not be repeated. However, the letter 

2 

-57- 
stated, to apply this waiver, it must be shown that the mechanical operating characteri sti cs 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 

reperformi ng 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 

-58- 
Report Detai 1 s 

I I I. Ensi neeri nq 

El Conduct of Engineering 

E. 1 4-kV Swi+chsear Conversion Desiqn Verification 

a. InsDection Scope 

In order to verify that PG&E and its contractors, NTS and PDS, had properly 

veri fi ed the interrupting capacity of the 350-MVA Yaskawa SF6 interrupter 

breakers, adapted to fit into 250-MVA GE Magne-B1 ast swi tchgear . the 

inspectors reviewed engi neeri ng and qual i ty 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 

e 

e 

Whether the design verification testing conducted on the complete 

conversion (consisting of the Yaskawa "modular assembly" plus the 

hardware to adapt it to the Magne-Blast cubicle) for PG&E (conducted for 

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 assembl i es, 

Whether taking credit for the Yaskawa ANSI testing of the modular 

assembly supplemented by technical evaluations was consistent with 

Section 5 of C37.59, 

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 sti 11 valid 

after those factory modifications, 

Whether the technical evaluations by NTS to demonstrate that its 

additional testing of the complete conversion as requi red by C37.59 was 

not invalidated by modifications made in response to installation, setup 

and operational problems made subsequent to the tests, and 

Whether PG&E adequately resolved the findings identified i n i t s audits 

of NTS, PDS. and Yaskawa. 

Also examined were: 

The material of the breaker secondary disconnect pins and 

The potential overtravel situation involving the stationary auxiliary 

switch (SASI in the cubicle possibly preventing full closure of the 

breaker, and several other interface and operation issues brought to the 

VIS inspectors ' attenti on by personnel from DCPP Operati ons . 

4 

-59- 
The inspectors reviewed the PG&E documentation related to the design tests of 

the 5GYB-1-1200 AND 5GYB-1-2000 circuit breakers performed by Yaskawa, PDS. 

and PSM; the evaluations performed by NTS which demonstrated the equivalency 

of the circuit breakers supplied to PG&E to the prototype circuit breakers 

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 

ANSUIEEE testing. The insksctor'c r:?: J :vas performed to verify that the 

circuit breakers supplied to PG&E were equivalent to the circuit breakers 

tested by Yaskawa. PDS. 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 representatives of the circuit breaker which Yaskawa would subsequently sell cis 

commerci a1 grade products (non safety-related) , This ci rcui t breaker was the 

Yaskawa SF, 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 breakers as Class 1E (safety-related) equipment. 

NTS purchased breakers from Yaskawa and subsequently subcontracted their 

modification and additional testing to PDS and PSM, Inc. (a high-energy test 

facility in Pittsburgh) , to be performed under the QA coverage of NTS. 

The particular models of these breakers that underwent design verification 

testing were Yaskawa Models 5GYB-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, ANSUIEEE Standard C37.59-1991, these components 

constitute a "modular assembly. 

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 apparatus, stationary auxiliary switch. truck-operated cubicle switch, 

and cubicle interlock +vices. 

PDS adapted the modular assembl ies for 

The applicable NRC regulations were Appendix A. "General Design Criteria for 

Nuclear Power Plants," to Part 50 of Title 10 of the Code of Federa7 

Regu7ations (10 CFR Part 50, Appendix A); 10 CFR Part 50. Appendix B, "Quality 

Assurance Cri teri a for Nuclear Power P1 ants and Fuel Reprocessi ng P1 ants ' '* 

10 CFR 50.59, "Changes, Tests, and Experiments, 'I 10 CFR Part 21, "Reporting of 

Defects and Noncompliance." The 10 CFR Part 50. Appendix B. criteria that are 

especially applicable to this project are Criterion 111, "Design Control .'I and 

Criterion VI I, "Control of Purchased Materi a1 , Equi pment and Services . 'I 

The principal industry standards applicable to switchgear conversions are 

ANSUIEEE Standards C37.59-1991, "IEEE Standard Requi rernents for Conversion of 

Power Swi tchgear Equipment: I' C37.09-1979. "IEEE Standard Test Procedure for AC 

5 

-60- 
High-Voltage Circuit Breakers Rated on a Symmetrical Current Basis," and 

C37.04-1979. "IEEE Standard Rating Structure for AC High-Voltage Circuit 

Breakers Rated on a Symmetrical Current Basis." Although there is no 

regulatory guide endorsi ng the standards i n 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 per C37.09 

has been performed on a prototype(s1 of the rriodular assembly. e.g., by the 

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. 

PG&E further i nterpreted the standard to provide that i f engineering eval ua - tions of all xbsequent modifications can demonstrate that function. or 

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

adversely affectea by the modifications such that the original tests would be 

invalidated, then the C37.09 design tests of those functions or characteri stics need not be repeated on the complete conversion. 

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

inspectors), the original circuit breaker design tests had been performed by 

Yaskawa 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 

its components parts to meet its assigned ratings and operate satisfactorily 

under normal service conditions/special conditions defined by the PG&E 

speci f i cati on. NTS Report No. 60431 -95N-C, "Equi valency Eva1 uat i on of ANSI 

Type Tests and ANSI C37 Test Reports," Revision 7, dated June 4, 1997, 

documented the required ANSUIEEE tests performed, specifying a description of 

each test, which company had performed the test, and including applicable test 

report. 

Paragraph 5.1.4.2(2) of C37.59 then requires specific additional testing of 

the complete conversion, which included dielectric, momentary (C37.20.3-1987), 

conti nuous current, i nterl ock and 'xi 1 i ary functions , and mechani cal 

endurance testing (C37.06-1987). The inspectors confi rmed 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 experts, i ncl udi ng , most 

notably, Dr. Ward Laubach of the Low-Vol tage Swi tchgear Device Subcommittee of 

the IEEE Switchgear Committee of the IEEE Power Engineering Society, sponsor 

of ANSIIIEEE 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 (1) 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- 
project involving French Mer1 i n-Geri n breakers for the Quad Ci ti es and Dresden 

Nuclear Stations), and (2) an employee of Square D Company, the U.S. 

representative for Merlin-Gerin (both of which companies are owned by the 

Schneider 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. 

(Attachment 2 to this report 1, the Chai rman of the Hi gh-Vol tage Ci rcui t 

Breaker Subcommi ttee of the Swi tchgear 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 standarcl 

regardi ng requi r ?d design veri fi cati on testing for conversions using modul ar 

assembl i es . 

In a letter to the NRC. dated August 21, 1997 

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 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 

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 Equi Val ency Eva1 uat i on 

The inspectors found that the original Yaskawa breaker required extensive and 

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 establish 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 cri ti cal to determi ni ng that the ci rcui t breakers suppl i ed 

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 ANSUIEEE 

testing was performed by approved vendors (or surveilled) and performed in a 

cal i brated test faci 1 i ty , and materi a1 s on the suppl ied circuit breakers were 

appropriately dedicated. In addition, NTS reviewed all design changes made by 

Yaskawa and PDS and certified that the modification did not invalidate the 

ANSI/IEEE testing 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 type tests had been 

performed. to the time of the NTS purchase. The Yaskawa changes 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 ANSIIIEEE 

type tests which had been performed by Yaskawa in 1993, and provided 

certification to that effect with the circuit breakers shipped to ,"TS. 

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 

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 

instal led 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 modi f i cati ons were pri mari ly mechanical changes to the ci rcui t 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-Vol tage Faul t 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 

complete conversion to adapt it to the auxiliary switches and mechanical 

interlocks of the Magne-Blast cubicle. These effects might cause the added 

hardware to impede breaker tripping on a fault: effects that may 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 complete 

conversion unit that was a1 ready undergoing the testing requi red by Paragraph 

5.1.4.2(2) of ANSUIEEE Std C37.59 in July 1996 at PSM. However, due to some 

problem with or unavailability of PSM's main high-current test facility 

8 

-63- 
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). 

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 con<::!? , ~ls only the current,, not the voltage 

gives rise to the magnetic fields. 

PG&E argued, that even though the 

C. Concl us i on 

The inspectors concluded that PG&E's approach to design verification was 

generally consistent with the applicable NRC regulations, primarily Criterion 

111, "Design Control ," of 10 CFR Par 50, Appendix 6. in that PG&E undertook to 

verify by engineering analysis and a testing program that DCPP's safetyrelated 4-kV electrical distribution system as converted and modified would 

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 from 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 performed primarily by NTS 

demonstrated that the modification made by Yaskawa and PDS had not impacted 

the validity of the ANSIIIEEE tests performed on the circuit breakers: 

therefore, the inspectors could not concl usi vel y determine the adequacy of a1 1 

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 Instal 1 ation and Performance Concerns 

a. InsDection 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 contact pin material, (2) breaker operation interference due to 

stationary auxiliary switch (SA9 overtravel, and (3) SAS adjustment/ 

performance, the inspectors reviewed the associ ated DCPP Action Requests ( ARs) 

and thei r di sposi ti ons , interviewed cogni zant engi neeri ng and operations 

personnel , and examined affected components. 

b. Observat i ons 

9 

-64- 
b. 1 Secondary Disconnect Pi n Materi a1 

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 addi Licn, 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 vertical 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 relnoval to ensure that the four segments of each pin we72 properly spread for adequate electrical contact when the breaker was fully racked up into its operate position. The representative from DCPP operations who had 

also related this concern to the team was satisfied that the corrective action 

was adequate. PG&E stated that there were sufficient replacement pins of the 

improved. stifferhore tempered material on site to replace any pins that 

should become i rreparably deformed despite improved hand1 i ng 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 experimentation), the 

inspectors determined that PG&E's minimum required gap (0.040") between the 

stationary auxi 1 iary switch operating rod and 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 procedures that required checking and establishing this gap each time a 

breaker was racked in, would prevent the overtravel condition from occurring. 

b.3 SAS Adjustment and Performance 

In addition, the DCPP operations representative had related concerns to the 

inspectors regarding the several other interface problems that had been 

encountered relating to the SASs. Having suffered several equipment fai lures 

since the instal lation of the converted Yaskawa breakers attributable to 

problems with the SASs (all sets of contacts not always changing state with 

breaker 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. breaker- open, plunger-SAS operating rod gap that would ensure that all SAS 

contacts would change state when the breaker closed given the worst case 

(largest) required stroke of the all the SB-12 switches. 

PG&E then determined the maximum allowable 

10 

-65- 
During thi s peri od, another re1 ated problem presented itself . Upon 

investigating the failure of a pump to start, PG&E found that another instance 

of 4-kV breaker-SAS adjustment to be 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 investigai.ion, PG&E discovered that wt-x~ the adjustment screw at the top 

of the breaker’s SAS plunger is retorqued after re lacing adjustment shims, 

the linkages expand to their maximum 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 procedures and conducted training 

to ensure that the plunger is tapped down into its fully withdrawn, 

breaker-open re;: basition before setting the plunger-to-SAS operating rod gap 

by manually adjusting the breaker elevation in the cubicle. 

c. Concl usi on 

With respect to the two 4-kV breaker installation, interface, and performance 

concerns, the inspectors determined that PG&E’s corrective action was 

appropri ate and adequate. 

the linkages that operate the SAS plunger become s Y ightly cocked as joints in 

11 

-66- 
PARTIAL LIST OF PERSONS CONTACTED 

Licensee 

Shawn LaForce 

Brad Olson 

Don Allen 

Stan Ketel sen 

Bill Bayne 

Bob Whi tgell 

Char1 i e Ni chol s 

Michael Jacobson 

Chuck Lewis 

Bill Colry 

Tom Bennett 

Dave Taggart 

Thomas W. Packy 

K1 emme Herman 

Ed Kahler 

Tom Fetterman 

Pat Colbert 

Eric Nelson 

David Oatley 

Jim Molden 

Terry Grebe1 

RS 

USNRC 

USNRC 

Proc. Svs 

NQS 

Materi a1 s 

NQS 

NQS 

NSAL - RS 

Reg. Svcs. 

os 

NQS-Engi neeri ng 

NQS -PA 

NTS/DES/EE 

NTYTec h 

NTS/ES 

NTS/ ES 

Mechanical Mai nt . Maint. Serv. 

Opera t i ons 

Reg. Services 

Dir. & Procurement 

Engi neer 

Project Engineer 

Resident Inspector 

supv. 

supv. 

supv. 

Di rector 

Sr. Engineer 

Engineer 

Engineer 

Di rector 

Di rector 

Lead Auditor 

Elect. Sup. 

PM 

Di rector 

Elect. Sup. 

General Foreman 

Manager 

Manager 

Director 

Open Items 

This report categorizes the inspection fi ndi ngs as unresolved i tems 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 

Office of Nuclear Reactor Regulation will issue any enforcement action 

resulting from their review of the identified unresolved items. An inspection 

follow-up item (IF11 is a matter that requires further inspection because of a 

potential probiem. because specific licensee or NRC action is pending, or 

because additional information is needed that was not available at the time of 

the inspection. 

Item Number Finding 

50-275; 50-323/97201-01 URI 

Title 

Meeting ANSI/IEEE C37.59-1991 (Section 5) 

12 

-67- 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, O.C. #)5654001 

July 3, 1997 

Mr. D. M. Smith, President 

PECO Nuclear 

PECO Energy Company 

Nuclear Group Headquarters 

Correspondence Control Desk 

P.O. Box 195 

Wayne, Pennsylvania 19087-01 95 

. Dear Mr. Smith: 

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. 

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 111 items from material suppliers without the complete documentation 

required by Paragraphs NCA-3861 (b) and NCA-3862.1 (b) of Subsection NCA, Section 111 

of the ASME Code. 

In addition, the staff identified a weakness regarding PECO Nuclear Quality Assurance 

(NQA) 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 8udit report 8s the basis for maintaining ACCUTECH as an approved 

vendor on the PECO Evaluated Vendors List (EVL). The NQA 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 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. 

-68- 
Mr. D.M. Smith -2- 

You are requested to respond to the weakness 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 identifying, 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 response 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 2.790 of the Commission’s regulations, a copy of this letter 

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. Should you have any questions concerning the attached 

inspection report, please contact the inspection team leader Mr. Richard P. Mclntyre at 

(301) 41 5-321 5. 

Sincerely, 

?w= Robert M. Gallo, Chief 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Docket Nos. 50-352 and 50-353 

License Nos.: NPF-39 and NPF-85 

Enclosure: inspection Report No. 50-352/96-201 

and 50-353/96-201 

cc: See next page 

-69- 
PECO Energy Company 

cc: 

J. W. Durham, Sr., Esquire 

Sr. V.P. & General Counsel 

PECO Energy Company 

2301 Market Street 

Philadelphia, PA 1 9 1 0 1 

Manager-Limerick Licensing, 62A-1 

PECO Energy Company 

965 Chesterbrook Boulevard 

Wayne, PA 19087-5691 

Mr. Walter G. MacFariand, Vice President 

Limerick Generating Station 

Post Office Box A 

Sanatoga, PA 19464 

Plant Manager 

Limerick Generating Station 

P.O. Box A 

Sanatoga, PA 19464 

Regional AdminiStfatOr, Region I 

US. Nuclear Regulatory Commission 

475 Allendale Road 

King of Prussia, PA 19406 

Senior Resident Inspector 

U.S. Nuclear Regulatory Commission 

Limerick Generating Station 

P.O. Box 596 

Pottstown, PA 19464 

Director-Site Support Services 

Limerick Generating Station 

P.O. Box A 

Sanatoga, PA 19464 

Chairman 

Board of Supervisors 

646 West Ridge Pike 

Linfield, PA 19468 

of Limerick Township 

-70- 

Limerick Generating Station, 

Units 1 & 2 

Chief-Division of Nuclear Saferty 

PA Dept. of Environmental Resources 

P.O. Box 8469 

Harrisburg, PA 17105-8469 

Director-Site Engineering 

Limerick Generating Station 

P.O. Box A 

Sanatoga, PA 19464 

Manager-Experience Assessment 

Limerick Generating Station 

P.O. Box A 

Sanatoga, PA 19464 

Library 

U.S. Nuclear Regulatory Commission 

Region I 

475 Atlendale Road 

King of Prussia, PA 19406 

Senior Manager-Operations 

Limerick Generating Station 

P.O. Box A 

Sanatoga, PA 19464 

Or. Judith Johnsrud 

National Energy Committee 

Sierra Club 

433 Orlendo Avenue 

State College, PA 16803 

Mr. George A. Hunger, Jr. 

Director-Licensing, MC 62A-1 

PECO Energy Company 

Nuclear Group Headquarters 

Correspondence Control Desk 

P.O. Box No. 195 

Wayne, PA 19087-01 95 
Docket Nos.: 

License Nos.: 

Report Nos.: 

Licensee: 

Facility Name: 

Location: 

Dates: 

Inspectors: 

Approved by: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

50-352 and 50-353 

NPF-39 and NPF-85 

50-352/96-201 and 50-353/96-201 

PECO Energy Company 

Limerick Generating Station, Units 1 and 2 

Correspondence Control Desk 

P.O. Box 195 

Wayne, Pa 19087-01 95 

August 5-98 1996 

March 10-14, 1997 

Richard P. Mclntyre, Team Leader, NRR 

Uldis Potapovs, Senior Reactor Engineer, NRR 

Bill Rogers, Reactor Engineer, NRR 

Gregory C. Cwalina, Section Chief 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

-71- 
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 procurement of 

products and services used in safety-related applications at the Limerick Generating 

Station, Units I and 2 US). 

The inspection team reviewed PECO's procurement program to assess its complience with 

the quality assurance (QA) requirements of Appendix E 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) N45.2.13, "Quality Assurance Requirements for the 

Control of Procurement of Items and Services for Nuclear Power Plants," and 

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 or verification of suitability of those 

items for their intended or approved safety-related applications. The inspection included a 

review of procedures and representative records, including approximately 40 procurement 

packages for mechanical, material and electrical items; interviews with PECO staff, 

including senior management and LGS site personnel; 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 concerned the 

PECO Nuclear Quality Assurance (NQA) 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 

i 

-72- 
List (EVL). The NQA 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 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. PECO did not identify that ACCUTECH's 

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 

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 Nonconformance identified during the December 1 994 inspection . 

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

Section 111 items from material suppliers without the complete documentation required by 

Paragraphs NCA-386 1 (b) and NCA-3862.1 (b) of Section Ill of the ASME Code. 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 certified 

material test report (CMTR). In several instances the approved supplier certifications were 

not furnished with the CMTR and were not included in the document package. 

ii 

-73- 
j3eDort Details 

II ~ I. Enaineering 

E7 

E7.1 Evaluated Vendo rs List 

Ouality Assurance in Engineering Acdvides 

I. Dection SCQPB 

The inspectors reviewed PECO procedure P-C-9, "Evaluated Vendors List," Revision 

1, dated December 15, 1993, and representative documentation to verify 

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

provided guidance for the preparation and maintenance of the PECO Evaluated 

Vendors List (EVL), the EVL Conditional Clauses, and the Alert List. The inspectors 

reviewed PECO procedure NQA-19, "NQA Evaluation of Vendors," Revision 4, 

dated February 20, 1996, and Revision 5, dated August 5, 1996, and 

representative documentation to verify implementation. NQA-19 established 

requirements, assigned responsibilities and provided guidance for the PECO Nuclear 

Quality Assurance (NQA) evaluation of vendors for acceptance on the PECO EVL. 

b. Dbservations and Findinas 

The EVL was a listing of vendbrs which PECO had evaluated to determine their 

capabilities of supplying equipment, components, and services, in accordance with 

the conditions df a purchase order or contract. NQA-19 indicated that PECO 

evaluated vendors by several methods including assessments, commercial grade 

surveys, QA Manual reviews, and annual evaluations. Evaluated vendors were 

listed on the PECO EVL as Approved (A), Conditionally Approved (C), Commercial 

Grade (GI, or Bidder (B). Approved vendors had a quality program complying with 

10 CFR Part 50, Appendix B or another applicable standard and PECO had verified 

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

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 

vendor evaluation activities. The applicable "C" vendor quality assurance 

conditional clauses were documented on the PECO EVL and imposed on PECO POs 

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 for 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 

obtained satisfactory 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 

Bidders (B), NQA reviewed the QA program documents to determine that the 

vendor had the capability to meet 10 CFR Part 50, Appendix B quality 

requirements. The bidder review process was documented on the PECO Vendor 

QA Manual Checklist. 

1 

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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 limited scope 10 CFR Part 50 Appendix B program or 

who required additional requirements on POs based on audit, surveillance, or survey 

results, but were capable of supplying safety-related items or services. The Vendor 

Alert List identified areas to be reviewed concerning equipment and services and 

associated vendors based on information obtained from NRC Bulletins and 

Information 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 determined 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 Re view o f PECO Vendor Assess ment and Su rveilla nce Process 

a. lnsoect ion Scooa 

The inspectors reviewed PECO procedure NQA-20, "NQA Vendor Asseqsments and 

Surveillance," Revision 7, dated March 8, 1996, and representative documentation 

to verify implementation. NQA-20 established requirements, assigned 

responsibilities and provided guidance for the coordination, preparation, 

performance, and reporting of PECO Nuclear Quality Assurance (NQAI vendor 

assessments and surveillance. 

b. Pbservat ions and Findinaa 

NQA-20 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 vendor, 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 assessment 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. NQA-20 defined 

"assessment" as a process equal to or exceeding the requirements of "audit" as 

defined in ANSI N45.2. 

2 

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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 Vendor QA documents, 

PECO procurement documents, the PECO Vendor Alert list, the INPO Nuclear 

Network, NRC Vendor Inspection Reports, previous assessment results, EVL 

corrective actions requirements, and nonconformance data. The Assessment Plan 

was to be performance based using methods such as comparing performance 

objectives and acceptance criteria to the final result and observation of ongoing 

activities. In addition, NQA-20 indicated that the NUPIC Performance Based 

Supplier Audit Checklist should be used in developing the Assessment Plan when 

appropriate. The objective in development of the Assessment Plan was to 

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, 

and 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 documented on Vendor Corrective Action Requests 

(VCR) and if the nonconformance could potentially affect hardware the ATL would 

request that the vendor document the identified nonconformance on a vendor 

Nonconformance 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 Deviation 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, strengths and weaknesses, recommendations, and 

evaluation of activities), Investigation Results (activity investigated, summary of 

acceptable and unacceptable results, reference to VCRs and recommendations 

(REC)), and Administrative Details (names of PECO personnel involved, vendor 

personnel contacted during the assessment, and VCRs and RECs). 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 Assessment 

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 performing 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 and 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 "acceptable as is" or "repair" in accordance with the vendor's QA program 

3 

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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 inciuding the PO item number, quantity, description, and PECO nuclear 

code number for the item being released. A copy of the Certificate of Surveillance 

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 concluded that PECO had developed adequate procedural 

requirements to establish a program to effectively 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. 

E7.3 Be view o f Selected PECO Assess ments and Surveillance 

a. JnsDection SCODQ 

The inspectors review of a listing of PECO performed assessments indicated that 

PECO typically performed twelve assessments yearly. A large portion of vendor 

placement 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 IngersollRand. 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 Findinas 

b.1 ACCUTEC H 

The inspectors reviewed the Comanche Peak Steam Electric Station TU Electric 

(TUE) OA Audit Report QAA-96-010 of ACCUTECH dated May 23, 1996. The 

audit, conducted April 29 through May 2, 1996, was led by TUE and performed in 

accordance with the requirements of the TUE QA program, under the auspices of 

NUPtC. The audit also included representatives of Iowa Electric Services (IES), 

Houston Light and Power, and Northern States Power. The audit was performed to 

assess ACCUTECH's QA program and its implementation in supplying ASME code 

and noncode materials to the applicable requirements of 10 CFR Part 50, Appendix 

B, ASME NQA-1, and ASME Section 111, Subsection NCA 3800. The audit included 

the areas of Order Entry; Commercial Grade Dedication; Procurement; Material 

Controll Handling, Storage, and Shipping; Fabrication, Assembly, and Special 

Processes; Tests and Inspection; Calibration; Document ControllProcedure 

Adequacy; OrganirationlProgram Compliance; Nonconforming Conditions; 

Corrective Actions; Internal and External audits; and Traininglcertification. 

4 

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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 B, 

NQA-45.2 and NQA-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 

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. 

The inspectors reviewed the applicable sections in the TUE audit report that 

addressed the nonconformances and issues identified in the December 1994 NRC 

Inspection (Report No. 99901 076/94-01) at Cardinal Industrial Products (former 

name of ACCUTECH), concerning commercial grade dedication, and sampling 

relating to commercial grade item (CGI) dedication. The audit 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 NUPIC 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 nonapproved suppliers. This method, in turn forms the basis for ACCUTECH's 

selection of the CGI sampling plans. 

FECO's review and acceptance of the NUPIC 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 1 994 

and they were still implementing 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 program for the products reviewed were deemed to be adequate. The 

inspectors stated that the TUE audit report described the NRC findings and followup 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 December 1994. In their formal AARF review conducted June 

27, 1996, PECO did not identify that ACCUTECH's sampling process needed 

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 (CAI 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 

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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 grade 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 requesting 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 NQA for review were either downgraded or removed 

from the EVL. 

During the March 1997 inspection at PECO, the inspectors discussed concerns 

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. PECO agreed that 

the CA requirements could be clarified and did so prior to the completion of the 

inspection. 

b. 1.1 Vendor InsDect ion at ACCUTFCH. 

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 99901 307/96-01, dated March 4, 1997. During the November 

1996 inspection at ACCUTECH the inspectors reviewed the sample plan 

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 QA program 

that supported the sampling process rationale described in various 1995 

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. 99901 076/94-01. 

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 examination. This document was dated 

November 7,1996, but had not as yet been implemented as part of the 

ACCUTECH QA program. 

6 

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In conclusion, the inspectors determined that ACCUTECH placed heavy reliance on 

visual and dimensional inspection to support the verification of lot homogeneity. 

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 EPRl guidelines for nondestructive 

testing (dimensional). The inspectors determined that visual inspection for shipping 

damage and manufacturing defects can not assure that all items 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 

99901 076/94-01 and related correspondence, the use of ASTM A-325 shipping lot 

sampling plan is inappropriate for this application. 

b.2 Praaon Valves 

The inspectors reviewed the Illinois Power Company (IPC) Audit Report of Dragon 

Valves, Inc. (Dragon), dated October 27, 1995, which documented the September 

25-28, 1995, nuclear assessment of Dragon. PECO had performed 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 performed the audit as a NUPIC Joint Utility Audit with 

Commonwealth Edison, Entergy Operations, and Illinois Power providing audit team 

members. The audit was performed to assess Dragon's 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 111, 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; Cali bration; 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 were not 

considered significant enough to have any major impact on work previously 

performed by Dragon. When reviewing the NUPIC audit package the inspectors 

noted that the audit identified that Dragon does not audit vendors who hold current 

ASME Quality System Cenificates (QSCs) and also that Dragon does not perform 

any commercial grade dedication. It further stated that CMTRs for material 

received from unapproved suppliers and QSC holders is validated through receipt 

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 IPC 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 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 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 placement of Dragon on the PECO EVL with 

7 

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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 several instances where 

activities were not conducted in accordance with NRC and ASME code 

requirements. These included: (1 1 failure to provide adequate control of material 

procurement procedures and to prepare the Unqualified Source Material list; (2) 

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 AS TA Enaineerina. 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 

GilbertlCommonwealth, Inc., personnel contracted by PECO to perform the 

assessment and prepare the Assessment Report. 

The assessment was performed to address ASTA's QA program and its 

implementation as applicable to engineering services 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 (1 0 CFR Part 21 I, imposed by 

customer PO, to a subtier 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 

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b.4 

b.5 

PJ amco Controls Co r DO ri at on 

The inspectors reviewed the IES Utilities (IES) Audit Report of Namco Controls 

Corporation (Namco), dated January 25, 1996, which documented the 

January 8-12, 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 Illinois 

Power, Tennessee Valley Authority, and IES providing audit team members. The 

audit was performed to assess Namco's QA program and its implementation in 

supplying electromechanical limit switches, limit switch receptacle assemblies, 

plug-in connector assemblies, and cable assemblies. The audit included the areas 

of Order Entry; Design; Procurement; Material Control, Handling, Storage, and 

Shipping; Fabrication, Assembly, and Special Processes; Tests and InsRection; 

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 OA program did not specify the frequency of 

internal 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 audit 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 NQA-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. 

Sorrento F lectronics 

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 

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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; 

Software Quality Assurance; Procurement; Material Control, Handling, Storage, and 

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

Calibration; Document Control; and Program Compliance. 

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 to 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. 

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 performed 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 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 are adequate for the PECO application. PECO had concluded that the Wolf 

Creek audit was had been applicable to PECO's purchases, that the findings had 

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 POs 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 ACCUTECY 

The inspectors concluded that the NQA third party 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 

requirements and did not question the basis for verification of lot homogeneity for 

10 

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c.2 

c.3 

c.4 

c.5 

E7.4 

a. 

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 Industriel Products (previous name of ACCUTECH). 

identified as a program weakness. 

PT~OOT~ Valves 

This issue was 

The inspectors concluded, based on the documentation reviewed, 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 inspectors 

verified during review of PECO POs to Dragon, that PECO had placed the 

appropriate conditional clause requirements in the POs. 

4STA Enaineerina. 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. 

Namco Co ntrols CorDoration 

The ikpectors concluded, based on /he documsntation reviewed, that PECO had 

taken adequate actions to place Namco on the PECO EVL as an approved supplier 

of nuclear limit switches and spare and replacement parts and had adequately 

documented these actions. 

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. 

P 1s 

JnsDection SCQPB 

The inspectors reviewed PECO procedure P-C-1, "Material Requisition Process," 

Revision 3, dated August 27, 1995, and representative documentation to 

determine the process used to requisition safety-related material from other utilities. 

11 

-84- 
P* ervations and Findim 

P-C-1 established purchase classification requirements and stated in Section 4.1 6 

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 

purchasing safety-related items from other Utilities. During the first phsse of the 

inspection PECO stated that they do not audit other nuclear utilities for placement 

on the EVL, based upon the fact that they consider their OA programs as approved 

by the NRC by acceptance of Chapter 17 of their FSARNSAR 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 purchase requirements 

prior to shipment. 

C. Conclusions 

The inspectors concluded that this method was consistent with 10 Cf R part 50, 

Appendix B requirements. 

E7.5 Cont rol of Purchased Mate rid 

a. ction SCOD~ 

The inspection scope included a review 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 adequacy and completeness of 

vendor-supplied documentation. The control of these activities is described in 

Section 1 7.2.7, "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. 

A representative list of the document packages reviewed and inspector's 

observations are included in the following paragraphs. 

12 

-85- 
b.1 PECO PO LS 60591 9, dated January 1, 1996, to ACCUTECH Division of B&G 

Manufacturing Co. Inc. 

Item 1 of this PO was for 44, all thread 1-1 /4 inch by 10 inches long, ASME SA193 Grade 87 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 material included ACCUTECH’s certification 

that the material had been manufactured in accordance with their ASME Quality 

Systems Certificate (QSC). The certification included ladle 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 visual inspection (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 

QSC, additional documentation should have been provided by the Material 

Organization (ACCUTECH) and maintained at the plant site. Specifically, Paragraph 

NCA-3862.1 (b) states that, when the required chemical analysis (including the 

melting mill heat analysis), heat treatment, tests, examinations, or repairs are 

subcontracted, the approved supplier’s certification for the operations performed 

shall be furnished as an identified attachment to the certified material test report 

(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 fall into that category. ACCUTECH‘s CMTR did 

not include the melting mill certification, including heat analysis, as an identified 

attachment. 

Failure to verify conformance with procurement document requirements was 

identified as an example of Unresolved Item 50-352;50-353196-201-0’1. 

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 controlied under a quality program which 

satisfies the requirements of Section NCA-3700/3800 of Section 111 of the ASME 

Code. 

For Items 1 through 3, PEC0,s document packages included ACCUTECH’s 

certification that this material had been manufactured in accordance with their 

13 

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QSC. The certification included ladle 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 B requirements, a 

statement that macroetch test results were acceptable, and a statement that 

ACCUTECH had conducted a satisfactory visual inspection (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 

documentation was not provided by the Material Organization and maintained at 

the plant site. The melting mill certifications, including heat analyses and, for 

item 4, macroetch certification were not provided with the material. 

Failure to verify conformance with procurement document requirements was 

identified as an example of Unresolved Item 50-352;50-353/96-201-01. 

b.3 PECO PO LS 157145, dated November 26, 1996, to Energy & Process 

Corporation (E&P). 

Item 1 of this PO was for four 3 inch, NPS, buttweld, schedule 40, seamless SA234, Grade WPB elbows, to be provided in accordance with ASME Code, Section 

111, Class 2. 

The document package for this material included E&P Material Certification which 

certified compliance with the ASME Code and PO requirements and noted that the 

material was supplied in accordance with E&P's ASME QSC. Included with this 

certification was a Material Test Report from Ladish Company, dated March 238 

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 

31 6L couplings, to be supplied in accordance with ASME Code, Section 111, 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, 19928 

and 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 1118 Class 2. 

In addition to E&P Material Certification, the documentation package included a 

Certificate of Analysis from Capitol Manufacturing Co., dated September 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 111, Class 2. 

14 

-87- 
in addition to E&P Material Certification, the document package included a CMTR 

from Bonney Forge, dated September 30, 198 1 , 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 data 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 

Items 1 through 4, the material producing mill certifications, including heat 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;50-353196-20141. 

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

Corporation (Nova). 

Item 1 of this PO was for ten 1 118 inch ASTM A-325, Type 1, heavy hex bolts, to 

be supplied in accordance with the applicable provisions of 10 CFR Part 508 

Appendix B. 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 material included Nova Certificate of Compliance 

which identified the material by heat number, described the heat treatment, and 

provided a quality program statement attesting compliance with the PO 

requirements and referencing Nova OA manual and their ISO-900 1 Certificate. 

Attached to the Nova certification was a CMTR 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 15521 2, dated August 28, 1996, to Nova. 

Item 1 of this PO was for twenty four 318 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 6. A Certificate of Conformance was 

required to be provided for this material. 

The document package for this material included 8 Nova Certificate of Compliance 

describing the material and attesting to compliance with the PO and the applicable 

specifications. It also included a quality program statement, certifying to 

compliance with the applicable portions of the Nova QA manual and stating that 

the material has been processed per IS0-9001(94), Certificate # GQC 21 1. 

15 

-88- 
C. 

A review of documentation packages for material purchased to the ASME Code 

requirements identified several examples of apparent failure to assure that the 

Material Organizations supplying this material provided all of the Code-required 

documentation. Specifically, Paragraph NCA-3861 (b) of Section 111 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 

several instances 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 50-352;50-353/96-201-01. 

Document packages for safety-related material purchased for non-Code applications 

were also reviewed. Such material was procured to 10 CFR Part 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 that some 

material orders required only the supplier's certificate of conformance, while others 

required traceable CMTRs including the actual results of all required chemical 

analyses and examinations. 

E7.4 Review o f Procurement from Amer Industrial Technoloaies. Inc. (A IT1 

8. JnsDection SCOD~ 

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. 99901 292/96-01) 

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 

-89- 
Pbservations and flndlw .. P* 

The inspectors reviewed the 1993 PECO audit and material supply history of AIT 

with PECO and determined that PECO had not identified problems with AIT 

supplied material received to date. The inspectors 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 heat 

exchanger, (2) a review of the NRC inspection report and the effectiveness 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 these actions PECO issued a Nonconformance report and requested 

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 Limerick Unit 1. This evaluation determined that the RHR heat 

exchangers were acceptable for current and continued use for all modes of 

operation at Limerick 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 

learned perspective and training sessions were conducted with all auditors on a 

lessons learned on vendor audit methodology covering areas such as ASME Code 

Section 111 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 1 993, Corrective Action 

Request (CAR) Q-4320/Quality Evaluation 00004320 was initiated by PECO to 

conduct a self-assessment to document and resolve problems associated with the 

NQA 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 personnel 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. 

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 sudit that required either 

additional actions or compensatory measures to be taken, and the actions taken by 

17 

-90- 
r E7.5 

a. 

b. 

I 

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. 

Review of Rece nt 1OC FR Part 2 1 Activitia 

bsoection SCODQ 

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

recent information which B&G Manufacturing Company (B&G) had provided PECO 

concerning potentially defective (inadequately heat treated) grade 67 fasteners. 

The fasteners had been originally 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. 

Qbservations and Findinu 

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 

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 Industrial 

Products KIP) had provided PECO with 3" 87 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 determined 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 A10231 18 

(Limerick) and A1 0231 19 (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. 

10 

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Nonconformance Report (NCR) LG 96-01 975 was initiated, which documented the 

PECO operability determination that the valves were operable 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. 

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 Meetin= 

X3 Exlt 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 

0. Fetters, Vice President - Station Support 

1. Niessen, Director, Nuclear Quality Assurance 

J. Cotton, Director, Engineering 

H. Birch, Manager, Supply Management 

W. Texter, Manager, Corporate Nuclear Quality Assurance 

1. 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 

G. Cwalina, Chief, Vendor Inspection Section, Special Inspection Branch 

F. Rinaldi, Project Manager 

J. Bednar, Foreign Assignee 

19 

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APPENDIX A 

This report categorizes the inspection findings as unresolved items and inspection followup 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 will issue any enforcement action resulting from their 

review of the identified unresolved items. An inspection follow-up item (IF11 is a matter 

that requires further inspection because of a potential problem, because specific licensee 

or NRC action is pending, or because additional information is needed that was not 

available at the time of the inspection. 

URI Failure to meet ASME NCA-3861 and 3862 

(Section E7.5 - b.1, b.2, and b.3) 

A- 1 

-93- 
I 

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

-August 14, 1997 

WASHINGTON. D.C. -1 

Mr. Jerry Ethridge, Project Manager 

Tritium Target Qua I i f i ca t io n Program 

Battelle Boulevard 

P.O. Box 999 

Richland, Washington 99352 

Subject: NRC INSPECTION REPORT 99900541 197-01 

Dear Mr. Ethridge: 

On July 18, 1997 tk U.S. Nuclear Regulatory Commission (NRC) c-.rn;leted an 

inspection of the Pacific Northwest National Laboratory (PNNL) activities relaring 

to the Tritium Target Qualification Program. The enclosed report presents the 

results of the inspection. 

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

implementation of your quality assurance program failed to meet certain NRC 

requirements. The specific instances are described in the enclosed report. 

However, as described in the report, the team noted that PNNL took 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 CFR 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, 

1 

Stuart A. Richards, Chief 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Docket No.: 99900541 

Enclosure: Inspection Report 99900541 /97-01 

-94- 

I 
Max Clausen 

Office of Commercial Light-Water 

Reactor Production 

Tritium Project Ofice 

U.S. Department of Energy 

1000 Independence Avenue, SW 

Washington, DC 20585 

Stephen M. Sohinki, Director 

office of Commercial Light-Water 

Oefense Programs 

U.S. Department of Energy 

1000 Independence Avenue, SW 

Washington, DC 20585 

Reactor Production 

DP-60 Records Management 

Office of Commercial Light-Water 

Reactor Production 

Tritium Project Office 

U.S. Department of Energy 

1000 Independence Avenue, SW 

Washington, DC 20585 

Mr. Oliver 0. Kingsley, Jr. 

President, lVA Nuclear and 

Chief Nuclear Officer 

Tennessee Valley Authority 

6A Lookout Place 

1101 Market Street 

Chattanooga, Tennessee 37402-2801 

Mr. Paul L. Pace, Manager 

Licensing and Industry Affairs 

Watts Bar Nudear Plant 

Tennessee Valley Authority 

P.O. Box 2000 

Spring City, Tennessee 37381 

Regional Administrator 

U.S. Nuclear Regulatory Commission 

Region II 

61 Forsyth Street, SW, Suite 23185 

Atlanta, GA 30303-341 5 

County Executive 

Rhea County Courthouse 

Dayton, TN 37321 

-95- 

County Executive 

Meigs County Courthouse 

Decatur, TN 37322 

Mr. Michael H. Mobley, Director 

Division of Radiological Health 

3rd Floor, L and C Annex 

401 Church Street 

Nashville, TN 37243-1 532 
Report No: 

Organization: 

Contact: 

Nuclear Activity: 

Dates: 

Inspectors: 

Approved by: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

9990054 1 /97-0 1 

Project 697 

Pacific Northwest National Laboratory 

Richland, Washington 

Jerry Ethridge, Project Manager 

Tritium Target Qualification Project 

(509) 372-4991 

Fabrication of the Tritium Producing Burnable Absorber 

Rods for the Tritium Target Qualification Project, Lead 

Test Assemblies 

April 29 - May 2, 1997 

July 7-1 1, 1997 

July 14-18, 1997 

Steven M. Matthews, Team Leader, VIS/PSIB/DISP 

Robert M. Latta 

Kenneth C. Heck 

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 (QAI program for supplying Tritium Producing Burnable 

Absorber Rods (TPBARs) for the Tritium Target Qualification Project (TTQP), 

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: 

0 Appendix A, ‘Genet al Design Criteria for Nuclear Power Plants,” 

General Design Criteria (GDC) 10, YReactor Design,” and GDC 12, 

‘Suppression of Reactor Power Oscillations,” to Part 50 of Title 10 of 

the Code of Federal Regulations (10 CFR Part 50) 

0 Appendix B, “Quality Assurance Criteria for Nuclear Power Plar,:s and 

Fuel Reprocessing Plants,” to 10 CFR Part 50 

0 10 CFR Part 21, “Reporting of Defects and Noncompliance” 

Tennessee Valley Authority (TVA) Nuclear Quality Assurance (NQA) 

Plan (TVA-NQA-PLN89-A), Revision 6, dated August 31, 1995. 

0 PNNL’s TTQP, ‘Quality Assurance Plan,” ETD-003, Revision 4, dated 

July 1997 

During this inspection, several nonconformances 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 

During the weeks of April 29 through May 2, 1997, and July 7 through 11, 

-2- 

-97- 
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 lTQP project documents that described the quality 

and technical requirements imposed on PNNL by the host utility (TVA) for 

these safety-related components. The team also examined the adequacy 

and implementation of PNNL's quality assurance plan described in EDT-003, 

"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 component safety 

classification; QA program adequacy; and QA program implementation in 

the areas of audits, corrective actions, training, and design control. 

.. . 3.1.1 Safety classiflcatron 

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-1141 9/UC-731), dated November 1996. The purpose of 

this report was to provide technical information related to the anticipated 

irradiation of TP6ARs in a commercial light 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 water reactor. 

I 

Based on NRC's review of the report, a request for additional information 

(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 Februqry 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 initial inspection Of 

I -3- 

-98- 
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-NQA-PLN89-A, and PNNLs "importance 

factors" described in procedure lTQP-1-046, Revision 0. 

Subsequent to a June 4, 1997, public meeting between the NRC and TVA, 

the staff provided further amplification 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 were 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., 

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 entirety (end plugs, 

getter, cladding, plenum spring, etc.). 

a. 

To evaluate the acceptability of PNNL's safety classification process 

for TPBAR components, the team reviewed PNNL's controlling 

procedure 77QP-1-046, "Tritium Target Qualification Project, TPBAR 

Component Characteristics and Related Importance Factors," Revision 

3, dated July, 1997. The following paragraphs summarize the results 

of this review. 

As determined by the team, procedure TOP-1-046, Revision 3 had been 

revised to comply with the Nhi's 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 TTQP-1-046 

designates those TPBAR components and critical characteristics, as either 

Category A or B. 

As defined in TQP-1-046, 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- 
characteristics are defined as those that could (1 ) significantly affect the 

mechanical integrity of the TPBAR, or (2) result in incremental tritium 

releases and either onsite or offsite doses, or (3) result in localized core 

power peaking. During the review oflTQP-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. Cnncluslnn 

Based on subsequent reviews related to this area, the team determined that 

procedure lTQP- 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 

PNNL's project quality assurance program is described in procedure 

ETD-003, "Tritium Target Qualification Project, Quality Assurance Plan," 

Revision 3. This ITQP project QA plan encompasses all quality activities 

related to the TPBARs, including design, procurement, process development, 

fabrication, inspection, testing, verification and assessment. 

a. 

The team evaluated the adequacy of PNNL's quality assurance program with 

respect Appendix B to 10 CFR Part 50 requirements. 

.. b. & 

NRC Regulatory Guide 1.28, "Quality Assurance Program Requirements 

(Design and Construction), " Revision 3, dated August, 1 Sd5 conditionally 

endorses ANNASME NQA- 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 ANWASME 

NQA-1-1989, and the referenced documents which implement the ITQP 

project QA program. The team also examined PNNL's QA procedures with 

respect to the conditions imposed by Reg Guide 1.28. Additionally, the 

-5- 

- 1 00- 
r 

~ 

I 

I 

I 

team performed a comparison of the programmatic requirements contained 

in ANWASME NQA-1-1983 versus those of ANSUASME NQA-1-1989 in 

order to determine if PNNL's quality assurance program appropriately 

addressed the Regulatory Positions contained in Reg Guide 1.28. 

c. Cnnclusinns 

Based on the team's review of ETD-003, Rev 3, and the implementing QA 

program procedures, it was determined that PNNL's TOP project QA 

program adequately addressed the requirements of 10 CFR Part 21 and 

Appendix B of 10 CFR Part 50. 

3.1.3 Prcqtam 

a. 

The team reviewed selected internal assessments and surveillances of PNNL 

activities controlled by the quality assurance program. 

.. b. Fmhgs 

Criterion XVIII, "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 program and to 

determine the effectiveness of the program. Contrary to this requirement, 

the team determined that not all aspects of the quality assurance 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 audits 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) 97-01 0. This CAR addressed this 

deficiency and, as documented by CAR 97-010, PNNL attributed the root 

cause to its reliance on internal assessments, 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. 

-101- 
(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 

Responsibility for scheduling annual audits and assuring that 

findings, and corrective actions taken to resolve them. 

- 

they are performed and reported to management has been 

assigned to the project lead quality engineer. 

(2) Internal audit procedure lTQP-7-048 has been issued which describes 

the planning, scheduling, 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 program are audited. 

Additionally, audits of all activities, with the exception of 

Organization (Criterion I), QA Records (Criterion XVII), and Audits 

(Criterion XVIII) 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. 

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- 

- 102- 
3.1.4 Corrective Actinn Pragram 

a. 

In order to evaluate the adequacy of PNNL's corrective action program, the 

team reviewed the status of actions taken to correct identified deficiencies. 

In particular, the team examined deficiencies that had been identified during 

an internal assessment conducted on May 1-9, 1996, deficiencies identified 

by an external audit conducted by Westinghouse on November 18-21, 

1996, with a follow up audit conducted on June 2-5, 1997, and other 

deficiencies identified by an external audit conducted by TVA on March 24- 

27, 1997. 

.. b. a 

The internal assessment report dated May 14, 1996, indicated that PNNL's 

corrective action programs were not being used because the tracking 

system employed by the QA department was cumbersome. 

The Westinghouse audit conducted in November 1 996 found that identified 

problems were being tracked by a TTQP project-specific action tracking 

system, but that the tracking system was not described by the QA program. 

Additionally, the Westinghouse follow up audit determined that 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. 

Nonconformance reports and deviation reports were not being used. 

Consequently, trending, root cause, 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: 

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

TIQP-7-037, "Corrective Action Processing Procedure for Deficiency 

Reports and Corrective Action Reports for the Tritium Target 

Qualification Project," Revision 2, dated July 1997. 

-103- 
TTQP-7-040, "Nonconformance Reports" Revision 0, dated May 

1997. 

The team reviewed a recently developed TTQP project corrective action 

tracking system printout. Based on the review of this printout the team 

found that the actions responsive to TVA's audit finding on the corrective 

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 TTQP Project Manager, dated 

July 3, 1997. 

The team reviewed the corrective actions procedures identified above, 

which had been revised or issued after the TVA audit. These procedures 

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 determined that deficiencies were being reported for all active project 

tasks, including material suppliers, design, 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. Cnnclllsrnns 

Based on evaluation of the 7TQP corrective action 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 trend 

analysis had not been performed. 

3.1.5 Imhi5.g 

a. 

In order to determine the adequacy of the TTQP training program, the team 

reviewed the governing procedure TTQP-7-07 1 , "Training and Qualification 

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

- 104- 
Additionally, the team reviewed a surveillance conducted on training for 

fabrication personnel and examined selected training records. 

.. b. Observationsanrf Finrhgs 

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 ?TOP project office. 

To verify the accessibility and retrievability of training records, the team 

requested the training records following issuance of TPQP-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 I TQP personnel 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 

requirements for personnel in the fabrication facility. The surveillance report 

found that fabrication personnel had met all training requirements and dere 

qualified to perform associated fabrication activities. The team also noted, 

during their inspection of the fabrication facility, that a matrix of training 

requirements and the current training status for fabrication personnel was 

posted prominently at the entrance. 

c. Conciuslnn 

Team inspection results confirmed that fabrication personnel satisfied the 

training requirements for job performance and that appropriate training 

records were maintained and retrievable. 

a. 

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 formal design outputs. 

-1 a- 

- 105- 
.. b. C)hsArvationsand F~rldrngs 

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 included: 

- TTQP-1-017, 'Design Analysis/Calculation and Associated 

Independent Review", Revision 1, dated June 1997 

- TOP-1-07 9, "Design Change Control", Revision 0, dated January 

1997 

- TOP-1 -021, "Design Interface Controls Process", Revision 1, dated 

TQP-1-022, "Design Requirements", Revision 1, dated June 1997 

June 1997 

- 

TOP-1 -058, "Design Change Impacts on Technical and Functional 

Requirements", Revision 0, dated March 1997 

PNNL-TOP-1-580, "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 

design control processes. 

The team interviewed representatives from PNNL's Task 1 design 

organization in order to gain 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 appIovaI 

-1 1- 

- 106- 
by a Design Review Board (ORB) 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 Bar 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 reload. 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 deqn control 

program. As determined by the team, all corrective actions 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 end 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 technicaVfunctional requirements. 

c. Cnncluslnns 

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 

a. 

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, and 

procedures followed. 

-1 2- 

- 107- 
.. b. (Ibservationsanrf Firuimp 

PNNL's procurement process is defined in Section 4.0, "Procurement," of its 

Commercial Light Water Reactor (CLWR), Tritium Target Qualification 

Project (TTQP), 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 TTQP-1-046, 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 

paragraphs 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 

TTQP-1-003, "Specification for 31 6 Stainless Steel Seamless Cladding 

Tubes," Revision 1, dated May 1996. lTQP-1-003 complied with ASTM 

Standard A-77 1-88, "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 TTQP-1-003 require 

double vacuum melted feed stock, and chemistry and inclusion limits on the 

product. The products to be produced from the 316 bar stock procured 

from Westinghouse Hanford for PNNL fabrication was TPBAR clad tubing 

and end-caps. 

PNNL verified the adequacy of the stainless steel material in accordance 

with TTQP-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 TTQP-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 TTQP Deficiency Report (DR) 07-066, dated 

May 1 , 1997. The team reviewed PNNL's corrective actions that included 

revising procedure TTQP-2-001 to address the weaknesses identified and 
correctly completing the Material Reverification Record. The team found 

these corrective actions adequate and closed the nonconformance. 

During this review the team also determined that TTQP-1-046 failed to 

adequately describe the actual sample size PNNL used to confirm the 

chemistry and inclusions of the stain'$:-s steel cladding bar stock or the 

actual characteristics verified during PNNL's reverification process. In 

addition the team determined that material specification TTQP- 1-004, 

"Specification for Target Rod End Cap Bar Stock Material," failed to agree 

with the importance factor sampling plans specified in TTQP-1-046 and that 

TTQP-1-046 did not adequately address the use of the ASTM standard 

sampling frequencies, where applicable, that PNNL actually used during its 

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 TTQP-1-046. In addition, both the 

sampling requirements and nomenclature used to describe 

characteristics in TTQP-1-004, Revision 1, "Specification for 

Target Rod End Cap Bar Stock Material," are inconsistent with 

the sampling requirements and nomenclature used in TOP-1 - 

046. 

On the generic basis of this nonconformance, PNNL required all of its 

material specifications to be compared to TTQP-1-046 to assure agreement 

in all cases. 

The team reviewed the effectiveness of this corrective action and 

determined that the actions ta'.?n by PNNL were adequate to address the 

specific item issue and the generic implications of this issue. The team 

therefore, closed this nonconformarse. 

b.2 Lithium Aluminate Pellets 

PNNL procured the Lithium Aluminate pellets from IC1 Advanced Ceramics. 

The pellets were procured in accordance with TTQP-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 

-14- 

- 109- 
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 IC1 to 

submit to PNNL the sampling plan for review and approval prior to use. 

In a memorandum to PNNL dated May 1, 1997, IC1 specified the sample 

sizes associated with the inspection levels as follows: 

95:95 

95:90 

95:75 

15 pellets 

7 pellets 

3 pellets 

According to ICI, the sample sizes were verified by the QC curves found in 

ANSVASQC Z1.9-1993, "Sampling Procedures and Tables for Inspection by 

Variables for Percent Nonconforming." The team determined that ASQC 

21.9 sample sizes are based on the assumption that the data represents a 

normal distribution. Therefore, in order for PNNL to use this standard to 

determine sampling sizes, PNNL would have to show a documentation of 

ICl'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 

pellets. 

(b) PNNL failed to include these sampling sizes in the lnspection/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 IC1 and established its 

basis for using the small sample sizes to accept the pellets. 

-1 5- 

-1 10- 
(b) The ITls for receipt inspection for all pellets for LTA lots received 

were revised to address the sampling matrix and 

confidence/inspection levels specified in TOP-1 -009. 

The team reviewed the effectiveness of these corrective actions and 

determined that PNNL's actions taken were adequate to address the issues. 

The team, therefore, closed this nonconformance. 

c. ConclLlslons 

The team identified concerns with the procurements of the stainless steel 

material for the TPBAR cladding tubes and end-plugs, and the Lithium 

Aluminate (LiAIO,) pellets. These concerns constituted nonconformances. 

However, PNNL responded with corrective actions that adequate! ; 

addressed the team's concerns and resulted in the team determining that 

the nonconformances were closed. 

.. . .. 3.3 TPBARn ACULIWS 

a. ImpemonSmpe 

During this portion of the inspection, the team evaluated the material control 

for the TPBAR fabrication, handling 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 

and procedures followed. PNNL's fabrication process and quality plan is 

defined in TIQP-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 TfQP-2-014, 

"Tritium Target Rod Fabricating Process Plan," Revision 1, dated June 

1997. 

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. 

-16- 

-1 11- 
b. 1 Material Control 

PNNL's material control process is described in the following: 

- TTQP-2-101, "General Receiving for the Tritium Target Fabrication 

Facility (TTFF)," Revision 0, dated February 1997 

- TTQP-2- 1 02, "Handling of Miscellaneous Components," Revision 0, 

dated February 1997 

- TTQP-2-105, "Control and Inventory for Received Storage," 

Revision 0, dated January 1997 

- TTQP-2-106, "Control and Inventory for Accepted Storage," 

Revision 0, dated January 1997 

- TTQP-2-109, "Control and Inventory for Rejected Storage," 

Revision 0, dated January 1997 

The team reviewed the established measures for material control contained 

in TTQP 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 7TFF were 

appropriately documented and controlled in accordance with established 

procedures. 

b.2 Cladding Tubes 

PNNL's process for handling cladding tubes were documented in TTQP-2- 

103, "Handling of Empty Cladding Tubes," Revision 0, dated February 

1997, lTQP-2-104, "Handling of Loaded Cladding Tubes," Revision 0, dated 

February 1997, and ITQP-2-211, "InspectionTTest Instructions for 

Inspecting Cladding Tubes," Revision 0, dated March 1997. The team 

reviewed the established measures for the control of cladding tubes in 

TTQP-2-103 and -1 04 and determined that the measures established were 

adequate. 

-17- 

-1 12- 
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 

records. 

b.3 Welding 

PNNL's process for welding qualification, performance, and inspection were 

documented in the following: 

- TTQP-2-018, "Qualification Plan for End Plug Welding," Revision 1, 

dated June 1997 

- TTQP-2-024, "Radiography Inspection Procedure Qualification Plan," 

Revision 0, dated June 1997 

- TTQP-2-117, "Top and Bottom End Plug Welding," Revision 0, dated 

June 1997 

- TTQP-2-303, "Weld Visual Inspection," Revision 0, dated May 1997 

- TTQP-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. 

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 the 

process was adequately controlled and performed in accordance with 

established procedures. The team also reviewed the radiographs taken of 

the 12 end-cap welds which were used by PNNL as additional information 

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 NDTRT-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 radiographs of the 12 end-cap welds using the beam-filtered tangential 

radiography technique. The team found the radiographs and the processes 

used to produce them to be adequate. 

-18- 

-1 13- 
On July 17, 1997, the team reviewed the completed welding qualification 

report for the qualification of the end-plug welding qualification. That report 

was documented in TOP-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 

second set of 8 top end-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 

TOP-2-125, "Pencil Assembly Loading," Revision 0, dated May 1997, and 

TOP-2-225, "Pencil Assembly Inspection," Revision 0, dated May 1997. 

The team reviewed the established measures for the control of pencil 

assembly and inspection and determined that the 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. 

For the fabrication activities reviewed, the team found that PNNL performed 

those activates with great care and attention to detail and that the 

personnel involved 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 management. In the exit meetings, on July 11 and July 18, 

1997, the inspectors discussed their findings and concerns. 

-1 9- 

-1 14- 
PARTIAL LIST OF PERSONS CONTACTED 

J. Ethridge 

G. Sorensen 

R. Latorre 

S. English 

D. Sensor 

C. Painter 

L. Erickson 

S. Bales 

D. Rittenhouse 

C. Thornhill 

R. Guenther 

J. Chardos 

QQE 

M. Clausen 

-1 15- 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. -1 

Aiigust 10, 19O7 

Mr. Lew Goetz 

President and CEO 

SOR, Inc. 

14685 West 105th Street 

Lenexa, Kansas 66215-5904 

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. (SOR) facility. The enclosed report presents the 

results of that inspection. 

The inspection was conducted to assess the adequacy of the corrective actions 

SOR took to correct manufacturing defects associated with certain safetyrelated pressure, vacuum, and temperature switches made by SOR and sold to 

1 icensees of nuclear plants. The inspector assessed specific attributes of 

the SOR quality assurance program and 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. 

insulation causing it to crack when it was bent during installation. 

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 SOR’s quality assurance program had not met certain 

NRC requirements imposed upon it by NRC licensees. 

The epoxy hardened the 

In 

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, SOR 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 

Uti1 ities Procurement Issues Committee audit, licensees did not evaluate SOR’s 

action to correct two manufacturing defects that SOR reported to its 

customers. 

-116- 
L. Goetz -2- 

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, 

Stuart A. Richards, Chief 

Special Inspection Branch Di vi si on of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Docket No. 99900824 

Docket No. 99900912 

Enclosures: 1. Notice of Nonconformance 

2. Inspection Report 99900824/97-01 

-1 17- 
NOTICE OF NONCONFORMANCE 

SOR, Inc. 

Lenexa, KS 

Docket No.: 99900824 

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 B to 10 CFR Part 50, "Instructions, Procedures, and 

Drawings," requires, in part, that activities affecting quality shall be 

prescribed by documented instructions and procedures of a type appropriate to 

the circumstances, and shall include appropriate quantitative or qualitative 

acceptance criteria for determining that important activities have been 

sat i sf actori 1 y accolilpl i shed. 

SOR Nuclear Quality Assurance Manual 8303-100, Section 5.2, "Instructions, 

Procedures, and Drawings," dated April 14, 1993, requires, in part, that 

instructions, procedures, and drawings must contain acceptance criteria to 

ensure compliance with customer quality requirements. 

Contrary to these requirements, SOR did not prescribe instructions or 

procedures to ensure that epoxy was not applied on insulated lead wires of 

safety-related 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. 

hardened the insulation causing it to crack when it was bent. 

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. 

cracked 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. 

The epoxy 

In September 

The 

In a June 19, 1997, letter to the NRC, SOR 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 Rockvi 11 e, Mary1 and 

this loth day of August 1997 Enclozure 1 

-118- 

-
Report No: 

Organi zat i on : 

Contact: 

Nuclear Industry 

Activity : 

Dates: 

Inspector: 

Approved by: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

99900824/97-01 

SOR, Inc. (SOR) 

Lenexa, Kansas 

Colbert 0. Turney 

Vice President, Qual i ty Assurance 

9 13 1888-2630 

Pressure, vacuum, and temperature switches 

June 16-19, 1997 

Ani1 S. Gautam, Senior Engineer 

Gregory C. Cwalina, Chief 

Vendor Inspection Section 

Speci a1 Inspection Branch 

Di vi si on of Inspection and Support Programs 

Enclofure 2 

-1 19- 
1 INSPECTION SUMMARY 

During this inspection, the NRC inspector assessed the adequacy of the actions 

taken by SOR to correct manufacturing defects associated with certain Safetyrelated pressure, vacuum, and temperature switches (hereafter referred to as 

switches). The defects included (1) cracked insulation of lead wires for 

switches, (2) leakage of O-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 qual ity assurance program acd 

reporting of defects under 10 CFR Part 21, and licensees' monitoring of SOR'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 sReaulations (10 CFR Part 50) 

SOR Nuclear Quality Assurance Manual (QAM) 8303-100, Revision 9, dated 

April 14, 1993, and associated implementing procedures 

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 Uti1 ities Procurement 

Issues Committee (NUPIC) audit, 1 icensees did not evaluate SOR's corrective 

actions regarding certain manufacturing defects reported by SOR in information 

notices to customers. Licensees' monitoring of SOR's control of quality is 

discussed in Section 3.2 of this report. 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

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 

high-energy-1 ine 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. 

microswitch was readjusted during factory calibration. 

screws allowed the switch diaphragm (seal) to be overpressurized during the 

test and caused it to leak. SOR stated that since other switches with the 

SOR believed that the screws had not been resealed after the 

Failure to reseal the 

2 

- 120- 
same type of housings did not suffer a similar test failure, the test failure 

was attributable to a random occurrence, not to an inherent weakness in the 

design. SOR’s corrective action consisted of (1) resealing the microswitch 

mounting screw threads if the microswitch was readjusted during factory 

calibration, and (2) applying a primer to the microswitch bracket screws to 

improve the curing of the thread sealant in stainless steel housings. 

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. 

SOR 

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 

milliamps (mA) of leakage current at 900 Vac. SOR could not find a root cause 

and believed that the 2 mA leakage current was a random’ anomaly and not 

indicative of a common failure mode. SOR’s basis was that the other specimens 

(1) passed the test at 1500 Vac, (2) had adequate insulation resistance at 500 

Vdc, and (3) had sufficient margin for service conditions because the switches 

were rated for 250 Vac and typically energized for 120 Vac or 125 Vdc 

applications. 

provided by SOR, and because leakage current from moisture intrusion would 

have been higher than 2 mA, the anomaly was satisfactorily addressed. 

issue is closed. 

The inspector determined that on the basis of information 

This 

3 INSPECTION FINDINGS AND OTHER COMMENTS 

3.1 Oualitv Assurance Proaram 

a. JnsDection ScoDe 

The inspector examined the adequacy of SOR’s Part 21 evaluations, 

corrective actions, conformance to procurement documents, and selfassessment of performance. 

b. Observations and Findinas 

The inspector observed that SOR’s QA program was based on the policies 

and criteria of 10 CFR Part 50, Appendix B. 

comprised of the Quality Assurance Vice President (QAVP) and 2 quality 

control (QC) inspectors. The QAVP reported directly to the SOR’s 

President/ CEO. 

a nonconforming i tem until the nonconforming conditions were corrected. 

The QA program staff was 

The QC inspectors were authorized to stop production of 

The inspector observed that SOR had posted sections of the federal 

Reuistec, dated September 19, 1995, concerning the latest changes to 

10 CFR Part 21 but had not posted the complete Part 21 regulation, as is 

required by 10 CFR 21.6. During the inspection, the QAVP posted copies 

of the complete regulation in appropriate locations. 

were ident i f ied. 

No further concerns 

3 

-121- 
The inspector assessed SOR's Part 21 reports and corrective actions for 

manufacturing defects associated with SOR switches during the past 5 

years. Defects included cracked lead wire insulation, leaking O-ring 

seals, and leaking epoxy seals in the switch conduit seal. 

inspector's review is summarized below: 

The 

(1) Cracked Insulated Lead Wires 

Insulated lead wires for the switch enter and exit an epoxy seal in 

the conduit adapter of the switch housing. The conduit adapter is 

potted (sealed) with epoxy to keep moisture from entering the switch 

hous i ng . 

In September 1994, Nebraska Public Power District (for the Cooper 

Power Statim) and Connecticut Yankee Atomic Power Company (for the 

Haddam Neck Plant) notified SOR of 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 

nucl ear-qual i f i ed sui tches (the NRC a1 so i ssued event not i f i cati on 

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 wire 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 inspectors 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. 

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 recomnended 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. 

SOR added 

4 

- 122- 
Leakage of O-Ring Seal 

The vacuum switch included an O-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 diaphragm. 'In 

1992 and 1993, SOR implemented a new program to qualify the vacuum 

switches (SOR test report 9058-102), and discovered that the O-ring 

seal in the vacuum switch was not capable of retaining the required 

maximum 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 O-ring had occurred during 

testing. 

On April 1, 1993, SOR issued "Information Notice Concerning Vacuum 

O-Ring Seal in SOR Nuclear-Qual ified Vacuum Switches" to applicable 

customers regardicg potential leakage in vacuum switches, and 

suggested to customers 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 

model number and JJTTX6, JJTTX7, JJTTX13, or JJTTX14 at the end of the 

model number. 

During this inspection, on June 19, 1997, SOR provided a written 

response (Attachment 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 eliminate the leak path by 

(1) welding the vacuum screw to the vacuum piston and (2) replacing 

the triangular O-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 ident i f i ed. 

Leakage of Conduit Seals 

The switch lead wires pass through the outer nipple of the conduit 

seal connector, through the epoxy seal potted in the nipple, and 

through a glass seal which is soldered inside the nipple. 

1994, during routine testing, SOR discovered leakage of pressure 

through the conduit epoxy seal of "NQ" switches. On June 10, 1994, 

SOR issued "Information Notice Concerning Conduit Seals in SOR 

Nucl ear-Qual if ied Pressure, Vacuum, and Temperature Switches" to 

inform the NRC and customers about the potential leak. 

SOR stated that the leak could lead to reduced insulation resistance 

or loss of function of the switch during or after a HELB. 

suggested to customers that all switches be returned to the SOR 

factory 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. 

In May 

In the notice, 

SOR 

5 

-123- 
During the inspection, on June 19, 1997, SOR provided a written 

response (Attachment 3) to the inspector. 

went undetected because (1) the leakage was a random problem, (2) SOR’s inspection steps were not adequate to identify the faulty 

condition seals, and (3) there was a manufacturing error in the heat 

cure of the epoxy because manufacturing personnel had not followed 

procedures. SOR 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. 

SOR stated that the problem 

In June 1994, SOR took measures to prevent recurrence by implementing 

more stringent testing on all conduit seals, including requiring (1) 

an insulation resistance test for conduit seals, (2) a 100 psi leak 

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 temperature 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 $OR’s implementation of 1 icensee purchase order 

requirements in SOR’s design documents (SOR assembly drawings 8520-264 

Revision 2, 8520-506 Revision 1, and 8215-659 Revision 2). No concerns 

were identified. 

The inspector observed that SOR’s General Instructions did not address 

the protection of switch components during handling (e.g., debris 

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 corrective actions. No concerns were 

i dent i f i ed. 

c. Conclusions 

In general, SOR’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. 

and steps to prevent recurrence of identified manufacturing defects. 

SOR took adequate corrective actions 

6 

- 124- 
3.2 Review of Licensee Monitorinci of SOR 

a. InsDection ScoDe 

The inspector evaluated 1 icensee monitoring of SOR’s control of qual i ty 

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 Gas 8 Electric Company, and Yankee Atomic Electric Company - 

audited SOR’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. The NUPIC audit team identified SOR deficiencies in 

the areas of (1) control of purchased materials, (2) test control, (3) 

corrective actions, and (4) control of measuring and test equipment. 

NUPIC considered the findings to be “administrative” and believed there 

was no adverse impact on the quality of SOR’s completed products. 

accepted SOR’s corrective actions for the above findings and closed the 

findings on April 11, 1995. NUPIC concluded that SOR’s QA program was 

adequate and that implementation was satisfactory. 

NUPIC 

The inspector observed that during its audit, NUPIC reviewed SOR’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 during the NUPIC audit, 

1 icensees did not evaluate SOR’s 1993-1994 corrective actions regarding 

leakage of O-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 qual ity 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 NUPIC report). 

NUPIC recommended, in part, that SOR 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. Foncl us i ons 

In general, licensee monitoring of SOR’s quality was in accordance with 

proper criteria, procedures, and check1 ists. 

SOR’s corrective action for two manufacturing defects reported by SOR to 

its customers. 

NUPIC did not evaluate 

7 

- 125- 
3.3 htrance and E xit Meetinaf 

At the entrance meeting on June 16,-1997, the NRC inspector discussed the 

scope of the inspection, outlined the areas to be inspected, and 

established interactions with SOR management. 

June 19, 1997, the inspector discussed his finding and observations. 

In the exit meeting on 

4 PARTIAL LZST OF PERSONNEL COKTACTED 

Lew Goett, President and CEO 

Col bert Turney, Vice President (VP), Qual ity 

Lind Coutts, Coordinator, Nuclear Engineering 

Joseph Modig, Enginew, Nuclear Engineering 

Landen Tuggle, Director, Manufacturing 

Harold Moddy, VP Sales 

Charisse Smith, VP Finance 

Tim Cei 11 esen , Product Manager 

Richard Johnson, QC Engineer 

Southern California Edison 

Jeff larson, Supervisor, Procurement Qual i ty 

ITEM OPENED, CLOSED, AND DISCUSSED 

99900824/97-01-01 Para 3.1 b NON 

Closed 

99900912!93-01-06 Para 3.10 of Open Item 

i nspect i on report 99900912/93-01 

inadequate instructions 

and procedures 

inadequate information 

regarding test anomal ie 

0 

- 126- 
14685 W 105th Street Lenexa Kansas 662155964 USA 

Tel 913 888 2630 Fax 913 888 0757 

Mr. .Ani1 S. Gautam. SRC 

Subject: 1 OCI- R2 1 datcd October 14. I994 

9001 

Dear Mr. Gautam: 

In July of 1993 SOR began to manufacture pressure. i'acuum. and temperature siLitches 

that were qualified b\ SOR test report 9058- 102. In September of I994 SOR \\as 

notified of a manufacturing defect b: h'ebrasha Public Pouer (Ref. RG.4 2 125. sei en 

defectite units) and Connecticut Yankee (Ref. RGX 21 17. one defective unit). In 

atdition SOR assc-nbl> personnel Fsd identified the same defect (Ref. hfRR 1479. one 

defecti\e unit). The defect \\as identified as a crack in the lead uire insulation 

prompted SOR to issue a 1 OCFRZ 1 and iniestigate the cause of the defect. The cause of 

the cracked insulation \\as a heai: coating ofepou? on the \\ires outside ofths putted 

area and \\as due to poor norkmanship. This condition nent undetected because SOR 

qualit: inspectors did not notice the coating ofepou: on the \tires. 

.-a . [lib 

As noted in the Pan 21 Notitication. the follo\iing correctiie action Mas taken in October 

1994: 

I. The M'ork Order formats for the conduit seals \+ere rexised to include specific 

instructions not to allow epoxy on the \\ires. In addition. there is an inspection step at 

the end of the N'ork Order that instructs the inspector to examine the wire and reject 

any that ha\.e epox). on the \{ire. Each of these steps must be signed off on the U'ork 

Order b). hlanufacturing and QX personnel for each order of conduit seals. 

2. Shrink tubing \vas added to the lead \\ires \\here they exit the conduit seal. The 

purpose of this tubing is to protect the \vie insiilation during shipping and handling. 

This step is signed off on the assembl! procedure by Xianufacturing and it is revieived 

b! QA pcrconnel for every snitch. 

3. The \\.ire manufacturers re. mmended minimum bend radius was added L the SOR 

General Instructions that are provided to the customer with each switch. 

The cpnduit seals are manufactured as a sub-assembly in a separate environmentaily 

controlled room. Ttlerefore. there is no danger of epoxy contamination on any other parts 

of the switches. 

Reference Corrective Action Report 0357. 

Regards. 

-127- 

Joseph G. Modig. Engineer 
SORt 

14685 W. 105th Street Lenexa. Kansas 66215-5964 USA 

Tel. 913.888.2630 Fax 913.888.0767 

-2 

RI~LI~IIYIII 

9001 

Mr. Ani1 S. Gautam, NRC 

Subject: Information Notice of April 1. 1993 

Dear Mr. Gautam: 

Prior to April I. 1993 SOR had been manufacturing nuclear qualified vacuum switches 

for approximately 10 years. These switches were qualified by a combination of Lesting 

and analysis as listed below: 

0 AETC Test Report 17344-82N-D. Rev.1 

0 AETC Test Report 1844 1 -83N. Rev. 1 

0 AETC Test Report 17344-82N-C, Rev. 3 

0 AETC Test Report 18577-83N. Rev. 1 

0 AETC Test Report 18878-84N-2. Rev. 2 

0 SOR Analysis 82 15-959 

In 1992 and I993 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 to meet 

hydrostatic requirements and qualified it by analysis (Ref. SOR Test Report 9058-102, 

Section 14, Appendix 4, Analysis 8923-2 19). 

This condition L as not discovered earlier because of inadequate engineering testirz and 

analysis of the vacuum switch. 

As noted in analysis 8923-2 19, 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 triangular 

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, eliminales 

the leak path mentioned above because the vacuum piston is now welded to the vacuum 

-128- 

14685 W 105th Street Lenexa. Kansas 66215-5964 USA 

Tel 913 888.2630 Fax 913 888 0767 

Mr. Ani1 S. Gautam. NRC 

Subject: Information Notice of June IO. 1994 

Dear Mr. Gautam: 

In July of I993 SOR began to manufacture pressure. vacuum. and temperature switches 

that were qualified bj SOR test report 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 affected utilities on the subject Information 

Notice. This problem went undetected by SOR for three reasons: 

1. The leakage was a random problem. 

2. Inspection steps were not adequate to identify faulty conduit seals. 

2. 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. 

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 signed off 

b! 2.4 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 signed 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 procedures 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 procedure by the 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: 

-130- 
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 

correct 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 Corrective Action Report 0338. 

Regards, 

b&i%t*S 7 ?[ 

Colbert Turney. V.P. Quality 

J 

Joseph G. Modig. Engineer 

-131- 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-OW1 

+**** July 22, 1997 

Mr. Richard P. Bender 

Vice President 

Yuasa Exide, Inc. 

2366 Bernvi 11 e Road, 

Post Office Box 14145, 

Reading, Pennsylvania 19612 

SUBJECT: NRC 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 

activities 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 federa7 Regu7ations (10 CFR 21.21), YE1 failed to inform all 

applicable licensees that certain YE1 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. 

letters to Southern California Edison (San Onofre Nuclear Generating Station) 

and C1 eve1 and Electric I1 1 uminating Company (Perry Nuclear Power P1 ant) 

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 P1 ant, Uni t-2). 

YE1 sent 

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. 

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 action is necessary to ensure compliance 

with regulatory requirements. 

In addition, the NRC found that the implementation of the YE1 quality 

assurance program failed to meet certain NRC requirements imposed on you by 

your customers. YE1 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 YE1 established for review of suitability of application of purchased 

parts and materials to be used in Class 1E batteries and verification that 

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. 

Please note that you are required to respond to this letter 

- 132- 
Mr. Bender -2- 

The failure of the San Onofre cells has potentially generic implications. 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 

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 required time under certain design basis conditions. 

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. 

potential deviation from YEI's pub1 ished product performance claims should be 

thoroughly investigated. 

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 in accordance with the instructions specified in the enc'losed 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). 

In particular, 

The NRC believes that this 

If it is found to be a deviation, all affected 

Sincerely, 

Stuart A. Richards, Chief 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Docket Nos. 39900358, 99900359 

Enclosures: 1. Notice of Violation 

2. Notice of Nonconformance 

3. Inspection Report 99900358,359/97-01 

-133- 
NOTICE OF VIOLATION 

Yuasa Exide, Inc. 

Reading, Pennsylvania/Richmond, Kentucky 

Docket Nos. : 99900358/99900359 

Report No.: 97-01 

During NRC inspections conducted at ydu. 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 

determines that it does not have the capability to perform the evaluation to 

determine if a defect exists, or if the failure to comply is assoc’ated with a 

substantial safety hazard, then the supplier must inform 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, pursuant to 10 CFR 21.21(a). 

Contrary to the above, YE1 failed to inform all affected licensees that 

certain YES GN type battery cells that were manufactured between October 1992 

and December 1992 could potentially have less than the manufacturer’s 

pub1 icized rated capacity of 8-hours. 

Edison (San Onofre Nuclear Generating Station) and Cleveland Electric 

IJluminating 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 GN 

type safety-related battery cells that were manufactured during the same time 

period. Violation 99900358,359/97-01-01. 

YE1 sent letters to Southern California 

This is a Severity Level IV violation (10 CFR Part 2, Appendix C, Supplement VII). 

Pursuant to the provisions of ln 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 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 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 - 32”>day of July, 1997 Enclosure 1 
NOTICE OF NONCONFORMANCE 

Yuasa-Ex 

Read i ng , 

de, Inc. (YEI) Docket Nos. : 99900358/99900359 

Pennsylvania/Richmond, Kentuckj Report No.: 97-01 

Based on the results of an NRC inspection conducted at YE1 facilities in 

Reading, Pennsylvania, ant Richmond, hsit..~~~, on March 4-7, 1997, and on 

April 28-May 2, 1997, respectively, it appears that certain of your activitieAs 

were not conducted in accordance with NRC requirements as follows: 

A. Criterion XV, "Control of Nonconforming Material," of Appendix B, 

"Qual i ty 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 

shall be reviewed and accepted, rejected, repaired or reworked in 

accordance with documented procedures. 

Section 15, "honconforming Material, Parts, or Components," of the YE1 

Qual i ty Assurance (QA) Manual required that QA personnel prepare 

material review reports in cases of nonconforming material and 

disposition the nonconformances. 

Contrary to the above, YE1 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. 

a review for suitability of application for safety-related structures, 

systems, and components. 

B. Criterion 111, "Design Control," of 10 CFR Part 50, Appendix B, requires 

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 1E station batteries for nuclear power 

plants were not adequate as follows: 

e Not all critical characteristics for certain items were 

identified, e.g., O-ring material and cure date, 

e Not all verification methods or acceptance 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. 

e Not a1 1 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 

-135- 
C. Criterion VII, "Control of Purchased Material, Equipment, and Services," 

of 10 CFR Part 50, Appendix B, requires that measures be established to 

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 YEIRichmond, Kentucky, for verification that purchased parts and materials 

to be used in the manufacture of Class 1E station batteries for nuclear 

power plants met the procurement specifications were not adequate as 

fol1 ows: 

0 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 LIPS. 

0 Commercial grade supplier surveys used to support verification of 

one or more critical characteristics for various items were broadbased 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, 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) the reason for the nonconformance, or if 

contested, the basis for disputing the nonconformance, (2) the corrective 

steps that have been taken and me results achieved, (3) the corrective stepr 

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. 

Dated at Rockville, Maryland 

this =day of July, 1997 

-2- 

-136- 

Enclosure 2 
ORGAN I ZAT I ON : 

DOCKET/REPORT NO(s) . : 

ORGANIZATION CONTACT: 

NUCLEAR INDUSTRY: 

ACT I V I TY 

INSPECTION 

CONDUCTED: 

TEAM MEMBERS: 

Approved By: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

Yuasa Exide, Inc. {YEI) 

2366 Bernvi 11 e Road, 

Post Office Box 14145, 

Reading , Pennsylvania 19612 

99900358/97-01 (Reading, Pennsyl vani a) 

99900359/97-01 (Richmond, Kentucky) 

Richard P. Bender, Vice President, Quality Assurance 

and Purchasing (610) 208-1972 

Lana West, Quality Assurance Manager (606)6;,-7346 

Manufactures and supplies stationary batteries and 

battery racks. 

Reading, Pennsylvania, March 4-7, 1997 

Richmond, Kentucky, April 28-May 2, 1997 

K.R. Naidu, Team Leader, NRR 

J.J. Petrosino, NRR 

S.D. Alexander, NRR 

S.N. Saba, NRR 

G.C. Cwalina, Chief 

Vendor Inspection Section 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Enclosure 3 

-137- 
1 INSPECT ION SUMARY 

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 YuasaExide, Inc. (YEI), used in Class lE/vital stc tion batteries at nuclear power 

plants. 

Conducted at YEI's engineering facility in Reading, Pennsylvania, and its 

factory 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 compl i ance, (3) procurement and acceptance of purchased parts and materi a1 s 

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 (POs) from NRC licensees and 

certificates of conformance (COCs) and associated documents provided to ARC 

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 

Reaulations (10 CFR Part 50, Appendix B) 

0 10 CFR Part 21, "Reporting Defects and Noncompliance" 

The inspectors identified two minor violations of 10 CFR Part 21 (§21.21(a) 

and 921.6) (See Section 3,l); 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 VI1 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 

3.1 

a. ScoDe 

INSPECTION FINDINGS AND OTHER COMENTS 

10 CFR Part 21 Procedure and Posting 

The inspectors reviewed YE1 ' s qual i ty assurance procedure (QAP) for reporting 

in accordance with 10 CFR Part 21 (Part 21), QAP 80.0, "lOCFR21 - Procedure 

for Reporting Non-Conforming Material ," dated November 2, 1994. QAP 80.0 was 

developed by YE1 to address Part 21 requirements at the two YE1 facilities 

2 

-138- 
which perform activities relating to "basic components." The inspectors also 

observed and reviewed the document that was posted at both facilities to 

comply with § 21.6 of Part 21, "Posting requirements." 

b. Observations and Findinas 

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 

provisions of 321.21 (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. 

inspectors observed that YEI-Reading, apparently opting for a 321.6(b) 

posting, had postea only a 1,otice that indicated that a copy of 10 CFR Part 21 

was available for review in its administrative office. However, the notice 

lacked the other information required by §21.6(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, YEI-Reading did not post Section 

206 of the Energy Reorganization Act of 1974 as required by both §21.6(a) and 

§21.6(b). At its Richmond, Kentucky, facility, YE1 had complied with the 

posting requirements of §21.6(a). 

The 

The inspectors discussed with the YE1 staff the provisions of 321.21(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 321.21(a)(l) 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 321.21(b) is not specifically required to be included 

in the procedures adopted pursuant to 10 CFR part 21, it is perhaps the most 

important provision for a particular vendor/supp?ier's disposition of 

deviations or failures to comply, because most vendors or suppliers do not 

have the capability to perform a 321.2I(a)(l) evaluation. 

Section 21.3, the 321.21(a)(l) 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. 

agreed that this process would normally be beyond its capability because, 

a'lthough YE1 can advise a licensee or purchaser of the effect of a particular 

deviation on the 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 

As also defined in 

YEI 

3 

-139- 
substantial safety hazard, even if they are not capable of performing a 

§21.21(a)(l) evaluation. However, a vendor who is not qualified to perform 

the §21.21(a)(l) evaluation should not perform the evaluation (in lieu of 

informing affected 1 icensees or purchasers in accordance with §21.21(b)) and 

determine that a report to the NRC is not required because the deviation does 

not appear to the unqualified vendor to be dafect 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 YE1 QA 

staff, the inspectors informed YE1 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 YE1 corporate staff drafted its revision to QAP 80.0 

and the draft document was discussed during the Richmond inspection. 

C. Conclusion 

The inspectors concluded that YE1 had not developed an adequate procedure to 

comply with the requirements of 10 CFR Part 21, and had not complied with 321.6 posting requirements at its Richmond facility. However, YE1 had 

complied with the 921.6 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 §21.21(b). 

3.2 10 CFR Part 21 - Inforrnina Affected Licensees 

a. ScoDe 

The inspectors reviewed an April 19, 1996, YE1 letter that was sent to 

Southern California Edison (San Onofre Nuclear Generating Station) (SONGS) and 

Cleveland Electric Illuminating Company (Perry Nuclear Power Plant) (Perry) regarding GN type battery cells that were supplied. 

provided a copy of the letter to the NRC staff for information. 

inspectors reviewed the letter to determine the adequacy of informing its 

customers df deviations or failures to comply. 

The YE1 author also 

The 

b. Observations and Findinas 

The YE1 letter indicated that certain YE1 GN type battery cells that were 

manufactured between October 1992 and December 1992 could potentially have 

less than YEI’s publicized rated capacity of 8-hours. 

the two licensees and NRC staff that only Perry and SONGS received GN type 

batteries manufactured in the suspect time period. 

The letter stated to 

During the Richmond facility inspection, the inspectors reviewed the letter 

and associated documents including YEI-Richmond’s 1992 “custom order status 

log” (status log) for the suspect time period. The status log is an internal 

YEI-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 

4 

- 140- 
imposed. 

lice~sees that received the YE1 letter had GF! type cells manufactured in the 

subject time period. 

However, the inspectors determined that there was a third licensee that 

received cells manufactured during this time period. It was noted that 

Washington Public Power Supply System (WPPSS) had also received eight GN type 

safety-re1 ated battery cell s. Further review of YE1 s records showed that the 

WPPSS Purchase Order (PO) 221020, dated October 1992, did in fact order eigh3 GN battery cells and imposed nuclear safety-related requirements on YEI. YE1 

committed to informing WPPSS of the deviation by July 25, 1997. 

The inspectors determined that YE1 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). 

The inspectors noted that the status log reflected that both 

This has been identified as Violation 

99900359/97-01-ul. 

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 

controls are nearly identical. 

may be susceptible to the deviation discussed above. Although YE1 has qualified only the GN cells for safety-related service, some licensees buy the 

commercial-grade GC cells and dedicate them. 

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, 1492 (FPC PO A7301166), and NYPA purchased 

four GC-33 cells (PO number undetermined), assembled November 23, 1992, and 

tested December 11, 1992. 

Therefore, the inspectors noted that GC cells 

The inspectors found that 

C. Concl us i on 

The inspectors concluded that YE1 did not perform an adequate review of its 

manufacturing records for the time period in question. 

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. 

Consequently, it did 

3.3 Batterv Manufacturina Process 

a. ScoDe 

The inspectors observed processes in the various stages of manufacturing of 

YE1 GN type battery cells. 

reviewed Jogs 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. 

reviewed production QA activities. 

b. Observations 

The 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, 

The inspectors examined items in production, 

The inspectors also 

5 

-141- 
trimming, etc.), cell assembly and sealing, leak testing (15 GC-25 cells, of 

which two cells had 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. 

fol1 owed. 

No concerns were identified in this area. 

The inspectors observed QC personnel performing various prescribed routine 

checks during the manufacturing process, including inspecting the plates for 

hairline cracks, plate thicknesses, loose/missing pellets, etc. No concerns 

were identified in this area. 

In general, manufacturing instructions were available and being 

Cperators were know1 edgeable ,n concerned with ensuring qual i ty . 

The inspectors noted that while the GN (nuclear qualified Class 1E) and 

GC (commercial) cells are generally similar in design and comparable in 

performance, there are some differences, including the following: 

0 The container material of GN is a type of polycarbonate. The GC 

container material is a type of styrene. 

0 The GN cover is polycarbonate; the GC cover is polyvinylchloride. 

0 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 

The GC cells have not been qualified by YE1 for seismic requirements. 

The GN and GC plate separators are of different materials. 

0 

GC plates. 

0 

0 

c. Conclusions 

The inspectors concluded that the YE1 manufacturing process controls observed 

were effectively implemented. 

3.4 Qualification of YE1 GN-Seri-; Batteries for Class 1E Service 

a. ScoDe 

The inspector reviewed Wyle Laboratories Report 45001-1, Revision A, dated 

January 15, 1989, the environmental and seismic qualification report for YEI's 

"GN"-Series batteries and battery racks for Class 1E service in nuclear power 

plants and interviewed YEI's Manager of Engineering Support, Large Stationary 

Batteries, who had witnessed the Wyle testing. The inspector also reviewed 

the report prepared by Flight Dynamics, Inc., in which it documented the 

seismic qualification analysis for the YE1 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 nonnuclear applications. 

6 

- 142- 
I 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 pro5r;m YJ:S 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. 

to save weight and cost without sacrificing ampacity. In order to qualify the 

improved design, YE1 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. 

However, the non-nuclear line had larger single posts 

No deficiencies were noted in this report. I 

c. Conclusions 

The original qual ification test program 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 1E equipment qualification), IEEE 

Std 344-1975 (seismic qualification of Class 1E equipment), IEEE Std 450-1987 

(Class 1E battery qualification), and IEEE Std 535-1987 (Class 1E battery 

seismic qualification). 

3.5 Purchase Orders 

The inspectors had no concerns in this area. 

a. ScoDe 

At the YE1 Reading and Richmond facilities, the inspectors reviewed selected 

licensee purchase orders (POs) for YE1 GN type batteries and associated 

records to determine whether NRC licensees imposed the necessary and 

appropriate technical and quality requirements on YE1 for the procurement of 

basic components. 

b. Observations and Findinas 

The inspectors noted that each of the licensee POs reviewed imposed the 

requirements of 10 CFR Part 50, Appendix B, and stated that 10 CFR Part 21 was 

applicable. Each of the YE1 customer files (called PO packages) contained 

7 

-143- 
documents such as test and manufacturing records, COCs, and Lab Test 

Assignment (LTA) sheets. 

manufacturing, test and quality assurance function traveler. 

particular job contains 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 PO 

packages reviewed contained appropriate quality and test requirements from the 

licensees except for one package discussed in Section 3.6 below. 

The LTA sheet is a form that is used as a 

The LTA for each 

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. ScoDe 

The inspectors reviewed Southern California Edison (SCE) PO 6L225004, dated 

March 1, 1995 (with subsequent revisions in October 1995 and August 1996), and 

associated documentation for ten 2GN-15 replacement cells 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 Findinqs 

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. 

cells finally passed with acceptance criteria modified by SCE. 

The 

Although the test failures were largely attributable to some weak cells (e.g., 

5000l), YE1 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. YE1 explained that they had experienced difficulty in 

maintaining lower temperatures due to inadequate air conditioning in the test 

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 YE1 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 hours (7 

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. 

temperatures , most of the cell s suffered degraded performance; a1 though , cell 

In this test at the highest 

8 

- 144- 
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 durini the previous tests run at lower 

temperatures. Thus, the data suggested that one or two cells were weak, but 

a?? the cells exhibited a sensitivity to high temperature. 

Section 15, "Nonconforming Material, Parts, or Components," and Section 16, 

"Corrective Action, I' of YEI's QA Manua? required that quality personnel 

prepare a material review report and disposition the nonconformances in 

accordance with the QA program. However, YE1 did not generate any material 

review reports to document the deviation from the 1 icensee's PO requirements 

and its disposition, as required by its QA procedures. 

YEI, after the fourth test failure, the YE1 Richmond facility QC staff 

contacted YE1 corporate engineering, and YE1 engineering contacted the SCE 

engineering staff. After consultation between the YE1 and SCE engineering 

staffs, SCE changed its testing requirements to 95% capacity at the 4-hour 

discharge rate. The cells wt this requirement during the fifth test an\, were 

shipped to SONGS. 

Instead, according to 

c. Conclusion 

The inspectors concluded that YE1 had established measures to assure that 

conditions adverse to quality, such as test failures, and deviations are 

promptly identified and corrected, and that YE1 had procedures to assure that 

nonconforming i tems would be reviewed and accepted, rejected, repaired or 

reworked in accordance with documented procedures. However, YE1 failed to 

document and disposition the four test failures of the SONGS cells in 

accordance with its 10 CFR Part 50, Appendix 8 quality assurance program. 

Accordingly, Nonconformance 99900359/97-03-02 with respect to Criterion XV of 

10 CFR Part 50, Appendix B, was identified in this area. 

3.7 Safetv ImDlications of Failure of SONGS Cells 

a. ScoDe 

The failure of the San Onofre cells is of technical concern with potentially 

generic implications. 

SONGS batteries, the inspectors were concerned that perhaps YEI's advertised 

8-hour capability for the 2GN cells i n i t s product literature was not always 

achievable, particularly under elevated ambient and cell temperature 

conditions. 

cells to pass their 8-hour capacity test discharge, the inspectors reviewed 

test data for several other groups of similar cells for other plants to 

determine whether there was any inherent difficulty in YE1 2GN cells meeting 

the manufacturer's published 8-hour discharge capability. 

In addition to concerns about the requirements for 

To try to resolve concerns raised by the failure of the SONGS 

b. Observations 

The 8-hour-rate capacity test data for several other batteries with similar 

cell types (2GC) with 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. 

suscept i bi 1 ity to premature loss of capacity at elevated temperatures. 

C. Concl us i on 

Based on the cell test data reviewed, the insyectors concluded that YE1 cells 

of this type should be able to meet an 8-hour rate, 90-percent capacity 

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 

SONGS was the poor performance of Cells 50001 and 50002. 

apparent susceptibility of the 2GN cell to premature loss of capacity at 

elevated temperatures may be indicative of an inherent weakness in some aspect 

of the design or manufacturing process. 

2GN cell is of a relatively low electrolyte volume design. This feature 

causes the cell to exhibit a capaci ty-versus-di scharge-rate prof i 1 e typical of 

lead-acid cells up o dbout 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. 

concerned that should this become limiting, particularly at high temperatures, 

it could result in the unexpected inability of station batteries to maintain 

required voltage for the required 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. 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 Fol lowup Item 99900358,359/97-01-05. 

Instead of ordering a replacement for the weak cells, SCE revised the test 

acceptance criteria. The inspectors were not able to determine at YE1 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 1E station batteries. 

Therefore, the inspectors could not rule out generic 

However, the 

For example, according to YEI, the 

The inspectors were 

3.8 YE1 Commercial Grade Dedication Prowain 

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. YE1 uses the provisions of its commercial grade dedication program as a systematic means to verify that purchased 

material and components used in the manufacture of Class 1E GN-type battery 

cells are suitable for safety-related service. In addition, the seismically 

qualified battery rack systems, manufactured to YE1 specifications by the KIM 

Company, are purchased by YE1 as comnercial grade items, dedicated by YE1 and 

supplied to NRC licensees as basic components. An adequate and effectively 

implemented commercial 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 requirements of Criterion 111, "Design Control ," and Criterion VII, "Control of 

Purchased Material, Equipment, and Services," of 10 CFR Part 50, Appendix 6. 

I 

10 

146- 
a. ScoDe 

a. 1 

In order to evaluate the effectiveness of the YE1 QA program and its 

implementation regarding controls applicable to the review for suitability of 

application and design verification in accordance with Criterion 111, the 

inspector reviewed YE1 Procedure QAP-70.0, "Dedication of Commercial Grade , 

Items (CGI) for Nuclear Safety Related Applications," revision dated April '20, 

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, YE1 design documents (principally drawings), and YE1 procurement specifications, were 

a1 so reviewed. 

Procedures and Technical Eva1 uat ionjRevi ew for Sui tabi 1 i ty 

a.2 Acceptance/Procur?ment Specification Compliance 

To evaluate the effectiveness of the YE1 QA program and its implementatior, 

regarding controls applicable to supplier selection and qualification, review 

of supplier documentation and examination of purchased material and components 

in accordance with Criterion VI1 for verification of satisfaction of 

procurement specification requirements, the inspectors reviewed acceptance 

process sheets, incoming inspection reports (IIRs), supplier audits and 

commerci a1 grade surveys, independent 1 aboratory materi a1 analysis reports, 

procurement documents and suppl ier certificates of conformance. The 

inspectors a1 so interviewed technicians and QA/QC personnel and examined 

selected purchased component parts and materi a1 s. 

b. Observations and Findinqs 

b.1 Procedures and Technical Evaluation/Review for Suitability 

The inspector determined that QAP 70.0 was not fully consistent with the 

requirements of Criteria 111 and VI1 of 10 CFR Part 50, Appendix 6, and the 

provisions of Electric Power Research Institute (EPRI) Report NP-5652, 

"Guideline for the Use of Commercial Grade Items in Nuclear Safety Related 

Appl icatiens (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, " issued 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 on 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 Generic Letter 89-02 on the 

use of Acceptance Methods 2 (commercial grade surveys) and 4 (product and 

suppl ier performance history) alone of EPRI Report 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 mu1 tiple acceptance methods for many items and re1 ied predominantly 

on Method 1 (special tests and inspections) when a single method was employed. 

11 

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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-6466, "Technical Evaluation of Replacement 

Items (NCIG-ll)." However, these definitions and the stated position (as in 

NP-6406) that critical characteristics fnr acceptance are a subset of critical 

characteristics for de;!gn are not cui,,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 characteristics 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 prccEss worksheets for the battery racks and various battery 

components and materials, the lists of critical characteristics for acceptance 

and their associated verification methods and acceptance criteria were not a!l 

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 

in some cases. For example, QAP 70.0, No. 100-Series Attachments, the 

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 O-ring 

material and cure date were omitted. Since the battery racks must remain 

structurally sound, and the cell connections remain tight during a design 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 acceptance 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 documei,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. 

example, the incoming inspection form specified the wrong durometer hardness 

value for post-seal O-rings. Although this had been corrected by a pen-andink 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 detect components that are 

not capable of performing their safety functions under all application design 

basis service conditions. 

In another 

12 

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b. 2 Acceptance/Procurement Specification Compl i ance 

For those critical characteristics that were identified, not all were 

correctly specified or adequately verified. 

material as an example of inappropriate verification method and acceptance 

criteria that was seen througnout tb, e'e,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. 

IIR forms (for which there was no procedure for preparation or for use and 

little training beyond OJT) where material was identified as a critical 

characteristic, would simply say "material 

characteristic column headings. For example, IIRs for intercell connector 

fasteners, specifically, bolts, specified "SS-316," for the material and the 

blocks for each sample specimen would simply say "yes." 

yes just reflected the belief that the material was correct based on markings and other factors. 

simi 1 arly inappropriately verified. 

The inspectors identified 

The 

under one of the critical 

In effect, entering a 

IIRs for post seal caps, showed that the material was 

The inspectors followed up on the question of intercell connector fastener 

material in more detail. The material specified for intercell connector 

fasteners, expressed as "SS-316," was supposed to be ASTM Type 316 stainless 

steel. YE1 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 suppl iers 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. YE1 also did not maintain lot 

traceability on the fasteners. YE1 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 YE1 documented the history of the 

consistency of this attribute (material). 

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 

material, 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 sui tabi 1 i ty. 

COCs from TSP were not in accordance with the PO requirements (B and N of QAP500). 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. 

Finally, the audits were not 

The inspectors a1 so noted that the 

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 COCs. The fasteners were confirmed to be nonmagnetic, consistent with an austenitic stainless steel such as 316. In 

addition, YE1 milled a bolt taken from stock to check for indication of 

irregularities such as welded-on bolt heads. 

apparent by visual examination. 

No such indications were 

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 YE1 purchase specification, acceptance process 

sheet, and the IIR form expressed this parameter in terms of ohms/square inch/mil of separator web thickness. 

par1 ance) i s actual 1 y conduct i on path 1 engt h-speci f i c resi stance (res i s t i vi ty 

normalized on conduction path cross sectional area) and is properly expressed 

in units of ohm-inch/mil of separator web thickness. 

resistance was specifiA using incorrect units, the inspector’s cakdation 

confirmed that the numerical values specified were consistent with the 

manufacturer’s specifications. 

Separator resistance (as used in battery 

Although the separator 

c. Concl us i ons 

The deficiencies in the YE1 commercial grade dedication program description 

documents is considered a weakness in the YE1 QA Program with respect to 

conformance to Criteria I11 and VI1 of 10 CFR Part 50, Appendix B. 

addition, the YE1 definition of the term critical characteristic in QAP 70.0 

was inconsistent with 10 CFR Part 21. 

In 

The inspectors concluded that contrary to these requirements, the measures 

established by YEI-Reading, and implemented by YE1 Richmond, Kentucky, for 

review 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 1E station batteries for 

nuclear power plants did not meet the requirements of Criterion 111 of 10 CFR 

Part 50, Appendix B. Accordingly, Nonconformance 99900358,359/97-01-03 was 

i dent i f i ed . 

In addition, YEI’s measures for verification that th 

materials met the procurement specifications did not 

Criterion VI1 of 10 CFR Part 50, Appendix B. Accord 

as part of Nonconformance 99900359/97-01-04. 

3.9 SuDDlier Qualitv Audits 

a. ScoDe 

se purchased parts and 

meet the requirements of 

ngly, this was identified 

The inspectors reviewed YE1 audit and commercial grade survey procedures and 

checklists and also reviewed audits performed by YE1 on its vendors. 

inspectors evaluated them to determine if YE1 auditors verified critical 

characteristics of the items furnished by those suppliers, and if the audits 

were performance based. 

The 

14 

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b. Observations and Findinas 

According to the audit reports reviewed, YE1 used checklists to perform the 

audi t/surveys of Cobra Wire and Cab1 e, Inc., herace-Mi croporous Products , Inc., I.E. Dupont De Nemours & Coy Threaded Screw Products, Inc., and KIM 

Engineering Company (KIM), who supplied cable, separators, acid, bolts & nuts, 

and battery racks respectively. 

Corrective Action Requests (ACARs) and requested suppliers to return the ACARs 

with the proposed actions to correct them. 

proposed corrective act i ons were acceptable. 

The inspectors determined that the attributes in the audit checklist were 

common to all suppliers and covered a broad programmatic overview of the 

supplier rather than focusing on the specific item being supplied. 

Furthermore, the audit did not attempt to verify the control of the quality of 

the critical c h a r a c t x i s t i c s o f the specific item being manufactured. 

YE1 documer.tzd adverse conditions in Audit 

YE1 then closed the ACARs if the 

For example, the audit of KIM failed to specifically verify the qualirications 

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 YE1 auditors did not 

verify this attribute. 

committed to take adequate corrective action by developing supplemental itemspecific checkl ists. 

YE1 management acknowledged this weakness and 

Similar concerns regarding the audit/survey of Threaded Screw Products is 

discussed in Section 3.8 above. 

C. Concl usi on 

The inspector's review of YE1 a supplier quality audit and commercial grade 

survey procedures and checkl ists and selected suppl ier audit/survey reports 

revealed that they were a broad-brush, programmatic review, not performance based. 

qualification of a commercial grade supplier (e.g., placing the supplier on an 

approved commercial grade suppliers list), and may be useful in managing 

suppl i er qual i ty resources , but they were not critical characteri st i c-speci f i c 

and item-specific, and inconsistent with NRC GL 89-02. The auditslsurveys did 

not adequately verify that the suppl iers' commercial qual ity 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 VI1 of 10 CFR Part 50, Appendix B. 

this concern is identified as part of Nonconformance 99900358,359/97-01-04. 

Such surveys/audits could provide a reasonable basis for preliminary 

Accordingly, 

15 

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3.10 Seismic Racks 

a. ScoDe 

To evaluate the requirements that YE1 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-Seismic Racks," dated November 16, 1994, QAP 30.9, 

"Rack Welding requirements and Welder Qualifications," dated October 15, 1982, 

and YE1 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-011, Section 07. 

American Welding Society (AWS) Standard 01.1, as the applicable standard for 

welding the racks 2nd specified the following: 

QAP 30.9 referenced 

0 Filler material to be AWS A5.18 2 70s-3 or equivalent 

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.l, Sections 2.7 and 3.6 

Weld Procedure Specification and Welder Qualification to be in 

accordance with AWS D1.l, Parts 6 and C 

Weld quality as required by AWS 01.1, Paragraphs 5.10.3 and 5.11.2 

Visual inspection in accordance with AWS 01.1, Paragraph 5.6.3 

WeJd 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 tolerartces 

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. YE1 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. Concl usi on 

-152- 

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. 

audits or surveys of the KIM Company not verifying welder qualifications were 

weaknesses in YEI's QA control of special processes which YE1 committed to 

strengthen. 

The lack of colocated specifications and the 

16 
3.11 Internal Audits 

a. ScoDe 

In order to determine if QC performed internal audits to verify the 

effectiveness of the quality assurance progr'tm, the inspectors reviewed 

Qual i ty Procedures Manual (QPM) 4.17, "Interiial Qual i ty Audits, " dated October 

7, 1996, as well as selected internal audit reports. 

b. Observations and Findincls 

According to the records, the internal audit 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 

nonconformances; three of them repeated. The inspectors noted that cwrective 

action for the repeat nonconformances was being implemented, but cull 

compliance had not yet been achieved. The inspectors determined that 

corrective 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. Concl us i on 

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 Traininq 

a. Scope 

The inspectors reviewed selected training records of craftsmen to determi ne i f 

they had pxeived training for the activities they were performing. 

b. Observations and Findinas 

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 

compl ete. 

c. Concl usi on 

The inspectors determined that personnel were trained in the areas in which 

they were working. 

17 

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4 PERSONS CONTACTED 

I.C. Baeringer, Vice President, Engineering 

R.P. Bender, Vice President, Quality Asscrance and Procurement 

C. Claypool, Quality Control Technician 

3. 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 

ODened 

99900358,359/97-01-01 VI0 Informing Affected Licensees of Deviation 

99900358,359/97-01-02 NON Documenting/Di sposi tioning Test Failures 99900358,359/97-01-03 NON Review for Suitability of Application 99900358,359/97-01-04 NON Product Verification/Suppl ier Audi ts/Surveys 

C1 osed 

None. No prior open items. 

Discussed 

99900358,359/97-01-05 IF1 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 

Information Notice 97-45 Environmental Qualification Deficiency for Cables and 

Containment Penetration Pigtails 

Information Notice 97-53 

Information Notice 97-59 

Circuit breakers Left Racked Out In Non-Seismically Qualified 

Positions 

Fire Endurance Test Results of Versawrap Fire Barriers 

-1 55-