NUREG - 0040 

Vol. 20, No. 4 

Licensee Contractor 

and Vendor Inspection 

Status Report 

Quarterly Report 

October - December 1996 

U.S. Nuclear Regulatory Commission I Office of Nuclear Reactor Regulation 
AVAILABILITY NOTICE 

Availability of Reference Materials Cited in NRC Publications 

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

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

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

Washington, DC 20402-9328 

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

20555-0001 

2. 

3. 

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

Referenced documents available for inspection and copying for a fee from the NRC Public 

Document Room include NRC correspondence and internal NRC memoranda: NRC bulletins, 

circulars, information notices, inspection and investigation notices; licensee event reports; 

vendor reports and correspondence: Commission papers; and applicant and licensee 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 

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

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

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

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1001 8-3308. 

A Year’s subscrhtion of this reDorf consists of four ouarterlv issues. 
and Vendor Inspection 

Status Report 

Quarterly Report 

October - December 1996 

WREG-0040 

Vol. 20, No. 4 

Licensee Contractor 

r L 

Manuscript Completed: March 1997 

Date Published: March 1997 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555 -0001 

DISCLAIMER 

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

Government. Neither the United States Gwernment nor any agency thereof, nor any of their 

employees, makes any warranty, express or implid, 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 docs not necessarily constitute or impiy its endorscmenl, mommendation, 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 by the NRC’s 

Special Inspection Branch, Vendor Inspection Section, that have been 

distributed to the inspected organizations during the period from October 1996 

through December 1996. 

iii 

Portions of this document may be illegible 

in electronic image products. Smage~ are 

produced from the best available original 

document. 
CONTENTS 

- PAGE 

Abstract .................................................................. iii 

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

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

ABB-Combusti on Engineering 

Windsor, CT 

C&D Charter Power Systems Incorporated 

Conshohoken, PA 

Dragon Valves Incorporated 

Norwal k, CA 

Framatome Cogema Fuels 

Lynchburg, VA 

Panal arm Business Unit 

Skokie, IL 

Westinghouse Electric Corporation 

Energy Systems Business Unit 

Pittsburgh, PA 

Zetec, Incorporated 

Issaquah, WA 

Sel ect Generi c Correspondence on the Adequacy 

Audits and the Quality of Vendor Products 

V 

(999013@6/96-01) ......... 2 

(99901304/96-01) ......... 14 

(99900264/96-01) ......... 40 

(99900001/96-01) ......... 53 

(99901303/96-01) ......... 94 

(99901307/96-01) ......... 111 

(99901037/96-01) ......... 128 

of Vendor .................. 145 

INTRODUCTION 

A fundamental premise of the U. S. Nuclear Regulatory Commission (NRC) 

licensing and inspection program is that licensees are responsible for the 

proper construction and safe and efficient operation of their nuclear power 

plants. 

for the inspection of commercial nuclear facilities to provide for multiple 

levels of inspection and verification. Each licensee, contractor, and vendor 

participates in a quality verification process in compliance with requiriments 

prescribed by the NRC’s rules and regulations (Title 10 of the Code of federa7 

Regu7ations). The NRC does inspections to oversee the commercial nuclear 

industry to determine whether its requirements are being met by licensees and 

their contractors, while the major inspect-ion effort is performed by the 

industry within the framework of quality verification programs. 

The Federal government and nuclear industry have establ ished a system 

The licensee is responsible for developing and maintaining a detailed quality 

assurance (QA) plan with implementing procedures pursuant to 10 CFR Part 50. 

Through a system of planned and periodic audits and inspections, the licensee 

is responsible for ensuring that suppl iers, contractors and vendors also have 

suitable and appropriate quality programs that meet NRC requirements, guides, 

codes, and standards. 

The Vendor Inspection Section (VIS) of the Special Inspection Branch reviews 

and inspects nuclear steam system suppliers (NSSSs) , architect engineering 

(AE) f i rms , suppl i ers of products and servi ces , independent testing 

laboratories performing equipment qualification tests, and holders of NRC 

construction permits and operating 1 icenses in vendor-related areas. 

inspections are done to ensure that the root causes of reported vendor-re1 ated 

problems are determined and appropri ate correcti ve actions are devel oped. The 

inspections a1 so review vendors to verify conformance with appl icabl e NRC and 

industry quality requirements, to verify oversight of their vendors, and 

coordination between licensees and vendors. 

These 

The VIS does inspections to verify the quality and suitability of vendor 

products, 1 icensee-vendor interface, environmental qualification of equipment, 

and review of equipment problems found during operation and their corrective 

action. When nonconformances with NRC requirements and regulations are found, 

the inspected organization is required to take appropriate corrective action 

and to institute preventive measures to preclude recurrence. 

imp1 ications are found, NRC ensures that affected licensees are informed 

through vendor reporting or by NRC generic correspondence such as information 

notices and bulletins. 

When generic 

vi i 
This quarterly report contains copies of all vendor inspection reports issued 

during the calendar quarter for which it is published. 

report lists the nuclear facilities inspected. This information will also 

alert affected regional offices to any significant problem areas that may 

require special attention. Appendices 1 ist selected bulletins, generic 

letters, and information notices, and include copies of other pertinent 

correspondence i nvol vi ng vendor i ssues . 

Each vendor inspection 

viii 
INSPECT ION REPORTS 

1 
+I)**+ '- 

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 205554001 

December 4, 7996 

Michael F. Barnosk 

President, Nuclear Operations 

ABB-Combustion Eng neering 

2000 Day Hi 1 1 Road 

Wi ndsor , CT 06095-0500 

SUBJECT: NRC INSPECTION REPORT 99901306/96-01 

Dear Mr. Barnoski: 

On October 31, 1996, the U.S. Nuclear Regu 

inspection at your ABB-Combustion Engineer 

presents the results of that inspection. 

atory Commiss on (NRC) completed an 

ng facility. The enclosed report 

The inspection was conducted to ascertain specific attributes and 

implementation of your quality assurance (QA) program, and whether 1 icensees 

effectively monitored your control of qual ity for safety-related 

instrumentation and control (IK) systems and associated spare parts purchased 

by licensees for nuclear power plantJ. 

dedication activities, screening of issues for Part 21 applicability, and your 

monitoring of the control of quality by your subvendors. 

We assessed your commercial-grade 

During this inspection, the NRC inspectors determined that the implementation 

of your quality assurance program did not meet certain NRC requirements 

imposed on you by your customers. Specifically, you failed to implement your 

documented QA instructions tc consider 1 icensee findings concerning the 

quality performance of your subvendor, as pequired by 10 CFR Part 50, Appendix B. 

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

and the circumstances surrounding it are described in detail in the enclosed 

report. 

specified in the enclosed NON when preparing your response. 

Please respond to the nonconformance and follow the instructions 

In addition, the inspectors determined that licensee monitoring of your 

qual i ty assurance program for I&C systems and components was not always 

adequate. 

dedication activities and your screening of issues for Part 21 applicability. 

Licensees did not monitor your review of subvendor commercial-grade 

2 
M. Barnoski -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, 

=w Robert M. Gallo. 

Special Inspection Branch 

Divi;ion of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Docket No. 99901306 

Encl osures : 1. Notice of Nonconformance 

2. Inspect ion Report 99901306/96-01 

cc: See next page 

3 
cc: 

Hr. Ronald Fitzgerald 

Director, Qual i ty Assurance 

ABB-Combustion Engineering 

2000 Day Hill Road 

Windsor, CT 06095-0500 

Mr. Michael P. Lilley 

Manager, Qual i ty Assurance 

Rochester Gas L Electric Corporation 

89 East Avenue 

Rochester, NY 14649 

Mr. Richard H. Fassler 

Supervisor, Qual i ty Assurance 

Ni agara Mohawk Power Corpordt i on 

Nine Mile Point Nuclear Station 

P.O. Box 63 

Lycoming, NY 13093 

Mr. Wallace Woodard 

Pr i nci pal Engi neer , Florida Power & Light Company 

P.O. Box 14000 

Juno Beach, FL 33408-0420 

Mr. Robert Ripley 

Audit Team Leader 

Union Electric Company 

1901 Chouteau Avenue 

P.O. Box 149 

St. Louis, MO 63166 

Mr. Ronald Casavant 

Technical Specialist 

Washington Public Power Supply System 

P.O. Box 968 

Richland, WA 99352 

Mr. James Young 

Section Leadet 

Arizona Public Service Company 

P.O. Box 52034 

Phoenix, A2 85072-2034 

Mr. Derek Mercurio 

Licensing 

Southern Califo ,iia Edison Company 

2244 Walnut Grove Avehue 

Rosemead, CA 91770 

Mr. Joel Rachal 

Procurement Engineer 

Entergy Operations, Inc. 

P.O. Box 31995 

Jackson, MS 39286- 1995 

4 
NOTICE OF NONCONFORMANCE 

ABB-Combustion Engineering (ABB-CE) 

Windsor, Connecticut 

Docket No.: 99901306 

On the basis of an inspection by the staff of the U.S. Nuclear Regulatory 

Comnission (NRC) on September 9 through 12, 1996, and October 28 through 31, 

1996, it appears that the following activity was not conducted in accordance 

wi t h NRC requ i remen t s : 

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 instruct ions and procedures, and 

accomp?ished in accordance with these instructions and procedures. 

ABB-CE's Quality Assurance Manual 100, System 7.0, Section 2.5.4, 

"Control of Purchased Items and Services," Revision 2, requires, in 

part, that the assessment of the overall quality performance of 

suppliers of safety-related items shall include consideration of reports 

of audits from other sources (i.e., customers, NRC, other ABB nuclear 

groups). 

Contrary to the above requirements, ABB-CE failed to consider licensee 

findings regarding commercial-grade dedication activities of its 

subvendor ABB-El ectro Mechanics (ABB-EM) . Two 1 i censees, Rochester Gas 

& Electric Company and Niagara Mohawk Power Corporation, identified 

problems with ABB-EM'S commercial-grade dedication of safety-related 

instrumentation components. ABB-CE's lack of consideration of licensee 

findings precluded it from assessing whether safety-related items it 

purchased from ABB-EM were compromised (99901306/96-01-01). 

Please send a ,.,,itten statement or explanation to the U.S. Nuclear Regulatory 

Comnission, ATTN: Document Control Desk, Washington, D.C. 20555-0001, with a 

copy to the Chief, Special Inspection Branch, Division of Inspection and 

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

date of the letter transmitting this Notice of Nonconformance. Your reply 

should be clearly marked as a "Reply to a Notice of Nonconformance" and should 

contain for the nonconformance (1) a description of steps that have been or 

will be taken to correct these items, (2) a description of steps that have 

been or will be taken to prevent recurrence of these items, and (3) the dates 

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

Dated at Rockvi 

this 4th day of 

le, Maryland 

December, 1996 

Enclosure 1 

5 
Report No: 

Organization: 

Contact: P 

Nuclear Industry 

Activity: 

Dates: 

Inspectors: 

Approved by: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

99901306/96-01 

ABB-Combuction Engineering 

Ronald J. Fitzgerald, Director 

Quality Assurance 

860/285-9816 

Instrumentation and control systems and 

associated spare and rep1 acement parts 

September 9-12 and October 28-31, 1996 

Ani1 S. Gautam, Senior Engineer 

Frank Gee, Electrical Engineer 

Gregory C. Cwaliqa, Chief 

Vendor Inspection Section 

Special Inspection Branch 

Division of Inspection and Support Programs 

Enclosure 2 
1 IJSPECTION SUMHARY 

During this inspection, the NRC inspectors reviewed activities associated with 

implementation of sel ected portions of ABB-Combust i on Engi neeri ng 's (ABB-CE's) 

quality assurance (QA) program and 1 icensee monitoring of ABB-CE's control of 

qual i ty . 

The inspection bases were as follows: 

Appendix B, Assurance Criteria for Nuclear Power Plants and Fuel 

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

Reaulations (10 CFR Part 50). 

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

NRC Regulatory Guide 1.144, "Auditing of Quality Assurance Programs for 

Nuc 1 ear Power Pi ants . 'I 

ABB-CE's Quality Assurance Manual (QAM) 100, Revision 4, dated May 1, 

The inspectors nc+ed one instance in which ABB-CE failed to conform to NRC 

requirements imposed upon it by NRC 1 icensees. 

discussed in Section 3.1 of this report. 

1996, and associated implementing procedures. 

This nonconformance i s 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

Nonconformance 99900538/95-02-01 (CLOSED1 

Contrary to Criterion V of Appendix B to 10 CFR Part 50, and paragraph 

D.2 of ABB-CE Nuclear Operations "Administrative Procedure for Reporting 

Defects and Noncompliance," dated February 28, 1996, ABB-CE did not 

document its screening of potEntial Part 21 issues. 

ABB-CE responded to the notice of nmconformance (NON) by letter of April 

26, 1996, stating that it would make appropriate changes to its Part 21 

imp1 ement i ng procedure. ABB-CE "Procedure for Eva1 uating and Reporting 

of Defects and Noncompliance Pursuant to 10 CFR Part 21," dated September 

30, 1996, addressed the process of screening and reporting of potential 

Part 21 issues by all Nucle.ar Systems and Nuclear Operations employees. 

The procedure requires each deviation or potential failure LO comply to 

be evaluated, the evaluation documented, and the disposition recorded. 

The inspectors found the revised procedure acceptable. 

3 INSPECTION FINDINGS AND OTHER COMMENTS 

3.1 Qualitv Assurance Proqram 

a. InsDection ScoDe 

The inspectors examined ABB-CE's QA program, pol icy, implementing 

procedures, conformance to procurement documents, corrective actions i n 

2 

7 
response to licensee audit findings, commercial-grade item dedication, 

Part 21 evaluations, and nonitoring of subvendors. 

b. Observations and Findinqs 

ABB-CE’s nuclear operations comprised five business units, including 

Total Quality, Engineering, Field Services, Fuel Operations, and 

Techno1 ogy Development . Each business unit was monitored by qual i ty 

control (QC) inspectors. 

QA managers. The QA director reported to the Vice President of Totcli 

Quality who reported to the President of Nuclear Operations. The 

inspector observed that QAM-100 System 1, Figure 1.2, indicated that the 

QC inspectors reported to the management of the individual business units 

rather than to QA. 

QAM-100 chart, the QC inspectors were monitored by QA. 

The inspector wit,iessed, in part, testing of a safety-related core 

cooling monitoring system for Millstone Unit 2 and observed equipment 

being subjected to various inputs to validate system performance and 

conformance to licensee procurement specifications. No concerns were 

i dent i f i ed . 

The QA staff comprised the QA director and four 

The QA management stated that irrespective of the 

The inspector evaluated ABB-CE’s audit reports EM-9605 and EM-9606, dated 

August 28, 1996, of its subvendor ABB-EM and determined that the audit 

reports did not provide adequate objective evidence of quality furnished 

by the subvendor. For example, the reports did not address compliance 

with Appendix B criteria, internal audit findings and corrective actions, 

negative observations or weaknesses, and any deficiencies identified by 

other customers. ABB-CE’s QAM-100, System 7.0, Section 2.5.4, “Control 

of Purchased Items and Services,” Revision 2, requires, in part, that the 

assessment of the overall quality performance of its suppliers of safetyrelated items shall include consideration of reports of audits from other 

sources (i .e., customers, the U.S. Nuclear regulatory Commission (NRC), 

other ABB nuclear groups). Durinq 1995-1996, Rochester Gas & Electric 

Company (RG&E) and Ni agara Mohawk Power (NMP) Corporation, purchased 

instrumentation directly from ABB-EM and identified problems with ABBEM’S commercial-grade dedication of safety-related instrumentation 

components. Licensee tindings included ABB-EM’S failure to identify or 

test critical characteristics of test jacks, diodes, and other printed 

circuit board (PCB) components. ABB-CE’s lack of consideration of 

1 icensee f,odings precluded it from assessing whether ABB-EM provided 

reasonable assurance that dedicated items will perform their intended 

safety function. This practice constitutes Nonconformance 99901306/96- 

01-01. The QA manager stated that the ABB-CE audit reports of ABB-EM 

were not typical and that QA planned to include additional details in its 

future audit reports. 

During 199“ 1996, ABB-CE purchased scfety-related instrumentation from 

ABB-EM for sale ’i3 licensees. -ABB-EM purchased these items as 

comnercial-grade and dedicated them for safety-related applications. 

During the NRC inspection on September 9-12, 1996, ABB-CE could not 

provide the inspector sufficient documentation to demonstrate that 

3 

8 
selected items were qualified to perform their safety functions. During 

the inspection on October 28-31, 1996, two inspectors visited the ABB-CE 

(Windsor, Connecticut) and ABB-EM (New Britain, Connecticut) facilities 

to further assess commercial-grade dedication of selected safety-related 

instrumentat ion The inspectors assessed selected purchase orders (POs) , procedures, 1 icensee surveillance, design drawings and documentation 

pertinent to commercial-grade dedication activities. The following ABBCE POs to ABB-EM and associated items were examined: 

- POS 

404676 

40661 0 

406260 

405516 

404700 

404489 

405651 

405842 

9607 145 

9405001 

9503 180 

ITEMS 

PCB 

Re1 ay 

Re1 ay and operational amp1 i f i er 

Switch 

Variable set-point card 

Test jack 

PC B 

Card assembly for combinational circuit 

Fiber optic switch 

Repair of bistable trip unit 

Diagnostic testing of setpsint card 

Upon the basis of commercial-grade dedication documents, the inspectors 

determined that ABB-CE and ABB-EM reviewed safety functions, identified 

appropriate critical characteristics, and performed tests and analyses to 

dedicate the aforementioned items. No concerns were identified. 

The inspector reviewed Part 21 reports issued by ABB-CE for design errors 

or failures during the past 3 years. 

nonconformance reports (NCRs) did not address Part 21 screening, 

applicability, and reportabil ity, and there was no mandatory training for 

the staff on Part 21 programmatic activities. 

identified in a previous inspection (see Inspection Report 99900538/95- 

02). 

implementing procedure to address documentation of screening on QA forms 

(including noncompliance forms), and self-study QA training for employees 

(see Section 2 of this report). 

The inspector observed that 

This concern was also 

In response to these concerns, ABS-CE revised its Part 21 

The inspector examined a summary of 1993-1996 NCRs concerning 

instrumentation and observed no trending of failures by QA to evaluate 

root causes in order to prevent recurrence of failures. The QA director 

provided evidence that an initiative was in place to improve trending of 

failures and expected it to be implemented by about January 1997. 

The inspector observed that there was no sharing of Part 21 information, 

subvendor audits, or hardware problems among ABB-CE and other ABB 

companies. For example, ABB-CE purchased items from ABB-Transmission and 

Distribution (ABB-T&D) but was not a4tare of failures of type MG-6 

electrically reset 1 atching re1 ayy (documented in NRC Report 50-280/96-05 

and 50-281/96-05, dated July 15, 1996). 

CE had not purchased the MG-6 relay, but that ABB-CE planned to expand 

the scope of its questions during subvendor audits. 

The QA manager stated that ABB4 

9 
The inspector observed two unmarked boxes in the holding area for safetyrelated items and asked whether the boxes contained any non-safety parts 

that could compromise the segregation of safety-related parts. 

inspector examined the boxes and confirmed that the shipping boxes 

contained safety-re1 ated parts and that the storage area was control 1 ed 

by receipt inspection and restricted access. 

The QC 

c. Conclusions 

The inspectors concluded that, in general, the QA manual, work 

instructions, and procedures were adequate, except for the nonconformance 

and screening of potential Part 21 issues described herein. 

3.2 Review of Monitorinq of ABB-CE by Licensees 

a. InsDection ScoDe 

The inspector evaluated licensee monitoring of ABB-CE’s control of 

quality for safety-related items purchased by licensees, including audits 

and surveillances of ABB-CE’s commercial-grade dedication, Part 21 

reports, and monitoring of subvendors. 

b. Observations and Findinss 

The inspector contacted the Nuclear Uti1 ities Procurement Issues 

Committee (NUPIC) audit team leakrs and selected licensees to discuss 

scope and findings of the audits and surveillance. 

NUPIC audited ABB-CE in October 1993 and October 1995. Utilities 

represented on the NUPIC audit teams included Arizona Public Service 

(APS) Company, Baltimore Gas & Electric Company, Consolidated Edison 

Company, Consumers Power Company, Entergy Operations (EO), Florida Power 

and Light (FP&L) Company, Illinois Power Company, Northeast Utilities, 

Pennsylvania Power & Light Company, Rochester Gas & Electric Corporation 

(RG&E), Union Electric Company, Virginia Electric & Power Company, 

Washington Pub1 ic Power Supply System, and Yankee Atomic Electric 

Company, 

NUPIC’s 1993 audit identified a finding concerning ABB-CE’s failure to 

identify critical characteristic- of certain commercial-grade items 

during ABB-CE’s dedication of instrumentation and control items. In 

response to this finding, ABB-CE developed procedure No. 00000-ICE-0509, 

“Technical Guidelines for Commercial-Grade Dedication,” to require 

identification of critical characteristics and specify form, fit, and 

function of dedicated items. 

concerning ABB-CE’s failure to use an approved vendor for calibration of 

measuring and test equipment (M&TE). The NUPIC audit team accepted ABBCE’s corrective actions. 

The NUPIC team also identified a finding 

NUPIC’s October 1995 audit identified findings concerning deficiencies in 

ABB-CE’s monitoring of subvendors, design calculation packages, and 

control of M&TE. ABB-CE’s monitoring of its subvendor MTS Systems 

Corporation (supplier of calibration services) was deficient in that ABBCE did not verify whether MTS implemented its out-of-tolerance5 

10 
C. 

notification program, established integrity of traceability 

documentation, or performed internal audits. NUPIC accepted ABB-CE’s 

corrective actions. 

APS audited ABB-CE in March 1995. The APS audit identified several 

weaknesses in ABB-CE’s implementation of its design control program 

including documenting design assumptions, identifying design interfaces 

between ABB-CE and APS engineering, verifying and identifying software, 

controlling computer output documentation, and maintaining QA records in 

Chattanooga, Tennessee. 

f i nd i ng s . 

FP&L audited ABB-CE in June 1995. 

concerning ABB-CE’s failure to monitor the adequacy of the design control 

activities conducted by its subvendor, and ABB-CE’s basing its 

engineering judgments on unverified design inputs. 

CE’s corrective actions for these findings. 

Based on FP&L surveillance reports 08.06.CEPSG.94.1, 95.1, 95.5, and 

96.1, FP&L conducted effective surveillances of ABB-CE and ABB-EM. 

Activities monitored incl uded commerci a1 - grade dedication activities 

(review of critical characteristics and verification), nonconformances, 

and M&TE calibration data. 

ABB-CE is in the process of resolving th? 

The FP&L audit identified findings 

FP&L accepted ABBDiscussions with EO and Southern California Edison procurement staff 

determined that they relied on NUPIC’s audit for monitoring ABB-CE and 

did not conduct any surveillance of ABB-EM. 

The inspector determined that NUPIC and the licensees did not audit the 

subvendor ABB-EM. 

identify that ABB-CE performed inadequate audits of ABB-EM (see section 

3.1.b of this report). 

Further, NUPIC and licensee’s audits of ABB-CE did not 

The insFectors observed that licensets, in general, did not identify 

specific safety functions of spare parts purchased from ABB-CE. POs 

identified spare part numbers and pertinent systems, not pertinent 

instrumentation circuits or parameters monitored. 

they did not ask licensees to identify specific safety functions, rather 

qualified the spare parts to the original design of the system. Lack of 

identification of specific safety functiocs by licensees could result in 

inadvertent mi sappl ication of procured i tems. 

Conclusions 

ABB-CE stated that 

In general, 1 icensees audited ABB-CE in accordance with proper criteria, 

procedures, and checklists. Licensee monitoring of ABB-CE’s QA program 

for instrumentation components was not always adequate: 1 icensee audits 

did rqt ensure that ABB-CE was contrl:lling the quality of their purchased 

i tems from ABB-EM. 

6 

11 
3.3 Entrance and Exit Meetinqs 

4 

In the entrance meeting on September 9, 1996, the NRC inspector discussed 

the scope of the inspection, outlined the areas to be inspected, and 

established interactions with ABB-CE management. In the exit meetings on 

September 12 and October 31, 1996, the inspectors discussed their 

findings and observations. 

PERSONNEL COMT ACT ED 

ABB-CE 

Don Allen, Vice President, Total Quality 

Ron Fitzgerald, Director, QA 

Mark Stewart, QA 

Gary Bloomqui st, Manager, Qual i ty Programs and Procedures 

Richard Bradshaw, Manager, Instrumentation & Controls Engineering (ICEE) 

Robert Oriscoll, Supplier Audits 

Edward Stepaneck, QC Inspector, ICEE Assurance 

Ian Richard, Manager, Operations Licensing 

Charles Molnar, Licensing Engineer 

Paul Rohan, Total Quality 

Mark Stewart, Supervisor, Spare parts QC 

Kenneth Tomany, ICEE QC 

Marty Ryan, ICEE 

Larry Bryan, Procurement Engineer, Nuclear Spare parts 

ABB-Electro Mechanics 

William Hadovski, Engineering Manager 

Alex Oja, Contract Manager 

William Wayland, Supervisor, QA Engineering 

Edward Roll i ns, Pl ant Manager 

K.Parekh, Senior Engineer 

M.J. Merlini, Contract Administrator 

Licensees (contacted by telephone1 

Mike Li.ley, QA Manager, RG&E Company 

Wall ace Woodard, Principal Engineer, FP8L Company 

Olga Hanek, Lic;,lsing Engineer, FP&L Company 

George Kuhn, Procurement Engineer , FP8C Company 

Robert Ripley, 1993 NUPIC Team Leader, Union Electric Company 

Ron Casavant, 1995 NUPIC Team Leader, Washington Public Power Supply System 

Derek Mercurio, Licensing, Southern California Edison Company 

Joel Rachal, Procurement Engineer, EO 

7 

12 
ODened 

99901306/96-01-01 

C1 osed 

9990U538/95-02-01 

ITEMS OPENED, CLOSED, AND DISCUSSED 

NON Failure to assess subvendor’s control of 

qual i ty 

NON Inadecuate Part 21 screening prqcess by 

ABB-CE 

8 

13 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-oOol 

' +**** 

December 17, 1996 

Dr. Leslie S. Holden 

Vice President Technology 

C&D Charter Power Systems, Inc. 

Washington & Cherry Streets 

Conshohocken, PA 19428 

SUBJECT: NRC INSPECTION REPORT 99901304/96-01 

Dear Dr. Holden: 

On October 9, 1996, thy U.S. Nuclear Regulatory Commission (NRC) completed an 

inspection at the C&W Charter Power Systems, Inc. (C&D) facilities at 

Attica, Indiana and Conshohocken, Pennsylvania, regarding your lead-acid 

battery cells that are manufactured arid supplied to nuclear power plant 

facilities for Class 1E electrical qqFety-related applications. The enclosed 

report presents the results of the inspection. 

During this inspection, the NRC inspectors determined that certain activities 

appeared to be in violation of NRC requirements. 

staff stated that your batteries were dedicated at the end of manufacturing 

and assembly activities, C&D failed to document the dedication methodology and 

basis. 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 its response. The NRC will use the C&D response, in part, to 

determine whether further enforcement action is necessary to ensure compl iance 

with regulatory requirements. 

Specifically, although C&D 

Please note that C&D is required to respond to this letter 

In addition, the NRC team determined that the implementation of the C&D 

quality assurance program failed to meet certain NRC requirements imposed on 

C&D by your customers. 

program applied to the manufacture of batteries intended for installation in 

nuclear power plant Class 1E electrical systems. The team observed that, even 

though w-itten procedures and instructions had been established to delineate 

manufacturing process controls at the Attica facility, craftsmerl were 

typically not provided nor made aware of these procedures and instructions. 

For example Nonconformance 99901304/96-01-05 revealed that C&D a1 1 owed a 

manufacturing procedural requirement to be performed differently than 

specified. 

The team found certain weaknesses in C&D's quality 

This failure of C&D to comply with its quality system program may have 

contribytcd to the degraded battery cc.'ls identified at the Hatch nuclear 

plant. 

the benefit of written procedures or instructions. 

The team also identified that seismic analyses were performed without 

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

14 
Dr. L.S. Holden -2- 

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

and C&D should follow the instructions specified in the enclosed NON when 

preparing your response. 

The team concluded that your batteries provided for 1 icensee safety-related 

applications may not have specifically met the 1 icensee's procurement 

requirements because C&D did not establish or implement a battery cell 

dedication program tQ ensure that each predetermined critical characteristic 

was verified under a 10 CFR Part 50 Appendix B quality program. 

the team determined that uti1 ity company auditors did not adequately evaluate 

CAD'S qual ity system program implementation because CAD characterized its 

battery cell manufacturing activities as comnercial grade to these auditors. 

In accordance with 10 CFR 2.790 of the NRC "Rules of Practice," a copy of this 

letter and enclosures will be placed in the NRC's Public Document Room (PDR). 

Additionally, 

Docket No 

Sincerely, 

Original signed by 

Robert M. Ga'llo, Chief 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

: 99901304/96-01 

Enclosures: 1. Notice of Violation 

2. Notice of Nonconformance 

3. Inspection Report 99901304/96-01 

cc: Mr. T.J. Kinden, Director Quality A;surance Mr. C. Dale Brown 

C&D Charter Power Systems, Inc. P1 ant Manager 

Washington & Cherry Streets 

Conshohocken, PA 19428 200 West Main Street 

C&D Powercom 

Post Office Box 279 

Attica, IN 47918 

15 
NOTICE OF VIOLATION 

CSD Charter Power Systems, Incorporated Docket No.: 99901304 

B1 ue Bel 1, Penrlsyl vani a 

During an NRC inspection conducted on September 16 through 18 and October 7 

through 9, 1996, a violation of NRC requirements was identified. 

accordance with the "General Statement of Policy and Procedure for NRC 

Enforcement Actions," NUREG-1600, the violation is 1 isted below: 

In 

Section 21.21, "Notification of failure to comply or existence of a defect and 

its evaluation," of 10 CFR Part 21 requires, in part, that a dedicating entity 

is responsible for maintaining auditable records for the dedication process. 

Further, Section 21.3, "Definitions," of 10 CFR Part 21 requires, in part, 

that in all cases, the dedication process must be conducted in accordance with 

the applicable provisions of 10 CFR Part 50 Appendix B. 

Contrary to the above, C&D did not establish and maintain auditable records 

for the dedication process used to dedicate battery cells destined for Class 

1E safety-related station battery applications. Adequate objective evidence 

was not provided to indicate that C&D's battery cell dedication process was 

being conducted in accordance with the applicable provisions of Appendix B to 

10 CFR Part 50. (99901304/96-01-01) 

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

Pursuant to the provisions of 10 CFR 2.201, C&D Charter Power Systems, 

Incorporated, is hereby required to submit a written statement or explanation 

to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, 

Washington D.C. 20555-0001, with a copy to the Chief, Special Inspection 

Branch, Division of Inspection and Support Programs, Office of Nuclear Reactor 

Regulation, within 30 days of the date of the letter transmitting this Notice 

of Violation. This reply should be clearly marked as a-"Reply to a Notice of 

Violation" and should include for each violation: 

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 ful i compl iance 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. 

(1) the reason for the 

Dated at Rockvill e, Mary1 and 

this 17th day of December, 1996 

16 Enc osure 1 
NOTICE OF NONCONFORMANCE 

C&D Charter Power Systems, Incorporated 

B1 ue Bel 1, Pennsyl vani a 

Docket No. : 99901304 

Based on the results of an inspection conducted September 16 through 18 and 

October 7 through 9, 1996, it appears that certain of your activities were not 

conducted in accordance with the requirerrents of the U. S. Nuclear Regulatory 

Comission. 

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

10 CFR Part 50 states that, activities affecting quality shall be 

prescribed by documented instructions, procedures, or drawings, of a 

type appropriate to the circumstances and shall be accomplished in 

accordance with these instructiops, procedures, or drawings. 

Instructions, procedures, or drawings shall include appropriate 

quantitative or qualitative acceptance criteria for determining that 

important activities have been satisfactorily accomplished. 

Section 3.0, "Responsibilities and Authorities," of C&D Procedure 

01-003.6, "Document and Data Control (Corporate), 'I January 1996, states 

in part, that department managers and supervisors are responsible to 

assure that the correct documents are available at the locations where 

needed and that they are being used. 

A.1 Contrary to the above, assembly line and manufacturing personnel in 

several areas at the Attica, Indiana facility were not aware of any 

procedures or instructions to control the work activities which they 

were performing. 

ensure that procedures or instructions were being used in areas such as 

plate pasting, pasted plate curing, cell assembly, post sealing and cell 

cover gluing . (9990l304/96-01-02) 

The Attica facility supervisors and managers did not 

A.2 Contrary to 

without the 

A.3 Contrary to 

battery cel 

Appendix 8, 

B. 

the above, C&D performed its site specific seismic analyses 

benefit of written procedures. (99901304/96-01-03) 

the above, although C&O stated that it was performing 

C&D did not establish a dedication procedure to identify and 

verify the adequacy of each of the critical characteristics important to 

ensure that the battery cells would perform satisfactorily in service. 

dedication for customers who imposed 10 CFR Part 50 

(99901304/96-01-04) 

Criterion 1 I I, "Design Control, I' states that, measures shall be 

established to assure that the design basis for those components that 

Appendix B applies are correctly trt2ns8ated into specifications, 

drawings, procedures, and instruct ions I These measures shall i ncl ude 

provisions to assure that appropriate quality standards are specified 

and included in design documents and that deviations from such standards 

are control 1 ed. 

Enclosure 2 

17 
Criterion V, Instructions, Procedures, and Drawings," requires that 

activities affecting quality shall be accomplished in accordance with 

instructions, procedures and drawings. 

Section 5.3.4 of CAD Product & Process Specification (PPS)-X4, 

Revision 3, states: Immediately following cover installation, the 

cell/unit shall be held in place for a minimum of 30 minutes with a 

pressure equal to 80-100 psig. The cover shall be held to the jar in a 

way to eliminate movement and to distribute the pressure uniformly over 

the entire seal ing area. 

B.l Contrary to the above, although C&D had translated a design requirement 

into Section 5.3.4 of Procedure PPS-X4, it did not ensure that the 

activity was accompli shed in accordance with the procedural requirement. 

Consequently, deviations from the design basis contained in the C&D 

Procedure were not controlled for that work operation. 

( 9990 1304 / 96-0 1-05) 

Please provide a written statement or explanation to the U.S. Nuclear 

Regulatory Comnission, ATTN: D~cilwrlt Control Desk, Washington, D.C. 20555- 

0001, with a copy to the Chief, Special Inspection 8ranch, Division of 

Inspection and Support Programs, Office of Nuclear Reactor Regulation, within 

30 days of the date of the letter transmitting this Notice of Nonconformance. 

This reply should be clearly marked as a "Reply to a Notice of Nonconformance" 

and should include the following for each nonconformance: (1) a description of 

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

description of steps that have been or will be taken to prevent recurrence of 

these items; and (3) the dates your corrective actions and preventive measures 

were or will be completed. 

Dated at Rockville, Maryland, 

this 17th day of December, 1996 18 
U.S. NUCLEAR REGULATORY COMMISSION 

Report No.: 

Organization: 

Contact Address: 

Nuclear Industry 

Act i vi ty : 

Dates of Inspection: 

Inspectors: 

Approved by: 

OFFICE OF NUCLEAR REACTOR REGULATION 

9990 1304/96-01 

C&@ Charter Power Systems, Incorporated * 1400 Union Meet i rlg Roaa 

B1 ue Bel 1 , Pennsyl vani a 19422-0858 

Terry J. Kinden, Director of Quality Assurance 

Washington & Cherry Streets 

Conshohocken, PA 19428 

(610) 825-2150 ext. 245 

Class 1E nuclear power plant station batteries, 

battery chargers, and seismically qualified station 

battery racks. 

September 16-18, 1996 @ Attica, Indiana, and 

October 7-9, 1996 @ Conshohocken, Pennsylvania 

Kamalakar R. Naidu, Senior Reactor Engineer 

Yueh-Li C. Li, Mechanical Engineer 

Joseph J. Petrosino, Q.A. Specialist 

Saba N. Saba, Electrical Engineer 

Gregory C. Cwalina, Chief 

Vendor Inspection Section 

Speci a1 Inspecti on Branch 

Division of Inspection and Support Programs 

Off ice of Nucl ear Reactor Regul at i on 

Enclosure 3 

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

14 
1 INSPECTION SWHARY 

During this inspection, the NRC inspectors reviewed the implementation of 

selected portions of C&D Charter Power Systems, Incorporated (C&D) quality 

assurance (QA) program, reviewed activities associated with the manufacture of 

lead-acid stationary batteries used in Class 1E applications, and reviewed .. pertinent design and engineering documents. 

The inspection bases were: 

0 Appendix By "Quality Assurance Cri Leria for hucle 

Fuel Reprocessing Plants," to Part 50 of Title IO 

Reaulations (10 CFR Part 50) 

b 10 CFR Part 21, "Reporting of Defects and Noncomp 

0 The American National Standard Insti tute/The Amer 

r Power Plants and 

of the Code of Federal 

i ance" 

can Society of 

Mechanical Engineers (ANSI/ASME) Standard N45.2-1977, "Qual i ty Assurance 

Program Requirements for Nuclear Facilities" (ANSI N45.2-1977), as 

endorsed by U. S. Nuclear Regulatory Commission Regulatory Guide 

(RegGuide) 1.28, "Quality Assurance Program Requirements (Design and 

Construction) ,'I Revision 2, February 1979 (RegGuide 1.28) 

The team found that C&D's quality program for its control of nuclear power 

plant Class 1E station batteries, at the Attica, Indiana facility, contained 

weaknesses regarding the consistency of assembly and manufacturing process 

controls as a result of not providing its assembly and machine operators with 

the required procedures or instructions as required. The team determined that 

even though procedures were established for C&D's manufacturing craftsmen, and 

C&D's quality system program required that they be used; the team identified 

that in many manufacturing areas the craftsman were not aware nor being 

provided with the required procedures/instructions as discussed in 

Section 3.6. 

A correlation to this process control inconsistency was seen when the team 

reviewed a C&D laboratory inspection report that discussed a phenomena 

regarding prematurely degraded LCY-35 battery cells in Class 1E safety-related 

systems at the Edwin I. Hatch nuclear plant (Hatch). 

investigation that was delineated in the laboratory inspection report prompted 

C&D to state that the excess sedimentation suggested that the cause was due to 

"improper positive plate processing, " most 1 ikely because of "improper or 

incomplete steam curing with resultant weak active material structure." 

The team was informed that C&D dedicated3 its battery cells and associated 

components at the Attica facility after all manufacturing and assembly 

activities were completed. 

time of a successful battery capacity discharge test, performed in accordance 

with guidance in The Institute of Electrical and Electronics Engineers, 

Incorporated (IEEE) Standard-450, "Recommended Practice for Maintenance, 

The results of the C&D 

C&D staff stated that dedication occurred at the 

~ ~~ 

As defined in Section 21.3 of 10 CFR Part 21. 

2 

20 
Testing, and Replacement of Large Lead Storage Batteries for Generating 

Stations and Substations”-1987. 

process be conducted in accordance with an Appendix B to 10 CFR Pzrt 50 

quality assurance program, the team found that C&D had not documented or 

established a battery cell assembly dedication process or program as required 

by NRC regulations. 

Although Part 21 requires that the dedication 

Additionally, since C&D claimed to dedicate their batteries after manufacture, 

it would be expected that the batteries would be manufactured as commercial 

grade items [CGIs), and that all of the critical characteristics would be 

verified upon dedication. However, even though C&D characterized its battery 

cells as being handled as CGIs to the NRC inspection team and NRC licensee 

auditors, as discussed in Section 3.3, it appeared that C&D was taking credit 

for in-process inspections and verifications during design and manufacturing 

act i v i t i es . 

The team also noted that C&D’s battery cell design for safety-related 

applications is controlled under C&D’s quality system program that was 

established to meet ANSI N45.2-1977. 

assessed the adequacy of C&D’s quality system program manual (QSPM) 

implementation since it appeared that the adequacy of the Class 1E batteries 

manufactured at Attica received the benefit of being processed in accordance 

with C&D’s documented QSPM controls. 

As a result of this finding, the team 

During this inspection, two violations of 10 CFR Part 21 were identified and 

are discussed in Sections 3.2 and 3.3 of this report, one was a non-cited 

violation. Four instances where C&D Charter Power Systems, Incorporated 

failed to conform to NRC requirements (10 CFR Part 50 Appendix B) imposed upon 

them by NRC 1 i censees were a1 so ident i f i ed. The nonccnformances are di scussed 

in Sections 3.6, 3.7 and 3.9 of this report. 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

This was the first NRC inspection of C&D’s facilities at Attica, Indiana, and 

Conshohocken, Pennsylvania. 

3 INSPECTION FINDINGS AND OTHER COHHENTS 

3.1 Oualitv Assurance Proqram 

a. ScoDe 

The inspection team (team) reviewed the establishment and implementation 

of selected portions of the C&D quality program, which was documented in 

the C&D Quality System Program Manual a The QSPM stated that its purpose 

was to establish the basic operating policies and procedures to be 

employed by C&D, and to meet applicable requirements of International 

Organization for Standardization (IS01 Standard 9001-1994, ANSI rNl45.2- 

1977 and other imposed specificatiorls. 

3 

21 
The team was apprised by C&D that it assures that its components conform 

to the specified requirements through implementation of C&D's quality 

system program that is delineated in the QSPM. The QSPM's foreword 

explains that C&D's quality system program is implemented through the 

QSPM, departmental standard operating procedures, work instructions, 

drawings, bills of materials, material specifications, process 

specifications, and qual i ty and test procedures. Section I, "Management 

Responsibility," of the QSPM indicates that C&D management has the 

responsi bil i ty and authority to ensure that the pol icy and objectives 

defined in the QSPM, and its surporting procedures and instructions are 

understood, implemented, and maintained at all levels in the company. 

b. gbservations and findinss 

The team noted that the C&D facilities that are under C&D's quality 

system program are specified in the QSPM, specifically: (1) Corporate 

Headquarters-Blue Bel 1, PA-Control and maintenance of design and 

qualification aspects of all C&D safety-related components; (2) Attica, 

IN-Class 1E station batteries; (3) Dunlap, TN-Class 1E station battery 

chargers; (4) Conshohocken, PA-qeismically qualified, Class 1E station 

battery racks; (5) Leola, PA-Class 1E station batteries, round cell; 

(6) Conyers, GA-commercial products; (7) Huguenot, NY-commercial 

products ; and (8) Ratel co , Incorporated, Seat tl e, WA-el ectroni cs , including battery charger assembl ies used in safety-related applications 

of C&D battery chargers. 

facilities manufacture and covtrol components that are destined for 

safety-related applications at nuclear power plants. 

The team ascertained that six of the eight 

Since six of the eight C&D facilities manufacture components that are 

used in safety-related applications, the team looked at the overall QSPM 

program establishment to determine whether a quality system that 

addresses 10 CFR Part 50 Appendix B requirements such as, ISO-9001 and 

ANSI N45.2, was in existence and controlled for the eight C&D facilities 

that supplied components for safety-re1 ated applications. 

The team observed that C&D's Policy Implementation stated that the C&D 

Quality Program has been designed to assure compliance with ISO-9001 

(1994), ANSI lNl45.2 (1977), and CSA-9001. However, a1 though the team 

was informed by C&D staff that each facility complied to ANSI N45.2- 

1977, the team determined that only 2 of the 8 facilities had been 

cel ,ified to ISO-9001 and none cf the 8 facilities had been certified to 

CLk 3001. 

The team was also informed that all of C&D's Attica facility 

manufacturing, in-process verification and assembly activities were 

considered commercial grade, and that C&D dedicated each station battery 

cell for Class 1E safety-related application at the time of successful 

final testing using guidance contained in IEEE Standard-450-1987. 

Based on that position, C&D considered Attica's manufacturing and 

assembly activities that occurred before the IEEE-450 test to be 

activities that did not fall under an Appendix B QA program even though 

4 

22 
its QSPM was established to meet ANSI N45.2-1977 requirements and its 

certificates of compliance for Class 1E batteries attested to compliance 

with licensee purchase orders (which imposed Appendix B) and C&D's QSPM. 

c. Conclusions 

The team observed that C&D' s qual i ty system program manual was yeneral ly 

developed to follow the requirements of ANSI-N45.2-1977, which can, if 

established and implemented in accordance with NRC RegGuide 1.28, 

"Quality Assurance Requirements (Design and Construction) ,If meet the 

requirements of Appendix B. 

The team concluded that C&D's position regarding battery dedication 

would have been acceptable had C&D complied with appropriate regulatory 

and industry guidance regarding dedication. 

the battery cell critical characteristics be identified and their 

acceptability verified by appropriate inspection, tests, 0.r analysis. 

In addition, the dedication process must be conducted and controlled in 

accordance with the applicable provisions of 10 CFR Part 50 Appendix B. 

That guidance requires that 

The team found that C&D had not established nor implemented a dedication 

program to comply with NRC regulations or industry guidance. The team 

did not find, nor was it provided with any battery cell dedication 

procedure documents. Additionally, the team was not provided with any 

documented basis of C&D's position regarding battery dedication. 

Further, the team determined that the "dedication" test was indicative 

of the battery capacity at the time of the test but did not adequately 

demonstrate the battery's functionality for its guaranteed 1 ife. 

Therefore, the team determined that several manufacturing steps were 

crucial to demonstrate that the battery would meet purchase order 

requirements. 

or controlled in accordance with Appendix B. Consequently, the team 

concluded that this quality assurance program area was not well 

established or implemented in accordance with 10 CFR Part 21 and 10 CFR 

Part 50 Appendix 6, and industry guidance. 

Those steps need to be verified by the dedication process 

The team also found that one area of the QSPM, regarding work 

instructions and procedures, was not well implemented at the Attica 

facility. The team's findings in this area are detailed in Section 3.6. 

3.2 

a. 

JO CFR Part LI Proqram 

Scooe 

The inspectors evaluated the procedure adopted by C&D to implement the 

requirements of 10 CFR Part 21 by reviewing Standard Policy and 

Procedure Number A-14-4, "Reporting of Defects and Nonconformances in 

accordance with Federal Regulation 10 CFR 21 (US Nuclear Regulatory 

Commission)," dated March 18, 1995. 

5 

23 
b. Observations and findinas 

After discussing the implementation of the Part 21 responsibility in 

detail with the C&D-QA Director, the team informed him that Procedure 

A-14-4 was inconsistent with the regulation. 

establish the requirements specified in 521.21 constitutes a violation 

of minor significance and is being treated as a non-cited violation, 

consistent with Section IV of the NRC Enforcement Manual (NUREG-1600). 

The failure to adequately 

Based upon discussions with the C&L staff the team was concerned 

regarding C&D compliance with other Part 21 requirements. 

the team discussed requirements that are contained in 921.21(b) of 

10 CFR Part 21 with C&D staff that require a supplier of basic 

components to inform purchasers or affected 1 icensees within five 

working days if the supplier is unable to perform an evaluation of a 

deviation or failure to comply. 

Director that its procedure did not define or reference 10 CFR Part 21 

specific terms such as, discovery and evaluation. 

Sp,ecifically, 

The team also informed the C&D QA 

As discussed in Section 3.4 of this report, over 40 of the 120 safetyrelated station battery cells that C&D manufactured and supplied in late 

1993-early 1994 to the Hatch plant exhibited premature degradation. The 

battery was degraded due to excessive sedimentation but continued to be 

operable according to the licensee. Therefore, the team reviewed the 

Hatch event with regard to C&D's responsi bil ity under 10 CFR part 21. 

The team noted C&D's June 1996 laboratory inspection report of the cell 

degradation indicated "an improper or incomplete steam curing." C&D 

staff stated that it reviewed its manufacturing and quality control (QC) 

records during that time period and did not identify other potentially 

affected customers. C&D further stated that no other customers 

(commercial or nuclear) with similar battery cells from that time period 

have reported similar problems. 

efforts regarding the Hatch issue. 

The team had no concerns with C&D's 

c. co ncl usions 

The team concluded that C&D's procedure adopted to implement Part 21 was 

not adequately established and lacked clarity. Specifically, the team 

found that the procedure did not address the preparation of an interim 

report if an evaluation of a deviation or failure to comply cannot be 

completed within 60 days of discovery. 

Additionally, the team concluded that C&D was not required to inform the 

NRC or other customers of the manufacturing process control weakness, as 

discussed in Sections 3.6 and 3.7, since it determined that no other 

customers were affected during its review of the circumstances of the 

probl em. 

6 

24 
3.3 Dedication of Class 1E Station Batteries 

The team found that the control of the manufacture of C&D’s battery 

cells and associated components could be viewed as being implemented in 

one of the following two ways: (1) performing dedication activities of 

its Class 1E battery cells, as conditionally permitted by 10 CFR 

Part 21, in accordance with the applicable requirements of Appendix B to 

10 CFR Part 50, or (2) manufacturinq its Class 1E battery cells as basic 

components, under its QSPM. 

b. Observations and findinas 

C&D staff informed the team that its station batteries were processed in 

their manufacturing and assembly areas as comnercial grade items (CGIs) 

and were not considered basic components until after each battery cell 

passed a test (which C&D considers its dedication activity) that C&D 

performs to meet guidance contained in IEEE-450-1987. Although, due to 

a lack of adequate documentation and conflicting information, the team 

was not able to determine how C&D does, in fact, produce its safetyrelated batteries (whether by dedication or by manufacture in accordance 

with Appendix B), the team discussed C&D’s dedication position to 

determine its compliance with NRC requirements. 

the manufacturing process for compliance with Appendix B, as discussed 

elsewhere in this report. 

The team also reviewed 

Additionally, the team noted that the C&D safety-re1 ated battery cell s 

are typically used at licensee facilities for back-up electrical 

requirements for instrumentation and control and other power needs to 

various safety-related circuits, inverter power and emergency and normal 

Class 3E vital bus applications. Typical critical characteristics of 

lead-acid battery cells that would be required to be verified include 

characteristics such as: ampere-hour capacity; cell voltage; dimensions 

and configuration; case and component materi a1 s, such as paste mixture 

for positive and negative plates; and grid, plate, or electrode 

construction and assembly, such as internal plate assembly-to-post 

connections. 

Section 21.3 of 10 CFR Part 21 states that dedication is an acceptable 

process undertaken to provide reasonable assurance that a commerci a1 

grade item will perform its intended safety function and, in this 

respect, is deemed equivalent to ari item designed and manufactured under 

a 10 CFR Part 50, Appendix B, quality assurance program. 

This assurance is achieved by identifying the critical characteristics 

of the item and verifying their acceptability by inspections, tests, or 

analyses, and that the dedication process must be conducted in 

accordance with the applicable prov‘sions of Appendix B. Section 21.3 

also states that a commercial grade item means a structure, system, or 

component, or part therefore that affects its safety function, that was 

not designed and manufactured as a basic component. Commercial grade 

7 

25 
items [CGIs] do not include items where the design and manufacturing 

process require in-process inspections and verifications to ensure that 

defects or failures to comply are identified and corrected (i.e., one or 

more critical characteristics of the item cannot be verified). Further, 

Section 21.21 of 10 CFR Part 21 requires a dedicating entity to maintain 

auditable records for the dedication process. 

The team determined that C&D: (1) did not establish or have a dedication 

procedure or program that complied with the applicable provisions of 

ApFendix B, or 10 CFR Part 21, ;2) did not identify or verify critical 

characteristics that were necessary to perform an adequate dedication 

process, (3) did not take exceptions to licensee purchase order 

requirements that imposed Appendix B to 10 CFR Part 50, and (4) 

indicated on its certificates of compliance that its Class 1E batteries 

are in compl iance with the appl icable 1 icensee purchase ordei-s and 

specifications which the team found to typically impose Appendix B to 10 

CFR Part 50 and C&D’s QSPM. 

Further, the team noted that C&D appeared to take credit for the 

implementation of manufacturing activities at Attica under its QSPM 

program for compliance to Apptdix B to 10 CFR Part 50, as attested to 

in its CoCs. Conversely, C&D appeared to characterize the same 

manufacturing activities to the NRC inspection team and licensee 

auditors, that occurred before IEEE-450 testing, as being commercial 

processes not subject to Appendix B. The team also noted that C&D’s 

dedication position of its battery cell assembly dedication had been 

found to be acceptable by recent licensee audit group inspections. 

c. Conclusion 

The team did not find any objective evidence to indicate that C&D was 

manufacturing its battery cells destined for Class 1E applications as 

commercial grade items followed by dedication. The team concluded that 

even though Section 21.21 of 10 CFR Part 21 requires that a dedicating 

entity is responsible for maintaining auditable records for the 

dedication process, and Section 21.3 requires that the dedication 

process must be conducted in accordance with 10 CFR Part 50 Appendix B, 

C&D did not have specific records to delineate its battery cell assembly 

dedication or to substantiate that its activities were carried out in 

accordance with 10 CFR part 50 Appendix B. The team identified 

Vi ’ation 99901304/96-01-01 in this area. 

The team concluded that although C&D stated that it was performing 

battery cell dedication for customers who imposed Appendix B to 10 CFR 

Part 50, C&D did not establish a dedication procedure to identify and 

verify the adequacy of each of the critical characteristics important to 

ensure that the battery cells would perform satisfactorily in service. 

Tnerefore, C&D did not ensure that its battery cell dedication 

activities affecting quality were prescribed and accomplished in 

accordance with documented instructions as required by Criterion V, 

8 

26 
3.4 

a. 

b. 

"Instructions, Procedures, and 

10 CFR Part 50. The team ident 

this area. 

Drawings," of Appendix B to 

f ied Nonconformance 99901304/96-01-04 in 

Additionally, the lack of a formalized dedication program to address 

critical characteristics and the lack of an effectively implemented 

Appendix B to 10 CFR Part 50 type of a quality system program could 

affect the overall life and functionality of C&D's battery cells that 

are currently i n-use at operating nucl ear power pl ant faci 1 it i es. 

Batteries for Hatch Nuclear Power Piant 

ScoDe 

The inspectors reviewed the circumstances surrounding premature 

degradation of over 40 of the 120 C&D LCY-35 type battery cells that it 

manufactured and supplied to Hatch in 1994-1995. 

team's review included: (1) C&D documentation to support the LCY-35 

design; (2) licensee purchase order requirements; (3) C&D's 

documentation substantiating C&D's certificate of compliance (CoC) for 

Hatch; and (4) the C&D analysis of the failure of the batteries supplied 

to Hatch. 

The scope of the 

Observations and findinqs 

Backqround In April 1996, the NRC was informed by the Georgia Power 

Company (GPC) that excessive sedimentation was observed in 21 of 127 

LCY-35 type batteries supplied by C&D, and by May 1996, approximately 43 

battery cells exhibited excessive sedimentation. The batteries were 

supplied by C&D in 1994 to the Hatch nuclear power plant (Hatch). 

conmi tted to the 1 icensee to rep1 ace the prematurely degraded battery 

cell s with new battery cell s. 

C&D 

Desicln Verification The inspectors reviewed the discharge 

characteristics curves of LCY (35 through 39 plates), and LCUN-33 (33 

plates) type batteries at C&D's facility in Conshohocken, PA. These 

batteries are used in switch-gear and! control applications at commercial 

nuclear power plants and are tested utilizing test procedures meeting 

the IEEE 450 standards. The inspectors reviewed the data for a battery 

with voltage terminal of 125 volts, 1.215 electrolyte specific gravity 

at 77°F (25°C) and found it acceptable. 

The team reviewed test data of a typical duty cycle of LCY-35 and LCY-37 

cells to verify that the cell voltage at4the one-minute discharge rate 

included the effect of the Coup de Fouet . The team's review of the 

test data confirmed that all the one-minute rates did include the Coup 

de Fouet. 

A phenomena of additional battery voltage drop when the battery is 4 

subjected to a discharge after long duration float charges. 

9 

27 
The inspectors reviewed the implementation of the engineering change 

notices (ECNs) and engineering change requests (ECRs). A typical ECR 

gave detailed information on the reasohs for the change and its impact 

on manufacturing, whereas, a typical ECN referred to the ECR, and was 

sent to different departments, including the manufacturing facilities. 

The team did not identify any concerns in this area. 

The inspectors established that one of the principal differences between 

Class 1E and comnercial grade cells is the separator. In Class 1E 

cells, C&D uses a rubber separator, whereas, 5n the commercial battery 

cell s C8D uses a microporous polyetnyl ene separator. The separator 

appeared to be the only component difference between the commercial and 

Class 1E battery cells. The installation of the rubber separator is an 

example of a manufacturing activity that requires inspection or 

verification during dedication or control by Appendix B quality program. 

Hatch Purchase Order The inspectors reviewed GPC’s PO 60135700000, 

dated December 1, 1994, to C&D for the supply of LCR-25, and LCY-35 

125/250 Vdc, lead calcium, 120 cell batteries and accessories for Hatch. 

The PO included the following requirements: 

Applicable portions of 10 CFR 50, Appendix B, apply 

Items shall be supplied new and not used or refurbished in any way 

10 CFR 21 applies to safety related nuclear items/services 

CoC certifying the material/components meet the requirements of 

Specification NP-93024, Revision 2, Section 3.A.4. 

Certificate’s of ComDliance (CoC) 

dated February 25, 1994, for Hatch and determined that it certified that 

127 LCY-35 wet cells with accessories were brand new and had been 

manufactured in accordance with the requirements of specification NP93024, Revision 2 and the quality standards specified therein, and that 

it met IEEE-450-1987 capacity acceptance test. 

459 amperes for a discharge time of 4 hours and the acceptance volts per 

cell was 1.75 average. The Capacity Discharge test sheets were attached 

to the CoC. These test sheets indicated that the average volts per cell 

exceeded 1.75V at the end of the test. 

The inspectors reviewed CAD’S CoC 

The discharge rate was 

The inspectors noted that C&D issued a CoC certifying that the batteries 

met all the GPC requirements implying that C&D’s QA program complied 

with 10 CFR Part 50, Appendix B, even though C&D was aware that its 

instructions and procedures affecting the quality of the components 

supplied were not comprehensive enough in some areas. For example, 

procedures and instructions that did not adequately address all of the 

Attica facility work activities, as discussed in Section 3.6. 

A review of a sample population of nther NRC licensee purchase order 

packages at the Attica facil ity determined that 1 icensees are procuring 

Class 1E battery cells and accessories on POs which typically impose 

10 

28 
Appendix B to 10 CFR Part 50, 10 CFR Part 21 and the nuclear related 

quality program that is delineated in C&D's QSPM (ANSI-N45.2-1977). All 

of the C&D CoCs reviewed in the different PO packages stated compliance 

with the applicable 1 icensee PO and 'I icensee specification. 

Additionally, the CoCs contained other compliance statements as 

required, such as, "a1 1 batteries/cell s are manufactured in accordance 

with the [licensee] PO requirements, C&D's quality system program policy 

manual, meeting the applicable requirements of 10 CFR 50, Appendix 6, 

ANSI N45.2-1977, and other applicable industry standards (e.g., IEEE 

450-1987, and IEEE 535-1986)." In the PO packages reviewed, the 

inspectors observed that C&D did not take any exceptlons to any of the 

licensee contractual requirements or additional requirements imposed on 

them, nor did any of the CoCs or QSPM discuss or address a dedication 

basis as discussed in Section 3.3 of this report. 

Service InsDection ReDort C&D examined the excessive sedimentation in 

the Hatch batteries. 

looks like sponge deposits, appeared to be the result of active plate 

material being shed due to contamination or impurities introduced into 

the plate material at the factory. The batteries were considered 

operable based on their tested performance to deliver over 116% of rated 

capacity. On June 25, 1996, after inspecting the batteries, C&D 

transmitted its findings and included C&D's battery laboratory 

inspection report. 

on C&D's analysis of one battery cell from Hatch, stated: 

According to C&D the excessive sediment, which 

The inspectors reviewed the report which was based 

The individual cell voltage was more than 2.29 V per cell. 

The positive and negative grids were in good condition. 

The positive plate active surface material was depleted (shed) with 

approximately 50-100% of the plate surface being affected. 

The negative active material was soft. 

There was no mossing present either on sides of the frames or the 

top. No sulfaration was evident. 

The sediment chamber was filled with shed active material that had 

oxidized after making contact with the feet on the negative plate. 

The glass mat was saturated at the bottom. 

The plate lug strap burns were good. 

Corrosion was normal on the positive post and strap assembly. 

29 

The report also stated Hatch's LCY-35 battery suffered from 

sedimentation, due to improper positive plate processing, and the 

condition of the positive plate active materi-1 suggested improper or 

incomplete steam curing with resultant weak active material structure. 

11 
During discussions with C&D personnel, the team was informed that C&D is 

currently in the process of considering additional design enhancements 

and process control changes to control the curing ovens more effectively 

for uniform curing of the positive plates. 

The team concluded that C&D’s laboratory report results and the team’s 

observations of weaknesses in C&D’s process control suggests that other 

Class 1E batteries that C&D provided in this same time frame may also 

prematurely degrade. Therefore, this aspect was discussed with the C&D 

Quality Assurance Director regadsding whether C&D should be considering 

informing its applicable customers of a potential deviation that needs 

to be monitored and evaluated by certain licensees. However, as 

discussed in Section 3.2, C&D investigated this issue and concluded that 

it was not applicable to other licensee battery cells. 

c . Concl us i on 

In general, the inspectors observed that the C&D adequately controlled 

the design of the batteries through ECNs and ECRs, including technical 

and engineering notices. 

The inspectors determined that C&D’s CoC implied that C&D’s QA program 

complied with 10 CFR Part 50, Appendix 8, even though CAD was aware that 

its instructions and procedures affecting the quality of the components 

supplied did not meet the requirements of 10 CFR Part 50 Appendix B or 

ANSI N45.2-1977. 

3.5 Manaqement Audits 

The inspectors reviewed the implementation of the QA program relative to 

management audits. The plant manager performed audits to satisfy the 

requirements of the QA Manual. Paragraph 1.5 of Section 1.0 of the C&D 

QSPM requires executive management to review the quality system program 

at defined intervals in accordance with Management Review procedures. 

This review is intended to ensure the continuing suitability and 

effectiveness of the established Quality Management System and the 

Company’s stated quality policies and objectives. As an integral part 

of the process, confirmation as to preventive actions and results of 

audits are to be submitted. Standard Operating Procedure (SOP) 01-022.2 

outlines the requirements to imDlement the above. 

r6,:ires the plant manager to cmduct a quality system program review on 

- +arterly basis one month prior to the end of the fiscal quarter or 

sooner as indicated by necessity or by direction of corporate Quality 

counsel. 

Paragraph 6.3 

The inspectors reviewed the plant manager’s audits performed on December 

31, 1995, and March 29, 1996, and determined that the audits covered the 

following: the need to develop additional process instructions and work 

instructions, the need to revise training for hourly employees, 

enhancements in process improvement programs and quality improvement 

programs, and status of corrective actions to a previous external audit. 

Copies of the audit were distributed to the “Corporate Qual i ty Counsel ‘I 

12 

30 
which consists of: the President/CEO, the Director of QA, the VP of 

Finance, Power Com., Motive, Technology, Power Com Operations, and the 

Director of Motive Products Services. The team noted that the audit 

recognized the necessity to develop procedures to control the 

manufacturing process as required by C&D’s quality system program. 

3.6 Manufacturinq Process Control 

a. ScoDe 

The team observed the process controls used by C&D in the manufacture of 

their batteries and related components. 

was to have all of its facilities certified to ISO-9001. The main 

document which establ ishes the basic operating policies and procedures 

employed to meet the requirements of ISO-9001 is the QSPM. 

states that the C&D quality system program is implemented through the 

use of operating procedures, work instructions, drawings, qual i ty and 

test procedures, etc. At the time of the inspection, the primary 

documents used to control manufacturing activities were Product & 

Process Speci f i cat i ons ( PPSs) and Factory Procedures ( FPs) . 

The team found that C&D’s goal 

The QSPM 

b. Observations and Findinss 

Adeauacv of Drocedure and instructions The team found that in 

approximately 1993-1994, when the current C&D QA Director assumed his 

current position, he recognized that the existing PPSs and FPs that had 

been established for many of the C&D facilities would not support ISO9001 certification. The team was told that the PPSs and FPs were 

written to encompass multiple facilities and, as a result, were not 

specific to the battery manufacturing process control. 

Therefore, C&D started a program to establish specific instructions and 

procedures for each particular work activity at each of its facilities 

to replace the existing PPSs and FPs. 

The replacement documents for the PPSs and FPs are process instructions 

(PIS) and operating/work instructions (01s). 

the PIS and 01s for the Leola facility had been completed but C&D had 

just begun to develop PIS and 01s for its Attica and Dunlap facilities. 

The team reviewed some of C&D’s draft 01s and PIS for the Attica 

facility and conducted discussions with C&D process engineering 

personnei who are responsible for developing these documents. The team 

noted that the new 01s and PIS appear comprehensive and easy to use, and 

the team was informed that the new work instruction documents have the 

banefit of receiving input from the personnel that are performing the 

work activity. 

PIS and 01s at Attica and Dunlap in the Fall of 1998. 

The team was informed that 

The QA Director stated that C&D expects to complete the 

31 

use of procedures and ingtructions Discussions with manufacturing 

personnel revealed that the personnel were experienced and know1 edgeabl e 

regarding their respective areas. 

was apparently obtained from comprehensive on-the-job training (OJT) 

The team found that this knowledge 

13 
programs. 

instructions were used. The team found that craft personnel typically 

were not aware of specific procedures or work instructions that 

controlled the activities that they performed, with the exception of 

those working in the paste mixing area. 

During the discussions, the team asked wha'. procedures and 

Section 3, "Responsibilities and Authorities,'' of C&D's QA Procedure 01- 

003.6, "Document and Data Control (Corporate) , 'I states that department 

managers are responsible for generating, approving, distributing and 

maintaining work instructions for their department. 

"department managers and supervisors are responsible to assure that the 

correct documents are available at the locations where needed and that 

they are being used". 

Additionally, 

c. Conclusions 

The team determined that C&D's quality system program of procedures and 

instructions was not adequately established to ensure effective and 

consistent manufacturing processes control. That determination is 

supported by C&D's effort to replace all of its PPSs and FPs. The team 

concluded that this weakness was not controlled in accordance with 

corrective action requirements of ANSI-N45.2-1977, 10 CFR Part 50 

Appendix B or C&D's QSPM, which has been established to address ANSI 

N45.2-1977. Although this matter could have been cited as a 

nonconformance being contrary to licensee requirements imposed on C&D, 

the team did not identify this matter as a noncompliance to 10 CFR Part 

50 Appendix B because the QA Director had identified the problem and had 

initiated corrective action. 

The team found that C&D has been aware of the practice of not providing 

instructions and procedures and has been attempting to correct the 

problem by establishing specific 01s and PIS for each of its facilities 

since approximately 1992-1993. However, the team noted that until this 

activity is complete, C&D's Attica activities are contrary to Criterion 

V, "Instructions, Procedures, and Drawings," of Appendix B to 10 CFR 

Part 50, and Section 6, "Instructions, Procedures and Drawings," of 

ANSI-N45.2-1977, regarding implementation of QSPM. Nonconformance 

99901304/96-01-02 was identified in this area. 

3.7 Observation of Work Activities 

a. JnsDection ScoDe 

The team inspected several areas at the Attica battery cell and 

associated product manufacturing and assembly facility including: paste 

mixing, paste application, curing, hydrosetting and drying, cell 

assembly, and leak testing. The team observed the manufacturing process 

control activities that were being performed and conducted discussions 

with crafts personnel. As discussed in Section 3.1, C&D's quality 

14 

32 
system program has been established to encompass all activities 

affecting product conformance to specified requirements through its 

system of operating procedures, xork instructions, drawings, bills of 

materi a1 , speci f icat i ons, and qual i ty aid test procedures. 

b. Observations and Fi ndi nqs 

Paste mixinq The inspectors observed the process of mixing paste for 

application on the grids. 

according to two tables observed to be in a C&D procedure; one table was 

for the positive plate and the other table was for the negative plate. 

These tables provided information on the components to be used, the 

materials specification, the quantities and tolerances, and the 

acceptance criteria for the resulting paste. 

paste mixing area was the only area at the Attica manufacturing facility 

where the craftsmen were observed to have a procedure or work 

instruction available for the craftsmen. 

The paste operator formulated the paste 

The team noted that the 

Paste amlication Paste is applied to the grids by gravity flow from a 

hopper to the grid’s pasting operations where it is applied to the 

positive and negative grids. Finished grids are visually inspected for 

non-uniform pasting. 

inspection are run a second time through the process and if the 

non-uniformity persists, the grid is rejected and scrapped. Following 

pasting, positive plates are sent to the curing area while negative 

plates go directly to hydrosetting. 

Grids which are rejected during the initial 

The inspectors observed the pasting of positive plates. 

period, a few plates which C&D personnel observed to be irregular were 

put through the process a second time. 

specific run appeared to operate smoothly with few slow-downs and stops 

during the period observed. 

were moved to the curing process area. 

During this 

The production rate of this 

After the pasting was completed, the plates 

Curing In this area, the positive plates were subjected to a humidity 

curing process (also known as steam curing). Curing is performed to 

obtain mechanical strength, grid paste adhesion, porosity and proper 

lead oxide crystal structure. The plates were cured in an oven required 

to be maintained at certain humidity and temperature levels specified by 

a chart. 

The inspectors observed that the only parameters of the curing that are 

recorded are the oven temperature and humidity. 

installed to announce any failure of the heating or the humidity 

sources. 

were not mechanically controlled. 

weakness in C&D’s curing process because plate curing is one of the 

critical and important stages in the manufacture of the lead acid 

batteries. 

dry, crack-free material with sufficient stre,,gth and adhesion to the 

grid plate. 

1 ongevi ty of the battery cell s. 

An alarm has not been 

The inspectors also observed that the access to the oven doors 

The inspectors observed this as a 

The aim of the curing F~-OC~SS is to convert the wet paste to 

Improperly cured plates can affect the performance and 

33 

15 
Hvdrosettina and drvinq Depending on the type of the lead oxide used in 

the manufacture of positive plates, they are either dried imnediately or 

subjectsd to a hydroset process before drying. 

hydroset before drying. 

All negative plates are 

Hydroset curing is performed at high humidity and relative low 

temperature to assure the required grid adhesion, mechanical strength 

and porosity. During the hydrosetting process, the free lead content of 

the active material is reduced to specific levels depending on the 

application of the battery. 

All types of plates are subjected to a drying process at 215 k 5°F. 

final moisture content depends on the type of plates. 

during several times of observing this area that the overhead doors are 

periodically left open without any obvious reconciliation or 

verification of the humidity and temperature levels. 

The 

The team noted 

Cell assembl y The inspectors observed craft-persons assemble different 

battery cell configurations (such as 4XTL13 and KCR9, that are used in 

safety-related applications) . The batteries were assembled in sequence 

including: (1) stacking, (2) plate-to-strap bonding, (3) element 

insertion into the container, (4) cover installation, (5) terminalconnector post-to-cover insert bonding, and (5; leak testing. 

The team noted that PPS-X4-3, "Standby Cell Assembly," March 28, 1991, 

described the assembly operat ion and emphasized proper a1 ignment of 

plates and separators. 

to the appearance, performance, and life of the battery cell. The 

element assembly consisted of two distinct procedures: stacking and 

plate to strap bonding. A stacked element consists of alternating 

negative and positive plates, interleaved with separators. 

The team determined these aspects were critical 

After the plates were stacked so that the lugs are aligned, the positive 

and negative burning lugs were on the opposite sides. The plate was 

then bonded to the strap and visually inspected for voids, porosity, 

entrapped dross, lead run downs, and poor bonding. (At this stage the 

assembly is cleaned if necessary.) 

raised the element into the vertical position and inserted support combs 

and moss shields. The element fitted snugly into the battery container. 

The cover was then glued on the container. 

After acceptance, the operator 

I$c team observed that after the glue was inserted onto the mating 

surface of the cover at the cover-to-jar mating area, the cover was 

placed on the jar, and tapped into place with a mallet. The craftperson then placed a wooden frame over the cover, and placed a banding 

type of nylon strap with a ratchet tightening device around the entire 

jar/cover assembly. The craft-person hand-tightened the ratchet until 

the strap appeared to be snug. 

station for some undetermined amount of time to allow the glue to dry. 

The cell was allowed to sit at that work 

16 

34 
The team observed that the above activity was contrary to the 

requirements of PPS-X4-3. 

controlled this area (Section 5.3.4 of PPS-X4-3), stated: "Immediately 

following cover installation, the cell/unit shall be held in place for a 

minimum of 30 minutes with a pressure equal to 80-100 psig. The cover 

shall be held to the jar in a way to eliminate movement and to 

distribute the pressure uniformly over the entire sealing area." 

The procedural requirement that supposedly 

As a result of comparing the actual work practice and the procedural 

requirement, the team noted that: '1) although the banding strap 

appeared to be firmly tightened, trte craftsmen did not have any 

indication of the actual pressure that was applied each time that the 

operation was performed, and (2) it was not obvious whether CAD'S wooden 

frame actually "distributed the pressure uniformly." The inspectors 

discussed the process with the craft personnel and determined the 

craftsmer! were not aware of any requirement for a specific pressure to 

be applied nor were they aware of a procedure that they were supposed to 

be using for that area of the assembly line. 

The team asked C&D management about the applied pressure requirement and 

was informed that the requirement had been modified in practice, but it 

was left in Procedure PPS-X4-3 because the procedure was scheduled to be 

replaced with the new 01s and PIS. The team asked C&D engineering if it 

had approved the deletion of the requirement because it appeared to the 

team that it was a design requirement affecting the operability and 

service life of the battery. C&D QA staff informed the team that C&D 

engineering had approved the removal of the specified pressure 

requirement. However, the team was subsequently told at the 

Conshohocken facility that engineering had not approved the 

discontinuance of the PPS specified activity. 

the battery area stated that the work activity step was a C&D 

engineering requirement and that no engineering approval had permitted 

the requirement to be discontinued. 

The C&D Chief Engineer in 

The team concluded that C&D management failed to ensure that craft 

personnel were supplied the procedure that provided the specific design 

requirement. That failure resulted in craft personnel implementing a 

change to the manufacturing process which, in turn, resulted in an 

unreviewed change to the design requirements. 

that this was contrary to Criterion 111, "Design Control," of Appendix B 

to 10 CFR Part 50, which requires that design changes be subject to 

design con+, dl measures commensurate with the original design and 

approved by the organization that performed the original design. 

inspectors were unable to determine how long this practice was 

prevalent. 

area. 

The team informed C&D 

The 

Nonconformance 99901304/96-01-05 was identified in this 

beak testina of cells The team noted that C&D procedures require 

operators to perform leak tests on all cells using a specific pressure 

for several seconds to c;.ifirm that the container will not leak. The 

team also noted that final formation of the cells, the electrolyte 

17 
filling of cells, charging currents and duration of tSe different 

charges are also discussed in C&D procedures. 

procedures were in the plant supervisor’s office. 

The team found these 

Overall, during discussions with the craft personnel, several of the 

craft personnel admitted that they were not aware of any specific work 

instructions or procedures for their specific job function, but relied 

instead on the OJT programs and co-worker experience and 

assistance.Genera1 ly, the team noted several weaknesses regarding C&D’s 

control of their manufacturing proress. Primary among those was the 

unreviewed design change to the battery assembly requirement. 

inspectors were also concerned with the lack of control for the curing 

oven temperature and humidity, including lack of process alarms and lack 

of control for the oven doors. In addition, the lack of control of 

manufacturing area overhead doors during certain manufacturing steps 

could also affect required humidity and temperature levels. 

The 

c. Conclusions 

The team concluded that craft personnel were knowledgeable and appeared 

to be experienced in their job. 

personnel , it was apparent that knowledge and experience was due 

primarily to OJT. 

discussed in Section 3.6. One nonconformance was identified in this 

area. 

Based upon discussions with craft 

The failure to use work procedures or instructions is 

3.8 CaDacitv Discharqe Tests 

The team observed capacity discharge tests in progress on a short 

duration and long duration battery to confirm that battery cell 

capacities met the design specifications. 

designed to supply relatively high currents for a short period of time. 

These type of batteries, used in UPS applications at nuclear plants, are 

manufactured as off-the-shelf items. 

designed to supply relatively small loads over a much longer period of 

time, typically eight hours. Long duration batteries are manufactured 

for a specific order. The team was informed that C&D monitors 

individual cell voltages during the discharge test to identify 

potentially defective cell s. 

A short duration battery is 

A long duration battery is 

15-minute discharae rate of a short duration battery The team noted 

that the test was successfully performed for the full 15 minutes. The 

team observed that the first low cell voltage alarm occurred at 15 

minutes and 40 seconds. The test was terminated at 17.10 minutes from 

the start indicating a capacity of over 114%. 

8-hour discharae rate of a loncl duration battery During observation of 

the test, the team noted that, when the test monitor indicated a low 

voltage on a cell, the test operatcr was able to identify the faulty 

cell using a hand-held voltmeter and confirmed that it was a bad wire 

18 

36 
3.9 

a. 

b. 

connection. The connection was repaired and the voltage on the cell 

returned to normal. The test continued thereafter without interruption. 

The test was not completed until aftc{- the inspection. 

The inspectors noted that C&D had adequate instrumentation to monitor 

the individual battery cell voltages and the discharge currents during 

the tests. 

rating of the battery before shipment. The inspectors had no concerns 

in this area. 

C&D routinely performed these tests to establish a benchmark 

SEISMIC OUAL I F I CATION REPORT 

ScoDe 

To assess C&D's seismic qualification basis for its LCR-29 and LCY-35 

batteries installed at Hatch, the team reviewed the seismic 

qua1 ification portions of C&D's Report QR-66171-01, "Environmental and 

Seismic Qualification Report of TYPE LCR-29 and LCY-35 Battery for Hatch 

Nuclear Plant," dated January 22, 1993. The team also reviewed Wyle 

laboratories, Seismic Simulation Test Reports 43450-1, "Seismic 

Simulation Test Program on a Battery Rack and Batteries," December 7, 

1976, and 44467-1, "Seismic Simulation Test Program on a Battery Rack 

Containing Two LC-25 Battery Cells and a Battery Rack Containing A 

3DCU-5 and 3DCU-7 Battery Cells," March 1979. 

Observations and findings 

The performance requirement specified in the report for the batteries is 

that the battery shall be capable of supplying the design 'loads for the 

required durations, without the voltage at the terminals falling below 

210 volts. The battery shall be capable of supplying design loads while 

experiencing any single or combination of the following normal and 

design service conditions: Temperature-77°F annual average, 65°F 

mi n i mum, 1 10 O F max i mum; Pres sure- a tmos pher i c ; Rad i at i on-1 es s than 10 

rads, 40 year total integrated dose; and Seismic-OBE & DEE Seismic 

Spectra for turbine building elevation 116'-0." 

The seismic qualification for the 125/250 Vdc lead-acid storage battery 

assembly for the LCR-29 and LCY-35 batteries at Hatch plant was provided 

in accordance with the requirements of Georgia Power Company Purchase 

Order 6010129, dated December 16, 1992, as well as the requirements of 

IEEE Standard 344-1975. 

of test and analysis. 

The basis for qualification was by combination 

Several seismic qualification tests were previously performed for 

various "L" type battery cells at Wyle Laboratories. 

performed in accordance with IEEE 344-1975 and the results were found to 

be well documented. The seismic sinlulation test included five Operating 

Base Earthquake (OBE) level tests prior to oIi: Design Base Earthquake 

(DBE) test. In addition to the seismic simulation test, pre-seismic and 

post-seismic battery discharge capacity tests were also performed. 

The tests were 

19 

37 
The seismic qualification of the Hatch LCR-29 and LCY-35 was 

demonstrated by similarity to those previously successfully tested 

battery assemblies with the same design and configuration. A detailed 

description of the similarity evaluation was found and reviewed. The 

evaluation included comparison of the battery design with emphasis on 

material, dimensions, weights, aging mechanism, battery service 

conditions, and service life. The comparison that was reviewed 

indicated that the design of the LCR-29 and LCY-35 battery assembly was 

equal or better than the originally qualified battery assembly. 

The evaluation also indicated that the seismic acceleration experienced 

by the specimen assembly enveloped the seismic response spectrum. 

Therefore, the loads experienced by the tested specimen enveloped the 

design loads for the plant Hatch batteries. 

The evaluation reviewed by the team was based on the well-designed LCR29 and LCY-35 battery assemblies and its data base did not include 

battery cells with excess amounts of sedimentation similar to the 

existing LCY-35 battery at Hatch nuclear plant. Therefore, the staff’s 

inspection findings and conclusions are not applicable to those existing 

batteries at Hatch. 

The team observed that C&D did not have a documented procedure for 

performing the seismic qualification by similarity. 

CAD management that Criterion V of Appendix B to 10 CFR Part 50 requires 

safety-related activities, such as the preparation of the Seismic 

Report, be performed in accordance with written procedures. The team 

emphasized that in order to assure consistency and quality of seismic 

qual if i cat ion reports, it i s necessary to have a documented procedure 

which includes appropriate quantitative or qualitative acceptance 

criteria for determining that important technical aspects concerning the 

similarity have been satisfied. 

The team informed 

The team also noted that C&D was not required by Hatch to perform any 

specific seismic simulation test with C&D’s Hatch battery to ensure that 

the existing Hatch battery will remain functional during and after a 

seismic event. 

c. Conclusions 

Thn inspectors concluded that C8.D has documented an adequate technical 

basis for the seismic qualification for the LCR-29 and LCY-35 battery 

assembly, and no concerns in this area were identified. 

The team concluded that because C&D did not have a formal dedication 

program to verify battery cell critical characteristics it was unclear 

as to how CLD had ensured that it has maintained its seismically 

qualified configurations during its manufacturing process control. 

20 

38 
The team apprised C&D that failure to have a written procedure for 

performing seismic analysis was contrary to Criterion V of Appendix B to 

10 CFR Part 50. Nonconformance 99901304/96-01-03 was identified in this 

area. 

3.10 PERSONS CONTACTED 

C&D Charter Power Systems. Inc.. Attica. Indiana 

t C. Brown, Plant Manager 

t J. DeSutter, General Supervisor, Casting 

t 8. Donavan, General Supervisor, West Side 

t W. Foster, General Supervi sor, Pas ti ng 

t M. Guthrie, Qual ity Control Manger 

R. Keller, Process Engineer 

t W. Lucas, Material Control 

C&D Charter Power Svstems, Inc., Conshohocken. Pennsylvania 

D. Heimer, Manager, Design & Documentation 

E. Urbanski, Manager, Process Engineering 

F. Wagner, Chief Engineer 

to*+ T. Kinden, Director, Qual i ty Assurance 

0 A. Williamson, Manager, Product Test Laboratories 

*+ C. Wood, Coordinator Supplier Quality 

C&D Charter Power Systems, Inc., B1 ue Bel 1. Pennsvl vani a 

a G. Mal ker, Manager, Applications Engineering 

+ = Attendance at entrance meeting on 9/16/96 

* = Attendance at exit meeting on 9/18/96 

= Attendance at entrance meeting on 10/7/96 

# = Attendance at exit meeting on 10/9/96 

21 

39 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. n#55Qoo1 

Octokr 29, 1996 

Mr. Robert E. Bond, President 

Dragon Valves, Inc. 

13457 Excelsior Drive 

Nomalk, CA 90650 

SUBJECT: NRC INSPECTION REPORT 99900264/96-01 AND NOTICE OF NONCONFORMANCE 

Dear Mr. Bond: 

On September 26, 1996, the U.S. Nuclear Regulatory Commission (NRC) completed 

an inspection at your Dragon Valves, Inc., (Dragon) facility. The enclosed 

report presents the results of that inspection. 

The NRC inspectors found that the implementation of the Dragon quality 

assurance program failed to meet certain NRC requirements imposed on Dragon by 

its customers. Specifically, the procedures for upgrading unqualified source 

material were not controlled and did not comply with the applicable ASME Code 

requirements, Dragon had accepted material certifications without the required 

qual i ty system program statements, the procedure for dedicating customer 

defined safety-related components did not provide a basis for the specified 

chemical sampling plan, and Dragon had not documented any actions for a 

significant condition adverse to quality, as required by 10 CFR Part 50, 

Appendix 6. 

These nonconfo,\mances are cited in the enclosed Notice of Nonconformance 

(NON), and the circumstances surrounding them are described in detail in the 

enclosed report. 

should follow the instructions specified in the enclosed NON when preparing 

your response. 

You are requested to respond to the nonconformances and 

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 (POR). 

Sincerely, 

Robert M. Gallo, Chief 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation Docket No. 99900264 

Enclosures: 1. Notice of Nonconformance 

2. Inspectign Report 99900264/96-01 

40 
NOTICE OF NONCONFORMANCE 

Dragon Valves, Inc. 

Noma1 k, Cal i forni a 

Docket No. 99900264 

Based on the results of an inspection conducted on September 23 through 26, 

1996, it appears that certain of your activities were not conducted in 

accordance with NRC requirements. 

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

10 CFR Part 50 and Paragraph NCA 3853.2 of Subsection NCA of ASME Code, 

Section 111, requires, in part, that activities affecting quality be 

prescribed by documented instructions or procedures and that these 

procedures include appropriate criteria for determining that these 

activities are satisfactorily accolrlplished. 

Paragraph NCA 3853.3, "Document Control" of Subsection NCA of ASME Code, 

Section 111, requires, in part, that instructions and procedures be 

controlled to assure that correct documents are being used. 

Paragraph NCA 3855.5, "Utilization of Unqualified Source Material", of 

Subsection NCA of ASME Code, Section I11 states, in part, that a 

Material Organization (MO) may accept certification of the requirements 

of the material specification which must be performed during melting, 

heat analysis, and heat treatment from an unqualified supplier and may 

furnish such material, providing that the MO performs or subcontracts a 

product analysis on each piece of such material and performs or 

subcontracts all other requirements of the material specification on 

each piece of the unqualified stock material. 

Contrary to the above: 

(1) The Unqual i fied Source Materi a1 Approved Vendor List , referenced in 

Procthre 17375, "Procedure for Procurement and Inspection of 

Unqual ified Source Material, I' Rev. N/C, had not been developed and 

was not available for use. 

(2) Procedures were not controlled, in that two procedures, 17375 and 

17053, with the same title, "Procedure for Procurement and 

Inspection of Unqualified Source Material ,'I but different 

requirements, were available for use. 

unqual ified source material based on subcontracting the performance 

of product analysis but without performing or subcontracting all 

other requirements of the material specification. 

(Nonconformance 99900264/96-01-01) 

(3) Procedure 17375 permitted the acceptance and furnishing of 

Enclosure 1 

41 
B. Criterion VII, "Control of Purchased Material, Equipment, and Services," 

of Appendix B to 10 CFR Part 50, requires, in part, that measures shall 

be established to assure that purchased material conforms to procurement 

documents . 

Paragraph NCA 3862.2, "Quality System Program Statement," requires 

material organizations qualified by parties other than ASME to show the 

revision and date of the applicable written quality system program on 

the certified material test report or certificate of compliance. 

Contrary to the above, Dragcn Valves, Inc., (Dragon) accepted materi a1 

certifications for 1 inch, type 316 stainless steel tubing from Sandvik 

Steel without the appropriate qual i ty system program statement and 

utilized this material without upgrading. (Nonconformance 99900264/96-01-02) 

C. Criterion VII, "Control of Purchased Materials, Equipment, and 

Services," of Appendix B to 10 CFR Part 50, requires, in part, that 

measures shall be established to assure that purchased material conforms 

to procurement documents. 

Contrary to the above, Dragon had not established a documented basis to 

substantiate that the chemical sampling plan for dedicating unqualified 

material prescribed in Procedure 16562, "Material and Parts Handling for 

Customer Defined Safety Related Components and Safety Related Items, 'I 

Rev. A, provides reasonable assurance that the dedicated material 

conforms to the procurement document requirements. 

(Nonconformance 99900264/96-01-03) 

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 

i s determined and corrective action taken to precl ude repetition. The 

identification of the significant condition adverse to quality, the 

cause of the condition, and the corrective action taken shall be 

documented and reported to appropriate levels of management. 

D. 

Section 16 of the Dragon QA manual, "Corrective Action," Fifth Issue, 

Revision 0, dated March 6, 1996, contains requirements, for significant 

conditions adverse to quality, to determine the cause, take corrective 

action to preclude recurrence, document and report these actions to the 

President and Manager of the departments involved, and to take follow-up 

action to verify implementation of the corrective action by the QA 

Manager and document these actions 

Contrary to the above, Dragon had not documented the cause, corrective 

actions, reporting to management, or follow-up action to verify 

correct've action implementation, for a Dragon Instrument Isolation 

Manifold which had been shipped to Duane Arnold with the "line" and 

"instrument" markings, stamped on the manifold body, reversed and where 

additional mismarked manifolds had been identified in the Duane Arnold 

and Dragon inventories. (Nonconformance 99900264/96-01-04) 

2 

42 
Please provide a written statement or explanation to the U.S. Nuclear 

Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555, 

with a copy to the Chief, Special Inspection Branch, Division of Inspection 

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

the date of the letter transmitting this Notice of Nonconformance. This reply 

should be clearly marked as a "Reply to a Notice of Nonconformance" and should 

include for each Nonconformance: (1) a description of steps that have been or 

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

been or will be takm to prevent recurrence; and (5) the dates your corrective 

actions and preventative measures were or will be completed. 

Dated at Rockville, Maryland 

this 29th day of October 1906 

3 

43 
Report No: 

Organization: 

Contact: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

99900264/96-03 

Dragon Valves, Inc. 

Norwalk, CA 90650 

Robert L. Snyder, Qual i ty Assurance Manager 

(31C) 921-6605 

Nuclear industry Manufactures instrumentation, manifold and specialty 

Activity: valves 

Dates: 

Inspectors: 

Approved by: 

September 23 - 26, 1996 

Richard McIntyre, Senior Reactor Engineer, Team Leader 

Uldis Potapovs, Senior Reactor Engineer 

Bill Rogers, Reactor Engineer 

Gregory C. Cwalina, Chief 

Vendor Inspection Section 

Special Inspection Branch 

Division of Inspection and Support Programs 

Enclosure 2 
1 INSPECTION SUnMARY 

During this inspection, the Nuclear Regulatory Commission (NRC) inspectors 

reviewed the implementation of selected portions of the Dragon Valves, Inc., 

(Dragon) quality assurance (QA) program, and reviewed activities associated 

with Dragon's activities in supplying instrumentation, manifold, and specialty 

valves to NRC licensees in accordance with the NRC licensee purchase order 

requirements. 

The inspection bases were: 

0 Appendix B, "Qual i ty Assurance Criteria for Nuclear Power P1 ants and 

Fuel Reprocessing Plants," to Part 50 o f T i t l e 10 of the Code of Federal 

Reaulations (10 CFR Part 50) 

10 CFR Part 21, "Reporting of Defects and Noncompl iance" 

American Society of Mechanical Engineers Boiler and Pressure Vessel 

Code, Section 111 

During this inspection, two non-cited violations of NRC requirements were 

identified. The violations are discussed in Section 3.1 of this report. 

During this inspection, four instances where Dragon failed to conform to NRC 

requirements imposed on them by NRC licensees were identified. These 

nonconformances are discussed in Sections 3.2, 3.3, and 3.4 of this report. 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

There were no open findings from previous NRC inspections of Dragon. 

3 

3.1 

INSPECTION FINDINGS AND OTHER COMMENTS 

10 CFR Part 21 Prosram 

The inspectors reviewed Dragon Procedure 14501 , "Procedure for 

Compliance to 10 CFR 21," Revision A, dated June 14, 1993, and 

representative documentation to verify implementation. Procedure 14501 

established the responsibilities and actions for the reporting of 

defects and the informing of deviations in accordance with the 

requirements of 10 CFR Part 21. The inspectors noted, and discussed with 

Dragon management, that the definitions of the terms deviation (nonconformance) and defect included in the Procedure 14501 were not in 

complete agreement with the definitions included in 10 CFR Part 21. 

Section 10 CFR 21.21 requires that entities subject to the regulations 

adopt procedures to evaluate deviations within 60 days of discovery, 

prwide interim reports to the KX within 60 days of discovery, and 

inform a director or responsible officer within five working days after 

the completion of an evaluation which determines that a defect or 

failure to comply exists. Procedure 14501 stated that, for items which 

had been shipped to a customer and subsequently identified to contain a 

2 

45 
deviation or defect, Dragon would perform an evaluation within fortyeight hours. Upon completion of the evaluation, Dragon would notify the 

customer and the NRC of the deviation or defect and the Dragon 

disposition of the items affected and impact on equipment. If the 

evaluation could not be performed within forty-eight hours, the customer 

would be notified of the deviation and would be requested to provide 

Dragon the current status of the delivered items. 

concluded that although Procedure 14501 included requirements in the 

areas discussed in 10 CFR 21.21, the procedure's requirements did not 

agree with, or include all of, the requirements specified in 10 CFR 

21.21. 

and concluded that there had not been an occurrence where Dragon's 

actions were not in accordance with the 10 CFR 21.21 requirements. 

The inspectors 

However, the inspectors reviewed Dragon's 10 CFR Part 21 records 

The inspectors reviewed Dragons's posting in accordance with the 

regulations in 10 CFR 21.6. Dragon had posted 10 CFR Part 21 and 

Procedure 14501 but had not posted Section 206. 

a copy of Section 206 which Dragon subsequently posted. 

The inspectors provided 

The failure to adequately proceduralize the requirements specified in 

10 CFR 21.21 and to meet the posting requirements specified in 10 CFR 

21.6 constitutes violations of minor significance which are being 

treated as Non-Cited violations, consistent with Section IV of the NRC 

Enforcement policy (NUREG-1600). 

3.2 Procurement and Utilization of Material for Amlications Which Are 

Reauired to Meet American Society of Mechanical Enqineers (ASME) Boiler 

and Pressure Vessel Code (Code). Section 111 requirements. 

A1 though Dragon's ASME Section I I I, Divi si on 1 Qual i ty Assurance Manual , Fifth Issue, Revision 0, dated March 6, 1996, (QA manual) contains 

provisions for upgrading unqualified source material, most of the 

material used in the manufacture of N-stamped valves supplied by Dragon 

is purchased from ASME-accredi ted material organizations or suppliers 

that have been qual i f i ed by Dragon. Since most Dragon-suppl i ed valves 

are in the 1/2 to 2-inch size range, much of the material used in their 

construction can be purchased under the ASME Code Section 111, 

Subsection NCA, Subarticle 3800 (NCA 3800) small parts exclusion or the 

provisions of ASME Code, Section 111, Subsection NX, Subarticle 2610 (NX 

2610) which exempt small product categories from most of the NCA 3800 

quality requirements. The application of NX 2610, when utilized, is 

indicated on Dragon's certificate of compliance. It was noted that the 

purchase orders of several utilities specifically prohibited the use 0; 

the NX 2610 exclusions and that most of the material in Dragon's stock 

was purchased to NCA 3800 requirements. 

3.2.1 Proclram Controls. 

The inspectors reviewed the QA manual, Section 7, "Control of Purchased 

Items and Service.," Fifth Issue, Revision 0, dated March 6, 1996, which 

describes their practice for purchasing material for applications 

3 

46 
required to meet ASME Code requirements. 

address the upgrading of unqual if i ed source materi a1 . Paragraphs 7.5.1 through 7.5.6 

Paragraph 7.5.1 states that material shall be purchased to Section I11 

requirements or upgraded by Dragon as permitted by NCA 3855,5(a)(l) 

through (a) (4), and out1 ined in Dragon's Procedure 17375. 

Paragraphs 7.5.2 and 7.5.3 require that a certification of the 

requirements which must be performed during tils melting and of the heat 

analysis be obtained from the manufacturer of the source matericl, and 

that "Dragon Valves, Inc, shall perform or subcontract a71 other 

requirements on each piece of the stock material." 

requires Dragon to perform a product analysis on each piece of source 

materi a1 . 

Paragraph 7.5.4 

In response to inspector questions concerning the above, Dragon's QA 

manager stated that "a1 1 other requirements," as referenced in Paragraph 

7.5.3 of the QA manual, pertained to the product analysis. In summary, 

no mechanical testing was procedurally required to upgrade unqualified 

source material. This position is consistent with Dragon's implementing 

Procedure 17375, Rev. N/C, "Procedure for Procurement and Inspection of 

Unqualified Source Material." For material used for valves over 3/4 

inch nominal pipe size, Section 4, "Material Verification," requires 

only chemical testing each piece of the source material. 

The inspector noted that the text of paragraph NCA 3855.5(a)(3) requires 

the upgrading organization to perform or subcontract "all other 

reauirements of the material mecification on each piece of unqualified 

source material .I' These requirements typically include tensile testing 

and may also include hardness testing, bend testing, hydrostatic 

testing, impact testing and other tests. Failure to establish source 

material upgrading requirements in accordance with the provisions of NCA 

3800 was ident i f ied as an example of Nonconformance 99900264/96-01-01. 

Dragon's qual i ty assurance manager stated that Dragon had not upgraded 

any unqualified source material exceeding the NCA 3800 small parts 

exclusion provisions since the implementation of this procedure, 

therefore the incorrect requirements contained in that procedure should 

have no effect on Code compliance of supplied items. 

sample of material procurements during this inspection tends to support 

this conclusion. 

The inspector also noted that paragraph 2.1 of Dragon's Procedure 17375, 

which is referenced in Section 7 of the QA manual for procurement of 

unqualified source material, states that material may be purchased from 

any suppliers on the Unqualified Source Material Approved Vendor List. 

However, the basis and use of this list was not discussed in the QA 

manual and the list had not been corqleted and was not available for 

review. The QA manager produced another procedure , ident i f i ed as 17053, 

Rev. A, "Procedure for Procurement and Inspection of Unqualified Stock 

Material" which had the same title and appeared similar to 

Procedure 17375, but was not referenced in the QA manual. 

A review of a 

Procedure 

4 
17053 did not reference the Unqual if i ed Source Materi a1 Approved Vendor 

List (which was referenced in Procedure 17375), but stated that material 

may be purchased from any supplier. 

on February 29, 1996, and were, apparently, in active status. 

Inadequate control of material procurement procedures and failure to 

prepare the Unqual if ied Source Materi a1 Approved Vendor List were 

identified as examples of Nonconformance 99900264\96-01-01 

Both procedures had been approved 

3.2.2 Proaram ImDlementation 

The material certification bases for approximately 15 valve orders from 

NRC licensees were reviewed to assess the effectiveness of Dragon's QA 

program implementation. 

fol 1 owing paragraphs : 

The results of this review are discussed in the 

Nebraska Public Power District Purchase Order (PO) 962080, dated 

September 03, 1996, for one 3/4 inch, 2500 psi needle valve, part 

number 5OOFO511SW-875. 

The PO requested compliance with ASME Code, Section 111, Class 2, 

1983 edition inc'luding summer 1983 addenda, but waived Data Report 

and stamping requirements. A Certificate of Compliance (COC) was to 

be provided instead. 

Review of the supporting documentation showed that the valve disc 

material (SA 564, Type 630) was purchased from Carpenter Technology 

in Reading, PA. Carpenter certification indicated that 492 bars of 

this material (heat 853543) were shipped to Dragon in 1985. 

Although Carpenter was on Dragon's approved vendor list at the time 

of this procurement, their certification did not contain a quality 

system program statement identifying the revision and date of the 

applicable written Quality System Program as required by NCA 3800. 

The file did contain a generic letter from Carpenter stating that 

material supplied by them has been manufactured, processed, and 

distributed in accordance with MIL-1-45208 and Carpenter quality 

assurance manual revision in effect at the time of manufacture. 

Tennessee Valley Authority (TVA) PO 92NNA-76504 B, dated June 29, 

1992, for several valves in accordance with TVA Standard 

Specification MEB-SS-10.19, "ASME Code Valves - 2 Inches and Smaller 

for TVA Projects." 

Review of supporting documentation for a 1/2 inch, 1500 psi globe 

valve (Fig. 10617-8) showed that the valve body forgings were 

supplied by Ajax Forge, which was on Dragon's approved vendor list. 

Ajax purchased the starting material, SA 182, Type 316 bar (41 

lengths) from ESCO Corp., a distributor, which, in turn, obtained 

the material from Altech Special'y Steel Corp. Neither ESCO or 

Altech certifications included a quality system program statement. 

From the available documentation, it could not be determined whether 

the distributor and steel manufacturer had been audited by Ajax to 

assure heat traceability of this material or whether Ajax had 

5 

48 
3.3 

upgraded this material. Review of the supporting documentation for 

two additional 1/2 inch valves from the same PO, with bodies 

provided by the same forge shop, showed that Ajax purchased the 

starting material from Fry Steel Co. of Santa Fe Springs, CA, which, 

in turn, obtained it from Atlas Specialty Steel. Neither Fry nor 

Atlas certifications included a quality system program statement. 

Review of supporting documentation for 1 inch, Type 316 stainless 

steel tubing used in the manufacture of valves covered by the same 

PO showed that Dragon had purchased this material in 1988 from a Los 

Angeles distributor, Tube Sales. lube Sales had obtained this 

material from Sandvi k Steel, Scranton, PA. A1 though Sandvi k Steel 

holds an ASME qual i ty systems certificate (QSC) , and a copy of the 

certificate was in the PO file, this material was apparently 

purchased as commercial grade tubing since Sandvik’s certification 

did not reference their QSC or contain a quality systems program 

statement. Sandvik certification identified the material as cold 

finished ASME SA 213, Type 316 steel (Heat 477898). In addition to 

reporting the heat analysis and tensile test results, the 

certification indicated that flattening and flaring tests had been 

performed on this material. The file also contained a letter from 

Tube Sales certifying that this material was annealed at 1900 

degrees Fahrenheit and water quenched. 

perform additional upgrading on this material. 

Dragon apparently did not 

The inspector noted that, lince the Sandvik certification did not 

reference their QSC or contain a quality systems program statement 

and since no upgrading was performed, this material did not fully 

comply with the requirements of NCA 3800. 

that there was no apparent basis for the heat treatment statement 

provided by Tube Sales. 

The inspector also noted 

The program implementation review indicated that the majority of 

material used by Dragon has been purcha5ed from qualified suppliers or 

certificate holders and, in most cases, sufficient documentation was 

available to demonstrate compliance with the applicable requirements. 

However, as discussed in the examples above, the inspectors identified 

several instances where Dragon had not verified supplier qualification. 

Inadequate verification of supplier qualification was identified as 

Nonconformance 99900264/96-01-02 

Procurement and Utilization of Material for Non-Code Amlications 

Reauirinq ComDliance With 10 CFR Part 50. Amendix B 

Most of the valves supplied for safety-related nuclear applications have 

been suppl ied as N-stamped components with procurement and manufacturing 

activities controlled in accordance with Dragon’s ASME QA manual. 

some cases, as discussed in parzt;paph 3.2.2, licensees have waived the 

Code stamping requirement. Under these conditions, the valves are 

supplied under the same QA program, except that the authorized nuclear 

inspector is not involved in the manufacturing process, a Code data 

report is not provided, and the valve is not Code stamped. 

In 

49 

6 
Although Dragon's QA manual does not address manufacture of safetyrelated (10 CFR Part 50, Appendix B) non-Code valves, they have 

developed a procedure (No. 16562, "Material and Parts Handling for 

Customer Defined Safety Related Components and Safety Related Items," 

Rev. A, Dated 11/03/93) which is apparently intended to control such 

activity, 

related component parts and defines the procurement, dedication, and 

processing requirements for each category. 

material for safety-related valve components can be procured in 

accordance with Dragon's QA manual or purchased to ASME or ASTM 

specifications and dedicated. 

involves cutting a sample from each size and heat of the material and 

having chemical analysis performed by a qualified vendor. Dragon's QA 

manager stated that no valves had been supplied to NRC licensees under 

this procedure. 

This procedure identifies the safety-related and non-safetyAccording to this procedure, 

The prescribed dedication process 

Although Procedure 16562 referenced Electric Power Research Institute 

(EPRI) report NP-5652, "Guideline for the Utilization of Commercial 

Grade Items in Nuclear Safety-Related Applications," the inspector noted 

that it did not follow the EPRI dedication methodology (identification 

and verification of critical characteristics). The procedure also did 

not assure material traceability or provide a basis for assurance that 

the prescribed chemical sampling (one sample from each size and heat of 

material from a non homogeneous product lot) was adequate to provide 

reasonable assurance that the material complied with the specification 

requirements. Failure to establish the basis for chemical sampling was 

identified as Nonconformance 99900264/01-03. 

3.4 Corrective Action Proqram and ImDlementation 

The inspectors reviewed the QA manual, Section 16, "Corrective Action," 

Fifth Issue, Revision 0, dated Marcn 6, 1996, and representati-;e 

documentation to verify implementation. 

responsibilities and actions for addressing conditions adverse to 

qual i ty. 

Section 16 established the 

Section 16 of the QA manual stated that conditions adverse to quality 

shall be identified promptly and corrected within a documented time 

period. For significant conditions adverse to quality, the cause shall 

be determined and corrective action taken to preclude recurrence. In 

addition, identification, cause and corrective actions shall be 

documented and reported to the President and Manager of the departments 

involved and follow-up action shall be taken to verify implementation of 

the corrective action by the QA Manager and documented. 

The inspectors discussed a recent occurrence in which a Dragon 

Instrument Isolation Manifold had been shipped to the Duane Arnold 

nuclear power plant with the "1 ine" and "instrument" markings, stamped 

on the inanifold body, reversed. The situation had been identified in 

April 1995 by Duane Arnold during the testing of a replacement 

instrument isolation assembly. Licensee documentation indicated that if 

the isolation assembly had been installed with the manifold reversed, 

7 

50 
the isolating valves would have closed with process pressure below the 

valve seat instead of above and the equalizing valve would have been 

located on the process side of the manifold instead of the instrument 

side. 

corresponding instrumentation and the equalizing valve opened, this 

could have caused a reduction in the differential pressure between the 

two process 1 ines which could have affected other instrumentation fed by 

the same lines. 

As a result, if the manifold was isolated to calibrate the 

Dragon manayellrent recall ed the event and the re1 ated discussions with 

Duane Arnold and that, in addition to the mismarked manifold in the 

assembly, Duane Arnold had located another one in inventory. 

instructed Duane Arnold to restamp the manifold and had reviewed the 

Dragon inventory and discovered additional mismarked manifolds. 

determined that the manuf;:: _. ing technician who marked the manifolds 

had used the correct drawings but had made an error when applying the 

markings and the inspector who performed the final inspection of the 

manifold had failed to find the mismarkings. 

Dragon had 

Dragon 

Further discussion with Dragon and review of the records indicated that 

Dragon had not documented the condition, had not entered the condition 

into the Drycn corrective action program, had not documented any 

corrective actions, and had not documented any actions to preclude 

recurrence. 

The inspectors concluded that shipping mismarked manifolds to a licensee 

was a significant condition adverse to quality and that Dragon had not 

met the requirements of Section 16 of the Dragon QA manual to determine 

the cause, take corrective action taken to preclude recurrence, document 

and report these actions to the President and Manager of the departments 

involved, and to take follow-up action to verify implementation of the 

corrective action by the QA Manager and document these actions. Failure 

to perform and document the required actions for a significant condition 

adverse to qual i ty was identified as Nonconformance 99900264/01-04. 

3.5 Indoctrination and Training 

The inspectors reviewed the QA manual, Section 2, “Program,” Fifth 

Issue, Revision 0, dated March 6, 1996, which described the 

indoctrination and training requirements for Dragon inspection and test 

personnel and auditors. Section 2 addressed job functions such as 

receiving, in-process, and final inspection; hydrostatic, seat leakage, 

and functional testing; welding; non-destructive examination; and 

internal and external vendor audits. 

The inspectors reviewed documented training records for various 

personnel for the job functions described above. The inspectors also 

reviewed the training files and qualification records for lead auditors, 

welders, non-destructive examination subcontractors, and the various 

inspection and test personnel. The inspeclors verified that a training 

plan and schedule was implemented for the job functions examined and 

that all training that had been planned and scheduled had been completed 

8 
for the last three years. The inspectors conc 

files contained all pertinent training records 

documents required by the Dragon QA program. 

3.6 Internal Audits 

The inmectors reviewed the OA manual. Section 

uded that the training 

and qual i f i cation 

18, "Audits," Fifth 

Issue, Revision 0, dated March 6, 1996, which described the system used 

for determining the overall effectiveness of the Dragon QA program. 

Section 18 requires that all sections of the QA manual be auditea 

annually by a lead auditor using approved checklists. The various QA 

manual section audits are performed by different lead auditors within 

Dragon so that no section of the QA manual is audited by an individual 

who has any responsibility within that area. The inspectors verified 

that the internal audits for all sections of the QA manual had been 

performed for the last three years. The inspectors concluded that the 

audits appeareti to be effective in reviewing all aspects of program 

implementation. 

3.6 Entrance and Exit Meetinas 

In the entrance meeting on September 23, 1996, the NRC inspector; 

discussed the scope of the meeting, outlined the areas to be inspected, 

and established interfaces with Dragon management. In the exit meeting 

on September 26, 1996, the inspectors discussed their findings and 

concerns. 

9 

52 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 2OSSOOO1 

November 25, 1996 

Mr. 3. R. Bohart 

Acting President 

Framatome Cogema Fuel s 

3315 Old Forest Road 

P.O. Box 10935 

Lynchburg, VA 24506-0935 

SUBJECT: NRC INSPECTION REPORT NO. 99900001/96-01 

Dear Mr. Bohart: 

On April 17, 1996, the U.S. Nuclear Regulatory Commission (NRC) completed an 

inspection of the Framatome Cogema Fuels (FCF), formerly the Babcock and 

Wilcox Fuel Company, activities at the engineering and fabrication facilities 

in Lynchburg, Virginia. This letter transmits the report of that inspection. 

During this inspection, the NRC inspectors did not identify any instances 

where activities controlled by quality assurance program 56-1177617-03, 

"Quality Assurance Program for B&W Fctel Company," May 30, 1994, failed to meet 

NRC requirements for the areas inspected. The enclosed inspection report 

contains a detailed discussion of the areas examined during the inspection and 

our conclusions. 

The inspectors did, however, observe certain FCF activities that were 

considered weaknesses that could affect quality. 

enclosed report, the inspectors considered the most significant weaknesses to 

be (a) the lack of detailed design instructions and (b) the lack of an overall 

multi-disciplinary review of the Three Mile Island Unit 1 Cycle 10 reload core 

design process by FCF to identify and evaluate the synergistic effect of 

design changes. To facilitate our closure of this matter, please provide a 

written explanation of the steps that have been or will be taken to address 

these weaknesses. 

Of those described in the 

Please send the response to the NRC, ATTN: Document Control Desk, Washington, 

D.C. 20555, and a copy, at the same address, 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. 

For the nonconformances cited in previous reports of NRC's inspections of the 

Babcock and Wilcox Fuel Company (NRC reports 99900001/90-01 and 99900001/91- 

01), the inspectors determined that t: Q corrective actims taken were adequate 

and affectively implemented. The enclosed report documents closure of those 

nonconformances. 

53 
J. Bohart -2- 

In accordance with Section 2.790(a) of Title 10 of the Code of Federal 

Reaulation (10 CFR), a copy of this letter and its enclosure will be placed in 

the NRC Public Document Room and made available to the public unless you 

notify this office by telephone within 10 days of the date of this letter and 

submit a written application to withhold the information contained therein. 

Such application must be consistent with the requirements of 10 CFR 

2.790(b)(l). 

the clearance procedures of the Office of Management and Budget, as required 

by the Paperw6tk Reduction Act of 1980, Public Law No. 96-511. 

Your response to this letter and its enclosure is not subject to 

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

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

Sincerely , 

Robert M. Gallo, Chief 

Special Inspection Branch 

Division of Inspection 

and Support Programs 

Off i ce of Nucl ear Reactor Regul at i on 

Docket No.: 99900001 

Enclosure: Inspection Report 99900001/96-01 

54 
Report No. : 

Organi rat i on : 

Contact: 

Nucl ear Industry 

Activity : 

Dates: 

Inspectors : 

U.S. NUCLEAR REGULATORY COClHISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

99900001/96-01 

Framatome Technologies, Inc. (FTI) 

Framatome Cogema Fuel s (FCF) 

Lynchburg, Virginia 

R. L. (Ronnie) Gardner 

Manager, Quality FCF 

FCF provides pressurized-water reactor (PWR) re1 oad 

core designs, safety analysis, and licensing, fuel 

assemblies, and fuel-related core components to the 

U.S. nuclear industry. 

March 18, 1996 - April 17, 1996 

Steven M. Matthews, DISP/PSIB 

David H. Brewer, DISP/PSIB 

Dr. John F. Carew, Brookhaven National Laboratory 

Carl 3. Czajkowski , Brookhaven National Laboratory 

Geoffrey R. Golub, DSSA/SRXB 

Rodney L. Grow, Parameter, Inc. 

Edward D. Kendrick, DSSA/SRXB 

Kamalakar R. Naidu, DISP/PSIB 

Billy H. Rogers, DISP/PSIB 

Approved by: Gregory C. Cwalina, Chief 

Vendor Inspection Section 

Special Inspection Branch 

Division of Inspection and Support Programs 

55 
1 INSPECTION SUWARY: 

From March 18, 1996, through April 17, 1996, the U.S. Nuclear Regulatory 

Comnission (NRC) conducted an inspection of Framatome Cogema Fuel s (FCF) 

[formerly the Babcock and- Wilcox (B&W) Fuel Company (BWFC)] activities at the 

engineering and fuel fabrication facilities in Lynchburg, Virginia. 

The inspection bases were: 

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

"Suppression of Reactor Power Oscillations," of Appendix A, "General 

Design Criteria for Nuclear Power Plants," to Part 50, "Licensing of 

Production and Utilization Facilities," of Title 10 of the Code of 

Federal Requlations (10 CFR) 

Appendix B, "Qual i ty Assurance Criteria for Nuclear Power P1 ants and 

Fuel Reprocessing Plants," to 10 CFR Part 50 

e 

a 

e 

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

Defect 'I 

Section 4.2, "Fuel System Design," of NRC NUREG-0800, "Standard Review 

Plan" (SRP), Revision 2, July 1981, and its Appendix A, "Evaluation of 

Fuel Assembly Structural Response to Externally Applied Forces," 

Revision 0 

FCF quality assurance (QA) program 56-1177617-03, "Quality Assurance 

Program for B&W Fuel Company," May 30, 1994 

During this inspection, the team noted weaknesses and observations concerning 

FCF activities that affect quality. 

observations described in this report require any specific action or written 

response by FCF. 

Neither the weaknesses nor the 

A list of acronyms used in this report is provided on page 37. 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

2.1 Nonconformance 99900001/90-01-01 (CLOSED) 

Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Section 5.0 

of the BWFC, Commercial Nuclear Fuel Plant (CNFP) procedure QC-506, 

"Receiving Inspection - Lower End Fitting Mark C," Revision 1, October 

27, 1980, CNFP failed to identify raised metal in the form of a burr 

adjacent to one grid pad bore hole on lower end fitting 15273-04. 

FCF personnel stated that the burr had been removed and that prior to 

use, a?l lower end fittings were clianed and reinspected for burrs. All 

receiving inspection personnel were instructed in the performance of 

vi sua1 inspection. 

-1- 

56 
2.2 

2.3 

2.4 

The inspectors reviewed FCF corrective act ons taken and determined that 

FCF had implemented actions to ensure that appropriate inspection 

standards were used to determine the acceptability of the lower end 

fittings . 

Nonconformance 99900001/90-01-02 (CLOSED) 

Contrary to Criterion V of Appendix I3 to 10 CFR Part 50, procedure 

QC-1414, "Retention and Storage of Quality Assurance Records," Revision 

4, September 13, 1990, did not provide requirements for retention of 

radiographs and did not identify whether the radiographs and the reader 

sheets were qual i ty records. 

The inspectors reviewed FCF corrective actions taken and determined that 

FCF had revised procedure QC-1414, Revision 5, August 7, 3991, to 

classify reviewer reader sheets and radiographs as 1 ifetime quality 

records. 

Nonconformance 99900001 /9 1-0 1-0 1 (CLOSED 1 

Contrary to Criterion V of Appendix B to 10 CFR Part 50, and Section 

6.3.6 of procedure QC-802, "Fuel Rod Inspection (In process and Final)," 

Revision 12, February 21, 1989, a 100% visual inspection was not 

performed as evidenced by felt cleaning plugs discovered in 22 fuel rods 

after visual examination and clean1 iness acceptance were performed by 

qual i ty control (QC) . 

The QC inspector responsible for performing the visual inspection was 

counseled on several occasions to ensure that the inspector understood 

that, by initialing the route card (RC), the QC inspector documented 

that the inspection had been completed to the requirements specified. 

To ensure that instructions were fully understood and followed, the 

Manager of Inspection and each Inspection Foreman emphasized the 

importance of and the need for compliance with inspection instructions. 

The inspectors reviewed FCF corrective actions taken and determined that 

FCF had implemented actions to ensure that appropriate inspection 

standards were used to accept fuel clad tubing. 

that subsequent to the implementation of these changes, FCF changed its 

method for cleaning the inside diameter of the tubing and discontinued 

the use of felt cleaning plugs. 

The inspectors noted 

Nonconformance 99900001 /91-01-02 (CLOSED1 

Contrary to Criteria V, XV, and XVI of Appendix B to 10 CFR Part 50, 

CNFP did not issue a component discrepancy report (CDR) and a contract 

variation approval report (CVAR) to identify, document, evaluate, 

resslve, and process the noncon mmance ider:tiFi -.if when felt cleaning 

plugs were found in 22 fuel rods. 

-2- 

57 
A CDR and a CVAR were initiated in August 1991. The Manager of QA 

conducted a series of meetings to identify and correct the causes of not 

reporting this nonconformance. In this case, the true cause for not 

issuing a CDR was that the responsible personnel viewed the recovery 

operation as an in-process reject that scraped each fuel rod containing 

felt cleaning plugs. During that time, the in-process reject practice 

was allowed without issuing a CDR. The current practice requires that a 

CDR be issued if a part or process deviated and the resolution of that 

deviation was not covered by a procedure or RC. 

The inspectors reviewed FCF corrective actions taken and determined that 

FCF had implemented appropriate actions to ensure that CDRs and CVARs 

would be issued as required. 

2.5 Nonconformance 99900001191-01-03 (CLOSED) 

Contrary to Criterion V of Appendix I3 to 10 CFR Part 50, procedure MA450 did not provide guidance to the operator for the extent of QC 

inspection required or what notification was to be given to QC to ensure 

that the visual examination for cleanliness was performed. 

Additionally, MA-450, Revision 16, March 8, 1991, and QC-802, Revision 

13, April 29, 1991, did not provide the extent of the inspections to be 

performed by the operator or the QC inspector. 

The inspectors’ review of the procedures referenced determined the 

following: 

0 

e 

Procedure MA-450 was at Revision 23, December 18, 1995, and the 

applicable paragraphs (4.2 and 4.3) were responsive to the 

concerns. 

Procedure QC-802 was at Revision 20, August 1, 1995, and paragraph 

6.3.5 was responsive to the concerns. 

2.6 Unresolved Item 99900001/91-01-04 (CLOSEDL 

The inspectors determined that BWFC had initiated actions to identify 

the root cause for the felt cleaning plugs ’left in loaded fuel rods. An 

evaluation in accordance with Section 21.21 of 10 CFR Part 21 may also 

be required. Pending receipt by NRC of a written response Tontaining 

the evaluations, this issue was considered an unresolved item. 

3 

The inspectors reviewed the FCF’s root cause evaluation 51-1205252-00 

and concluded that FCF had adequately evaluated the event. 

INSPECTION FINDINCS AND OTHER COMMENTS 

-3- 

58 

This inspxtion included an evaluation of the FCF fuel operations management, 

engineering, and quality staff, and certain activities for providing reactor 

1 icensees PUR neutronic design information, reload core designs, fuel 

assembl ies, fuel re1 ated core components, and fuel re1 ated field services. 
The emphasis of the inspection was on the fuel design and engineering, 

organization, management , training and qual if i cat i ons , corrective actions , and 

compliance with the requirements of 10 CFR part 21. 

3.1 

a. 

10 CFR Part 21 Program 

Inspection Scope 

The inspectors reviewed FCF compliance with the NRC requirements in 10 

CFR Part 21. The FCF procedures implementing the requirements of 10 CFR 

Part 21 were BWNT-1707-01 , "Processing Safety Concerns, I' Revision 24, 

April 13, 1994, and BWNT-1707-03, "Reporting of Defects and Failures to 

Comply (10 CFR 21)," Revision 1, 1994. The inspectors reviewed selected 

10 CFR Part 21 evaluations, known as preliminary report of safety 

concern (PSC) . 

b. 

b. 1 

Observations and Fi ndi ngs 

PSC 5-94 

PSC 5-94 concerned the fuel peak cladding temperature (PCT) during the 

reflood stage of the large break loss-of-coolant accident (LOCA). The 

inspectors discussed the PSC with the LOCA and safety analysis task and 

supervising engineers. 

values for (a) the inlet enthalpy in the core thermal analysis and (b) 

the assumed initial inventory in the core flood tank (CFT). The 

nonconservative inlet enthalpy was due to the use of a coarse time step 

in the reflood analysis, and an inadequate averaging of the nodal 

enthalpy following node dryout and rewet. 

The PSC addressed the use of nonconservative 

The nonconservative initial CFT inventory was first observed on 

September 30, 1994, during the development of the new FCF RELAP 5/MOD 2 

- B&W methodology for application to the MARK-B11 LOCA an;alysis. At 

this time the licensed LOCA analysis was based on CRAFT 2 . A 

subsequent CRAFT 2 analysi s conf i rmed the nonconservat i sm in the assumed 

initial CFT inventory. 

These nonconservatisms in the LOCA analysis resulted in a reduction in 

the LOCA kW/ft limits. FCF evaluated the effect of the reduction and 

determined that, because of available margin in the operating limits. no 

change W~S required in the FCF determined LOCA-based operating 1 irrii ts. 

During these discussions, the inspectors noted that four months had 

elapsed from the time the LOCA anaiysis initial CFT inventory 

nonconservatism was first identified (SeDtember . -I 1994) and the date NRC 

'"CRAFT2 - Fortran Program for Digital Simulation of Multinode Reactor 

Plant During Loss of Coolant," J. 3. Cudlin, B&W, October 1982 

-4- 

59 
b.2 

received notification2. 

initial identification was made using the unlicensed RELAP 5/MOD 2 - B&W 

methodology, the inspectors concluded that a final confirmation of this 

nonconservatism could have been made using existing CRAFT 2 models. 

Although the inspectors recognized that the 

Procedure BWNT 1707-01 required that the NRC be notified within 60 days 

after the PSC is initiated. PSC 5-94 was initiated on November 28, 

1994, and was reported to the NRC on January 27, 1995; within the 60 day 

requirement. However, Section VIII, of this procedure also required, in 

part, that a responsible engineer initiate a PSC for any concern 

discovered during, and which is related to, the design and analysis that 

has or may have safety implications. 

The inspectors determined that approximately 60-days had elapsed between 

the time the engineer had reasonable indication that the emergency core 

cooling accident (ECCS) LOCA analysis was nonconservative and the time 

that the PSC was initiated. 

PSC 1-95 

The inspectors discussed the identification and status of PSC 1-95 with 

the FCF responsible engineer. 

a possible return to critical during the recovery stage following a 

small break LOCA (SBLOCA). During this period the plant may spend some 

time in the boiler/condenser mode and may accumulate deborated water in 

the reactor coolant pump suction piping and downcomer. 

accumulation, an operator performs a pump bump, or under natural 

circulation, this deborated water enters the core, the core boron 

concentration may not be sufficient to ensure subcritical ity. 

concern was also the subject of PSC 8-81 which was resolved in 1985.) 

PSC 1-95 identified the concern regarding 

If, after this 

(This 

PSC 1-95 was a result of several analyses which indicated that the plant 

conditions assumed in the 1985 resolution of PSC 8-81 are no longer 

valid for B&W designed PWRs. 

letter . FCF recently issued a report on the initial Phase-I 

investigation of the issues associated with PSC 1-95. 

performed for the B&W Owners Group, and was not intended as a detailed 

evaluation but rather as a path finding activity. 

follow-on Phase-I1 activity including a set of detailed transient 

analyses and an evaluation of special effects for resolving PSC 1-95. 

The NRC ?as notified of PSC 1-95 by 

This study was 

The study proposed a 

'B&W letter to NRC, "Report of Preliminary Safety Concern Related to 

3B&W letter (06-1521) to NRC, "Repori of Potential Core Criticality 

4FCF 47-1244436-00, "Preliminary Report to the B&W Owners Group on PSC 

Large Break LOCA ECCS Analysis," January 27, 1995 

Concern," June 13, 1995 

1-95 Investigation," January 1996 

-5- 

60 
C. 

The inspectors concluded that the prel iminary Phase-1 evaluation 

adequately included the major issues associated with PSC 1-95. 

Conclusions 

Based on its review of PSC 5-94, the inspectors concluded that FCF could 

have initiated the PSC before November 28, 1995, and that FCF’s failure 

to evaluate the PCT during LOCA concern when indications of the problem 

were first identified in September 30, 1994, was considered a weakness 

in the implementation of FCF’s program. 

3.2 Qual i tv Assurance 

The inspectors reviewed FCF’s QA program described in 56-1177617-03, 

“Quality Assurance Program for B&W Fuel Company,” May 30, 1994. The 

inspectors observed that the FCF Manager of Quality had the 

responsibil ity, authority and organizational freedom to enforce and 

monitor compl i ance with the program. 

The inspectors concluded that the QA program was generally implemented 

in an acceptable manner for the areas evaluated during this inspection. 

3.3 

a. 

Reload Core Desicm. Safety Analysis and Licensina Process 

Inspection Scope 

To evaluate FCF’s reload core design, safety analysis, and licensing 

processes, the inspectors reviewed the performance, interfaces, and 

documentation of the reload process. The reload core design, analysis, 

and licensing activities consisted of determining the 1 icensee 

requirements and the preliminary fuel cycle design; performing the 

steady-state and transient neutronic, thermal-hydraul ic and fuel 

performance analysis; and updating the cycle-specific reload 1 icensing 

analysis and the process computer databank. 

b. Observations and Findings 

b. 1 Licensee Requirements 

The inspectors observed that the process started with the FCF/l icensee 

reload fuel contract and the FCF reload project manager (PM) interacting 

with the licensee to define the requirements for a new reload core. The 

licensee specified the fuel assembly type and options for the new 

reload. Preliminary operating requirements and design data were 

documented in preliminary energy requirements. 

The inspectors determined this was an iterative process between FCF and 

the 1 icensee and found that the Larrespondence dc-umented the request 

for the licensee to approve the requirements and design data. 

-6- 

61 
The PM created a contract requirement document (CRD) which outlined the 

scope of supply and special licensee requirements and used a document 

release notice (DRN) to release the CRD to FCF fuel engineering. 

The inspectors concluded that FCF's 1 icensee requirements process did 

not use formal checklists or computerized design data. The inspectors, 

however, did not observe any instances where the lack of formality led 

to a design error. 

b.2 Reload Core Design Process 

The inspectors reviewed the reload core design process and observed that 

the preliminary and scoping fuel cycle designs were based on the 

preliminary operating requirements specified by the 1 icensee. This 

initial design effort determined the preliminary reload batch size and 

enrichment; the analysis uti1 ized the two-dimensional (20) lattice 

physics code (CASMO 3) results and the three-dimensional (3D) core 

simulation calculations from the NEMO computer program. This initial 

analysis was documented in the preliminary fuel cycle design (PFCD) 

report and this information was released via the DRN process. 

According to FCF, the reload licensing analysis task engineer (RELATE) 

was a key individual in the reload core design process and was 

responsible for maintaining the technology for the task. 

were also told that the reload analysis and licensing services (RALS) 

inspectors maintained the quality of the service and ensured timely 

interaction with the licensee. 

cycle design requirements (FCDR) following receipt of the CRD. Based on 

the requirements of the FCDR, the final fuel cycle design (FFCD) was 

produced and documented in a report to the licensee. 

consisted of CASMO 3 and NEMO calculations which were documented in 

calculation packages. 

The inspectors 

The cognizant RELATE generated the fuel 

The FFCD analysis 

The inspectors reviewed prior cycle FFCDs against the new reload FFCD 

calculation packages. From its review, the inspectors determined that 

the process was controlled by administrative procedures. FCF had not 

developed detailed engineering instructions. The inspectors also noted 

that the calculation packages did not contain documentation identifying 

what the reviewer checked or verified (e.g., no check marks or 

check1 ists) . The inspectors made similar conclusions rega+ing the 

reload safety analysis process discussed in Section 3.3.b.3 of this 

report. 

b.3 Re1 oad Safety Analysis Process 

The reload safety analysis process was performed to ensure that the 

updated version of the final safety analysis report (FSAR) remained 

val'd for the reload core design. 

were evaluated to ensure that the reload core thermal performance during 

lhe accidents analyzed in the FSAR 

-7- 

62 
hypothetical transients was not degraded and that site doses were below 

the acceptance criteria of 10 CFR Part 100. 

safety analysis process included the determination of the core operating 

1 imits and reactor protection system (RPS) setpoints. 

The reload safety analysis consisted of both the non-LOCA and LOCA 

analyses. The safety analysis used the flux/flow limits determined by 

the thermal-hydraul ic analysis, the power peaking 1 imits determined by 

NEMO irr the nuclear analysis, and the RPS offset limits determined in 

the NEW maneuvering analysis. In the non-LOCA analysis, the core 

parameters that impact the FSAR transients were compared to the plant 

1 icensing basis values (LBVs) to ensure that the 1 icensing basis 

remained valid for the reload core. The specific parameters evaluated 

included the moderator and Doppler reactivity coefficients, initial 

boron concentration and inverse boron worth, maximum ejected and dropped 

In addition, the reload 

rod worths at hot-full -power (HFP) and rod group worth-at hot-zero-power (HZP) . 

The transients considered included both overheating and overcool ing 

events. The overheating events included moderator density dilution, rod 

withdrawal and ejection, loss of flow, loss of power, reactivity change 

and the startup event. The overcool ing transients considered were the 

steam line break, cold water event and the dropped or stuck rod event. 

In the LOCA analysis, the linear heat rate (LHR) effects of fuel 

prepressure, enrichment, and plenum volume, and the cycle-burnup 

dependence of the fuel rod pressure and temperature were evalyated. The 

fuel rgd parameters were calculated with computer codes TACO'' or 

GDTACO in the case of gadolinia (Gd) loaded fuel. 

that can affect either the reload LOCA analysis (e.g., ECCS assumptions) 

or the plant transient analyses were reviewed against the licensing 

basis analysis. 

LBVs or a re-analysis was performed to demonstrate their acceptability. 

Plant modifications 

These changes were either verified to be within the 

The power/imbalance/flow trip setpoints and safety 1 imits were 

calculated using the cycle-specific offset limits determined in the NEMO 

maneuvering anal ys i s . 

5BAW-10180-A, "NEMO-Nodal Expansion Method Optimized," Revision 1, March 

1993 

'BAW-10141P-A, "TACO 2: Fuel Performance Analysis, 'I Revision 1, June 

1983 

'BAW-10162P-A, "TACO 3: Fuel Pin Thermal Analysis Code," Revision 1, 

November 1989 

8BAW-10184P, "GDTACO, Urania-Godol inia Thermal Analysis Code," May 1992 

-a63 
The reload evaluation also included a NEMO reactivity analysis to 

determine the refueling boron concentration required to ensure that the 

core remains subcritical following a refueling boron dilution event. 

The NEW analysis also provided core reactivity requirements for 

determining (a) the minimum borated water storage tank (BWST) 

concentration to ensure that the core remains 1% shutdown following a 

LOCA and (b) the minimum volume of borated water in the BWST and the 

boric acid storage tank to independently ensure that the core can be 

held at cold shutdown. 

The plant exclusion area and low-population zone boundary doses were 

evaluated for each reload to insure they were within the 10 CFR Part 100 

acceptance criteria. The reload fuel and coolant radionuclide 

inventories were calculated and compared to the LBVs. A reanalysis was 

performed to demonstrate the acceptability of the boundary doses if the 

reload values were outside the LBVs. 

The inspectors found that the reload safety analyses were documented in 

formal cycle-specific calculational files. The safety analysis 

calculations were reported in calculational summary sheets. 

From its review of the reload safety analysis process, the inspectors 

determined that if a core parameter value was not bounded by the LBV for 

a given transient, then the value was justified by reanalysis of the 

transient. 

calculation or a complete reanalysis of the transient. 

This reanalysis may consist of s simple sensitivity 

The inspectors reviewed prior cycle calculational files against the new 

reload calculation packages. From its review, the inspectors determined 

that the documentation and analysis technique tended to be guided by the 

prior cycle(s) documentation. The inspectors observed that the process 

was controlled by administrative procedures and that FCF had not 

developed detailed engineering instructions. The inspectors a1 so noted 

that the calculation packages did not contain documentation identifying 

what the reviewer checked or verified (e.g., no check marks or 

checklists were used). 

Based on these observations and similar observations made by the 

inspectors during their review of the reload core design process (see 

Section 3.3.b.2 of this report), the inspectors examined numerous 

calculation packages for the reload design process and the reload safety 

analysis process (see Section 3.3.b.4. (1)-(4) of this report). 

Re1 oad Licensing Process 

The reload licensing process defined the specific tasks to be performed, 

1 icensing schedule, task coordination and interfaces, and required 

design reviews. 

for the reload. 

which were identified numerically (e.g., task 4 was the ECCS analysis 

and task 11 was the nuclear analysis). A task engineer was assigned to 

each task and was responsible for the basic technology and methods 

b .4 

The reload PM provijed the primary licensee interface 

The reload licensing process consisted of 18 tasks 

-9- 

64 
employed in carrying out the task analysis. 

documented in a series of FCF reports. 

document were the primary output of the reload 1 icensing process. 

Each of the following reload designs was reviewed by the inspectors with 

the PM, the RELATE, and the engineers who performed the reload core 

design and the reload safety analyses. 

The task analyses were 

The FCOR and the site support 

(1) Three Mile Island 1 

The FCF 1 icensing analysis provided the technical justification 

supporting Cycle 10 operation of the GPU Nuclear Corporation 

(GPUN), Three Mile Island Unit 1 (TMI-1). The general reload 

licensing information consisted of the fuel assemblies, burnable 

poison rod assemblies (BPRAs), and reactor rod cluster control 

assemblies (RCCAs) for TMI-1 Cycle 10 and was specified in the 

GPUN/FCF reload fuel contract. The latest safety analysis and 

plant systems input for the Cycle 10 reload evaluation were 

provided in the plant FSAR. 

the initial number of fuel assemblies and plant operating 

conditions (e.g., flows and temperatures). These requirements 

were finalized based on the FCF fuels engineering preliminary fuel 

cycle analysis of the cycl e-1 ength, number of fuel assembl i es , axial blanket, and Gd zoning. 

The reload data checklist specified 

The reload licensing analysis was initiated by the release of the 

reload licensing schedule, which specified the tasks to be 

performed, schedule, and del iverables, The TMI-1 Cycle 10 

1 icensing analysis consisted of the standard FCF task-labeled 

analyses. 

As part of the evaluation of the FCF reload licensing process, the 

inspectors reviewed the nuclear analysis with the supervising 

engineer, the RELATE, and the enginger responsible for the 

analysis. 

complete and we1 1 documented. 

The inspectors concl uded that the cal cul ati ons were 

The inspectors found that FCF's QA program for the use of computer 

codes in the reload licensing process required that approved code 

versions be used in licen>ing analyses or, when approved versions 

were not used, a justification must be provided in the calculation 

files. During its review of nuclear analysis 32-1218970-01 and 

32-1219600-01 , the inspectors noted that punapproved versions of 

NEMO (NEMO 3.1RS and NEMO 4.1RS) and SHUF (SHUF 3.lRS) were used 

in the nuclear analysis. However, no justification for the use of 

these unapproved versions was documented as required by the FCF QA 

program. 

I 

I 

'The SHUF program is used to shuffle the fuel isotopic concentrations between 

core 7 ocat i ons . 

- 10 - 

65 
When the inspectors brought this omission to the attention of the 

FCF staff, revisions to these calculation files were provided that 

contained a justification for the use of unapproved computer codes 

on the basis that the errors in the unapproved versions had no 

impact on the results of the nuclear analysis. 

I 

The TMI-1 Cycle 10 reload safety analysis was reviewed by the 

inspectors with the responsible task engineer and unit manager of 

Framatome Technologies, Inc. (FTI). The safety analysis compared 

the Doppler and moderator temperature coefficients (MTC), initial 

boron concentration, maximum ejected and dropped rod worths, and 

rod group worths to the bounding values used in the FSAR. 

values used in the FSAR were shown to bound the values for TMI-1 

Cycle 10. The safety analysis demonstrated that the generic set 

of LOCA analyses bounded the TMI-1 Cycle 10 reload. 

The 

The TMI-1 Cycle 10 reload fuel rod analysis was reviewed by the 

inspectors with the responsible fuel rod analysis engineer. The 

fuel rod mechanical design analysis used the power distribution 

and power history data determined in the fuel cycle design 

analysis to evaluate the cladding creep collapse and the stress 

and strain for the Cycle 10 fuel. The design data for each of the 

Cycle 10 batches was taken from the applicable documents list 

(ADL) . The axial locations of the fuel stacks, control rod poison 

stacks, burnable poison stacks and the axial blankets were 

determined from the core and vessel specifications and drawings. 

The stress parameters for each of the four Cycle 10 fuel rod 

designs (MARK-B8, MARK-B8V, MARK49 and MARK-B9-Gd) were bounded 

by conservative 1 icensing basis analyses. 

for both the MARK-B8 and MARK-B9 fuel assemblies was bounded by a 

previous generic analysis. 

As part of the inspectors’ review of these generic analyses, the 

fuel rod growth calculation for the MARK-B9 fuel (32-1202547-00) 

was reviewed in detail. The growth analysis included a 

statistical combination of growth tolerances and used a 

conservative fluence growth model. 

and included the necessary verification reviews. 

The creep collapse time 

The documentation was complete 

The 28 MARK-B9-Gd fuel assemblies iwluded Gd fuel rn’s. The fue 

mechanical analysis (32-1222994-00) concluded that the creep 

collapse analysis for the Gd fuel was bounded by the generic 

analysis performed for the MARK-B9-Gd fuel (32-1219532-00). The 

generic analysis was reviewed and it was no+.ed that the analysis 

employed an unapproved version of GDTACO. 

However, upon the inspectors’ request, the FCF staff provided a 

subsequent creep col 1 apse ana 1;tsi s (32-1224702-00) which 

demonstrated that there was no significant difference in the 

results of the calculations performed by the approved versus 

unapproved versions of GDTACO. 

- 11 - 

66 
The inspectors concluded that the documentation of all analyses 

reviewed complied with the FCF documentation and verification 

requi remen t s . 

(2) Sequoyah 1 

The inspectors selected the Tennessee Val ley Authority (TVA) , Sequoyah Unit 1 Cycle 9 reload for review because Sequoyah was a 

non-B&W designed plant (Westinghouse). Cycle 9 was the first 

reload provided by FCF for Sequoyah. The inspectors were also 

concerned about the handling of design inputs for a non-B&W plant 

and the analysis to support a mixed core reload. 

were also told that the fuel design for Sequoyah, the Mark-BW, had 

been used in fuel for Catawba and McGuire. 

supplied fuel and LOCA analysis for McGuire, a sister plant to 

Sequoyah. ) 

The inspectors found that this reload was not a typical FCF reload 

in that a first-of-a-kind (FOAK) analysis was performed prior to 

initiating the typical cycle specific analysis. 

found that the FOAK effort addressed changes to the FSAR, 

technical specifications, core operating 1 imits report (COLR) and 

also produced the topical report BAN 10220P, "Mark - BW Fuel 

Assembly Application for Sequoyah Nuclear Units 1 and 2," Revision 

0, March 1996. The inspectors were told that the FOAK analysis 

addressed the Sequoyah plant specific design and focused on the 

pl ant safety analysis. The inspectors observed that the topical 

report documented the LOCA and nan-LOCA transient analysis, 

thermal-hydraul ic analysis, and mechanical and containment 

analysis. 

The inspectors 

(FCF had previously 

The inspectors 

The neutronics analysis was largely addressed in prior Mark-BW 

documents. 

FCF fuel in the Sequoyah reactor and identified the bounding 

physics parameters for the cycle-specific reload analysis. 

inspectors found that the Sequoyah design inputs were documented 

via correspondence between FCF and the licensee and the 

documentation showed that the inputs had been verified by the 

1 i censee. 

The topical report documented the licensing basis for 

The 

The inspectors found that the PFCD had been completed and FCF had 

received the final energy requirements from the 1 icensee (TVA). 

The inspectors reviewed the 1 i st of Sequoyah reload del iverables 

and contract documents list. The inspectors concluded that the 

handling of design data inputs and the documentation of FCF 

activities through the preliminary design were consistent with the 

FCF administrative manual procedures. 

To assess the analysis process the inspectors examined calculation 

package 32-1228786-00, "Sequoyah Catchup and Accident Analysis. It 

The inspectors found that this package documented the benchmarking 

of the CASMO 3/NEMO neutronics model for Cycles 1 to 8. 

- 12 - 

67 
Comparisons to INCORE (computer code used by TVA for technical 

specifications survei 11 ance) power distributions, measured rod 

worths, isothermal temperature coefficients, and boron letdown 

values were documented. 

the main steam line break analysis with comparisons to the FSAR 

results. The inspectors observed that this calculation package 

was prepared consistent with the FCF administrative manual 

procedure BWNT-0402-01, "Preparing and Processing BUNT 

Cal cul at ions , '' Revi sion 29. 

The cal cul at ion package a1 so documented 

The inspectors concluded that the Sequoyah reload package was 

adequately documented, the mixed core issues addressed, and the 

design inputs were adequately controlled and verified by the 

1 i censee. 

(3) Crystal River 3 

The inspectors reviewed the reload process for Florida Power 

Corporation, Crystal River Unit 3 (CR3), Cycle 11 with the reload 

PM and cognizant RELATE. 

FCDR, FFCD, and reload report, the inspectors observed the 

following unique characteristics about the CR3 Cycle 11 fuel 

design: 

(a) 

(b) 

(c) 

From its review of the CR3 Cycle 11 

the longest cycle length to date (670 effective full power 

days (EFPD)) 

the highest enrichment to date (4.96 w/o UZ3*) 

first time use of Gd 

The inspectors found that the FFCO was documented in calculation 

package 32-1239620-00 and this package was reviewed by the 

inspectors relative to the Cycle 10 FFCD in calculation package 

32-1224533-00. 

analysis packages were prepared by the same individual and the 

same reviewer. 

The inspectors noted that the Cycle 10 and 11 

Documentation for the MICBURN 3, CASMO 3, and NEMO calculations 

was reviewed by the inspectors. Based on its review of several 

specific items including the use of Gd for the first time, the 

change in axial nodalization in NEMO, for the new cycle, the choice 

of the K-effective bias or critical K-effective for NEMO and the 

cross sect ion generation process, the inspectors concl uded that 

the design process was effective for CR3. 

The CR3 Cycle 11 reload process was examined by the inspectors 

with the major focus on the engineering analysis which supported 

reload report BAW-2262, "Crystal River Unit 3 Cycle 11 Reload 

Report," January 1996. 

the functional areas (fuel system, nuclear design, thermalhydraulic, transient and technical specification/COLR). 

The rcload report was examined for each of 

- 13 - 

68 
C. 

The use of GDTACO for fuel performance, the appl icat ion of the 

statistical core design methods for thermal-hydraul ics, and the 

main steam line break analysis were among the topics reviewed by 

the inspectors. 

The inspectors observed that the quality of the calculation 

packages varied significantly in terms of readability and in 

overall content and in documentation of the verification process. 

The inspectors concluded, however, that the reload analysis 

process for CR3 was adequate. 

(4) Arkansas Nuclear 1 

The Entergy Operations, Incorporated, Arkansas Nuclear Unit 1 

(ANO-l), Cycle 13 reload core was an FCF full-scope design. The 

design included 57 MARK-68 and 60 MARK-BSZL reload bundles, and 60 

MARK-B9 fresh fie1 assemblies. Unlike TMI-1 Cycle 10, the fresh 

assemblies were loaded in a symmetric checker board pattern 

throughout the core that reduced the potential for crud deposition 

on the fuel (see Section 3.4.b.2 of this report). 

The inspectors reviewed the ANO-1 Cycle 13 nuclear analysis with 

the PM and RELATE. 

require any modifications to the reload methodology and the 

standard FCF reload analysis codes and methods were employed. 

inspectors reviewed the Cesign files documenting the results of 

the Cycle 13 calculations for the on-line computer input (86- 

1232957-00) , the physics data supporting the Cycle 13 siteoperation (61-1232956-00) , and the generation of key safety 

parameters for the evaluation of transients, reactivity effects 

and control rod worths (32-1232950-00). 

The specific Cycle 13 fuel loading did not 

The 

While the inspectors review of the calculational files showed 

that , in several instances, unapproved computer codes were 

employed, the inspectors also noted that each of these 

applications were justified by an evaluation of the effects on the 

results of the nuclear analysis. 

the calculations were we1 1-documented and the necessary 

verification was indicated. 

The inspectors determined that 

Concl us i ons 

The inspectors discussed with each individual the technique used to 

perform certain design functions and learned that no design instructions 

existed for either performing the analysis or for review and 

verification of the analysis. The inspectors found that the responsible 

individual follows the analysis process documented for the prior cycle. 

The inspectors also noted that t-e PM, RELATE, avj the RALS inspectors 

share the responsibility for the reload analysis. The inspectors 

determined that FCF performed these analysis processes based on the 

methodology documented in the prior cycle calculation packages in 

conformance with the guidance in the administrative manual procedures. 
3.4 

a. 

b. 

b. 1 

The lack of detailed design instructions to perform the reload 

calculations could result in a nonconservative value that may not be 

apparent from following the previous cycle methodology as a template for 

the current reload calculations. The inspectors, therefore, concluded 

that the lack of detailed design instructions was a weakness in FCF’s 

reload design process. 

However, the inspectors did not identify any instance where the lack of 

design instructions caused or contributed to an. error. Specifically, 

the inspectors noted that for the CR3 Cycle 11 reload, the first time 

use of Gd was handled appropriately. 

Fuel Assembly Mechanical Desian 

Inspection Scope 

The inspectors reviewed recent FCF experience and activities related to 

fuel assembly bow and the distinctive crud patter (DCP) following TMI-1 

Cycle 10 and CR3 Cycle 10 operation. 

Observations and Findings 

Fuel Assembly Bow 

According to FCF, all fuel assemblies experience some assembly bow. 

Fuel assembly bow affects local margin to departure from nucleate 

boiling (DNBR) as a result of reduced cooling as the assembly-toassembly gap was decreased, and reduced the margin to LOCA F limits as 

increased . a result of increased power peaking as the assembly-to-assem B ly gap was 

Because of the stronger fast flux gradient on the core periphery, the 

differential growth-induced bow was larger for fuel assemblies near the 

core-edge and, therefore, the core reload pattern typically employs a 

cross-core shuffle. FCF post-irradiation examination data taken over 

three operating cycles indicates that most of the assembly bow occurs 

during the first cycle and that the increase in later cycles was not 

1 arge. 

Because of recent control rod insertion kroblems experience? by NRC PWR 

licensees, the inspectors reviewed an assembly bow event at the Ringhals 

3 and a subsequent related event at Ringhals 4. 

included second generat ion Framatome fuel assembl ies. In the Ringhal s 4 

incident an RCCA jammed in the dashpot region and did not fully insert 

following a reactor trip. 

of the dashpot) for all other RCCAs was within the value used in the 

safety analysis. Only fresh or once-burned fuel assemblies were located 

in +he RCCA locations. 

Both of these plants 

However, the drop times (measured to the top 

As part of the investigation of these incidents, UT and Eddy current 

tests were performed on eight fuel assemblies to evaluate swelling and 

wear. Fifty-eight fuel assemblies were also inspected for bowing, 

- 15 - 

70 
b.2 

elongation and the extent of rod-to-nozzle gap closure. These tests 

indicated larger than usual (both "S" and "banana" shape) assembly 

bowing. Based on subsequent measurements and cal cul at ions, it was 

concluded that the assembly deformation was the cause of the failure of 

the RCCAs to insert. A review of the second generation Framatome fuel 

indicated that, while the fuel rod, guide tube and hold-down system 

designs were identical, the decreased grid height resulted in reduced 

lateral stiffness which increases the susceptibility to assembly bow. 

FCF analysis of the assembly bow included the effects of 

(a) differential growth due to flux gradients and (b) creep 

amp1 ification of existing deformations. The resulting fuel assembly 

design modifications resulted in increased guide tube stiffness and a 

reduced fuel assembly hold-down force. 

The inspectors concluded that FCF was actively evaluating the Ringhals 

bowing events and their relevance to FCF fuel designs. 

Fuel Cladding Distinctive Crud Pattern 

Examination of the fuel pins following TMI-1 Cycle 10 and CR3 Cycle 10 

operation indicated a distinctive crud pattern (DCP), on more than 182 

fuel rods at TMI-1 and on four fuel rods at CR3, which was significantly 

different than the "normal" corrosion pattern. 

burned) fuel rods at TMI-1 with thP DCP had through-wall leaks and many 

of the other rods with DCP had significant wall-thinning. 

Nine fresh (one-cycle 

FCF provided a detailed description off their investigation into the 

cause of the DCP observed at TMI-1 Cycle 10 and CR3 Cycle 10. The 

inspectors reviewed this material in order to identify (a) the factors 

that could have potentially contributed to the DCP and (b) the actions 

that FCF may have taken to prevent the occurrence of the DCP. 

For TMI-1 Cycle 10, the DCP tended to occur in the sixth-axial grid-span 

(second from the top of the fuel assembly) and on the outer-row of fuel 

pins of the assembly. Most of the fuel pins with DCP were located in 

one of eight symmetrically loaded "T" loading patterns. These 'IT" 

loading patterns consisted of four fresh high enrichment (4.0 to 4.75 

w/o U,,) fuel assemblies, arranged with three assemblies across the top 

of the "T" and one assembly centered below. 

located on the core periphery in the low-flow core region with one 

assembly facing the reflector (see Figure 1 on page 35 of this report). 

Following the TMI-1 Cycle 10 fuel failure presentations by FCF, the 

inspectors decided to evaluate the DCP event based on the following 

possibilities: 

These "T" patterns were 

(a) 

(b) 

the presence of crud in the coolant was caused by inadequate water 

chemistry 

the deposition of crud on the fuel pins was accelerated due to 

higher clad surface temperatures in relatively low-flow regions 

- 16 - 

71 
(c) 

The inspectors pursued investigations into various aspects of the 

failures and possible mechanisms. 

(1) 

the corrosion failures (crud/oxide spa11 ing) were caused by local 

boil ing in or under the crud 1 ayer 

Reactor Cool ant System Chemistry 

During operation of a 177 fuel assembly B&W plant, a minimum 

coolant pH level is maintained in order to prevent any significant 

increase in the deposition of crud on the fuel rods. However, the 

recent aggressive 24-month high-reactivity core designs, 1 ike TMI1 Cycle 10 and CR3 Cycle 10 have increased boron levels (to -1500 

parts-per-million (ppm) at startup) which has resulted in a 

reduction in the coolant pH to below the target value of -6.9. 

This pH reduction is generally compensated for by maintaining an 

increased lithium (Li) concentration in the coolant which 

increases the pH. 

In its letter to GPUN, "1991 Revision of BAW-1385 - Water 

Chemistry Manual 177 FA P1 ants, Li thium/Boron Control ,I1 November 

11, 1991, BWFC provided GPUN the recommended Li/boron control for 

TMI-1 Cycle 10. This recommendation was to maintain a pH of 6.9 

initially and then, if necessary, to increase the pH to a higher 

value later in the cycle. Because of the high initial boron 

concentration in the coolant for this high reactivity core, this 

would have required a Li concentration higher than the usual BWFC 

recommended limit of Li (2.2 ppm. 

designed to protect against fuel cladding corrosion.) 

(The upper limit on Li was 

In a subsequent letter to GPUN, "Lithium/Boron Control 

Requirements," December 11, 1991, BWFC recommended a second and 

different Li/boron control. This recommendation was to startup at 

a Li concentration of 2.2 ppm and hold at this value until the pH 

increases to the desired value above 6.9 (as a result of the 

decreasing boron), and then to decrease the Li in order to 

maintain the desired pH. Responding to a request from GPUN for 

BWFC to clarify its position, BWFC indicated that the second 

recommendation should be followed. 

The inspectors concluded that the final BWFC recommendation, to 

limit the maximum Li to 2.2 ppm, resulted in a pH less than 6.9 

and significantly increased the tendency for deposition of crud on 

the fuel. The low pH was of special concern at TMI-1 Cycle 10 

because of the high level of "circulating crud" believed to be 

present due to (a) the five years of plant shutdown that had 

occurred earlier and (b) the fact that the reactor coolant system 

(RCS) makeup flow was less thitn 50% of nominal. 

situation existed at CR3 Cycle 10, although the lowest pH was 6.6 

and the level of circulating crud was believed to be substantially 1 ess. 

A similar 

- 17 - 

72 
In its letter to GPUN, "TMI Cycle 11 Lithium Levels - Revision to 

Previous Reconmendations," August 28, 1995, BWFC relaxed the Li 

limit to 3.0 ppm for TMI-1 Cycle 13 in order to maintain a pH 

above 6.9 and thus reduce the potential for fuel rod crud 

depos i ti on. 

During the initial months of operation of TMI-1 Cycle 10, GPUN did 

not add Li to the reactor coolant but rather allowed lithium-7 to 

"burn-in" as a result of the 6-10 (n, a ) Li reaction. Because of 

the high Cycle 10 boron concentration, this resulted in an initial 

pH of 5.7, which increased to 6.9 after several months of 

operation as a result of the Li buildup and the decrease in boron. 

Based on the above evaluations, the inspectors made the following 

concl usions : 

The FCF communication to GPUN concerning the TMI-1 Cycle 10 

Li/boron control was not clear and may have. contributed to 

the decision by GPUN to ignore the recommendation and 

initiate Cycle 10 without Li. 

this a weakness in the FCF/customer interface. 

The inspectors considered 

(b) The FCF Li/boron control guidance provided to GPUN did not 

ensure a pH >6.9 and provide adequate protection from crud 

deposition. 

weakness in the FT: water chemistry program to establish 

adequate Li/boron control procedures. 

The inspectors considered this a result of a 

(2) Nuclear and Thermal-Hydraul ic Design 

The nuclear and thermal-hydraulic reload core design is an 

important consideration in preventing fuel rod crud, since both 

low-flow velocity and high fuel rod power enhance crud deposition; 

in part, due to the reduction of solubility with increased 

temperature. In addition, since several of the mechanisms 

suspected of causing the DCP failures involved boiling under the 

accumulated crud, the relatively small margin of 10-20°F to 

coolant boiling (in the sixth-axial span where DCP occurred) makes 

this design relatively sensitive to small increases (decreases) in 

fuel rod power (flow veic;:ity). This was especially true in TMI-1 

Cycle 10, where a BOC power-tilt of +15% occurred at the sixthaxial span in the quadrant of the core where the DCP occurred. 

(This tilt is believed to be due to boron "hideout" in the crud.) 

In the fuel assembly designs for both the TMI-1 and CR3 core 

reloads, the highest powered fuel rods were located on the outside 

row of the fuel assembly, where the lowest assembly flow velocity 

occurs. 

assemblies in the "Tl' loading pattern on the core periphery 

resulted in the unusual situation in which the highest powered 

rods in the core (they were in the highest eight percent) were 

located in the peripheral (lowest velocity) locations. In 

The placement of :he fresh high t?-,-ichment fuel 

73 

- 18 - 
addition, the relatively low burnable poison loading of both the 

TMI-1 Cyc e 10 and CR3 Cycle 10 cores resulted in higher critical 

boron concentrations and consequently, 1 ower reactor cool ant pH. 

It was well known that fuel rod crud deposition is enhanced by 

high fuel rod power, reduced flow velocity, and low coolant pH. 

However, the inspectors determined that several features of the 

TMI-1 Cycle 10 core design resulted in a significant reduction in 

the margin in these parameters to fuel rod crud deposition. These 

include (a) the highest-to-date critical boron concentration which 

lowered the coolant pH and (b) the location of the "T" loading 

pattern of high powered fuel assemblies (with the highest-to-date 

enrichment of 4.7 w/o U235) in a low-flow velocity region on the 

core periphery. 

The inspectors concluded that this design weakness together with 

the (allowed) operation at low Li, which further lowered the 

coolant pH, resulted in DCP on the high powered rods in the "T" 

pattern loaded fuel assemblies. Based on the above, the 

inspectors concluded that there was a basic weakness in the TMI-1 

Cycle 10 core design. 

materials analysis of the TMI-1 Cycle 10 core were considered weak 

by the inspectors because the reload core could not accommodate 

the expected lower coolant pH values and protect from deposition 

of crud on the fuel rods. 

The nuclear, thermal -hydraul i cs and 

Fuel Assembly Bow 

Fuel assembly bow, due to guide-thimble/fuel-rod differential 

growth, occurs preferentially on the core periphery where the flux 

gradients are strongest. As the fuel assembly-to-assembly gap 

increases, the power in the fuel rods located on the outer row of 

the assembly increase (-3% per 50 mil of gap closure). 

In the TMI-1 Cycle 10 core design, the "T" pattern of fresh high 

enrichment high-powered fuel assemblies were placed near the core 

periphery which maximized the flux gradient across these 

assemblies and the resulting assembly bow. The inspectors 

determined that this resulting increase in rod power would 

increase the potential for crud de9osition in exactll' the location 

where the DCP occurred; the outer row of fuel rods. for TMI-1 

Cycle 10 operation, both the location of the DCP failures in the 

outer row of the fuel assemblies and on the core periphery was 

consistent with the assembly bow mechanism which would have 

resulted in increased fuel rod power and local boil ing . 

The inspectors concluded that FCF failure to perform an evaluation 

(including measurements where practical) of assembly bow as a 

potential mechanism for the deposition of the DCP was a weakness 

in its DCP true cause analysis. 

- 19 - 

74 
(4) Neutronic 

One theory pursued by the inspect rs was that the high clad 

surface temperature and local boiling was caused by the inability 

to correctly predict the heat flux or pin power distributions in 

the fuel assemblies with DCP. The inspectors evaluated the 

likelihood that the accuracy of the neutronics as applied to the 

TMI-1 Cycle 10 core design was A contributor or the potential 

cause of the fuel failures. 

The inspectors were told by FCF that the majority of the pin 

failures had occurred in fuel assemblies located in the fresh fuel 

"Ts" (a four fresh fuel assembly group, which has 3 assemblies 

face-adjacent in a row forming the top of a T, see Figure 1 on 

page 35 this report). Within these "T" assemblies, the majority 

of the pin failures occurred in fuel pins on the outer row (along 

the water gap) of the assembly and in the corner pin where the 

fresh assembly faces meet. The failures occurred in the top one 

third of the fuel rod (typically 100- to 130-inches from bottom of 

the fuel ) . The inspectors evaluated FCF's abi 1 i ty to accurately 

predict the pin power in these locations. 

The inspectors reviewed the TMI-1 Cycle 10 "measured power 

distributions" (seven axial values for each of the 57 assemblies 

containing fixed incore deteq-tors) versus those predicted by NEMO. 

Particular attention was given to the comparisons in the sixth and 

seventh axial location which corresponds to the elevation with the 

crud deposition and fuel failures. Comparisons from four points 

in time throughout the cycle were examined. 

The inspectors observed that for the assemblies of interest, NEMO 

consistently over-predicted the power in the seventh level (on the 

average, 5-10% over-predicted) and more closely predicted the 

power in the sixth level (on the average 1% under-predicted to 3% 

over-predicted). This would result in the temperatures being 

cooler in the actual core in these locations as compared to the 

predicted core. 

In addition to the "measured" vs NEMO predicted power 

distributions presented in the NEMO topical report, the inspectors 

reviewed TMI-1 Cycle 9 radial power distributions. The inspectors 

observed that the radial power was typically predicted within 2% 

(root-mean-squared (RMS) or standard deviations were less than 2% 

for various databases). The inspectors also observed that the 

power in the face adjacent fresh fuel was predicted as well as 

other radial locations. 

The inspectors reviewed analyiical power distribution benchmarks 

such as comparisons between NEMO and other nodal codes or between 

NEMO and PDQ. Data from TMI-1 Cycles 8, 9, and 10 was examined 

and the inspectors observed that, compared to the other code 

prediction, NEMO tended to predict the same or higher power in the 

- 20 - 

75 
face adjacent fresh fuel locations. Based on the above 

evaluations, the inspectors concluded that for the specific "T" 

assembly locations, there was no evidence to suggest that NEMO 

under-predicted the assembly power. 

The inspectors also observed that there was nothing unusual about 

the TMI-1 Cycle 10 assembly power distributions relative to the 

benchmarked cycles (in terms of control rod presence or unusual 

burnable poison configurations). The inspectors did note that the 

fresh assembly "TI' was placed in a core location which had a steep 

power gradient (due to presence near the edge of the core). Based 

on the inspectors conclusion that NEW was predicting assembly 

power reasonably well, the inspectors next focused on the NEMO pin 

power reconstruction methodology and NEMO's ability to predict the 

local pin power. 

The inspectors reviewed single assembly CASMO 3 calculations and 

NEMO quarter core with pin power reconstruction calculations to 

determine what the pin power distributions were for current FCF 

designs. The inspectors observed that the newer designs, e.g., 

CR3 Cycle 11, have a flatter assembly pin power distribution and 

the peak/average pin power ratio was typically less than 1.10. 

The older designs may have peak/average pin power ratios of up to 

1.2. The inspectors observed that the NEMO calculations for the 

pin power ratios in the fresh "T" locations show a very large 

gradient across the assembly (e.g., 1.4 to 0.6 or 1.5 to 0.8). 

The inspectors also noted that the peak pin power often occurs in 

the corner pin where the fresh assembly faces meet. 

The inspectors reviewed mu1 ti assembly (2x2 array) comparisons of 

NEMO pin power distributions vs MCNP and comparisons of NEMO vs 

CASMO 3. 

than 2% and often less than 1%. The inspectors concluded that 

NEMO was reconstructing the pin power distribution satisfactorily 

for these analytical comparison; however, the actual core pin 

power distributions were more challenging (steep gradients). The 

inspectors also noted that their concern about the NEMO accuracy 

partially arises from the lack of directly applicable benchmarks. 

That is, FCF had no benchmarks which duplicate the TMI-1 Cycle 10 

core design and the pin power distributions gradients present in 

these core designs. 

The inspectors observed that the differences were less 

Based on the above review of the neutronic aspects of the TMI-1 

Cycle 10, the inspectors concluded that the NEMO methodology can 

accurately predict the assembly and pin power distributions. 

there was no evidence to support the hypothesis that the 

inaccuracy of the neutronic models contributed to the fuel 

corrosion failures. 

Thus 

21 

76 
However, the inspectors also concluded that the TMI-1 Cycle 10 

design challenged the methodology more than other FCF designs via 

the "T" placement of fresh assemblies. The TMI-1 Cycle 10 design 

resulted in large pin power peaks in the corner pins and steep 

power gradients across the assembly and these local conditions can 

not be found in any of the NEMO benchmarks; thus, the methodology 

was stressed to near its maximum capability. Other FCF core 

designs have either not used the "T" loading, or used an "L" 

loading which had lower pin peaking. 

that these factors resulted in a weakness in the TMI-1 Cycle 10 

reload core design. 

The inspectors concluded 

c . Concl us i ons 

Based on the review of the DCP and associated fuel rod failures, the 

inspectors concluded that the FCF design process was also a contributing 

factor to the DCP. 

of the TMI-1 Cycle 10 core design which by themselves were not 

significant contributors to DCP, but when taken together, resulted in a 

reduction in the margin to crud deposition and localized boiling. 

factors included: 

The inspectors noted that there were many features 

These 

(a) increased critical boron and the resulting decrease in reactor 

coolant pH 

(b) inadequate Li/boron control 

(c) inadequate communication io TMI-X regarding the establishment of 

the Li/boron control 

(d) low coolant pH was not accommodated by the nuclear, thermalhydraulic, and material design 

(e) challenged the NEMO methodology by using "T" loading pattern 

(f) both the fuel core and lattice designs placed the highest powered rods in the locations of lower-flow velocity and higher assembly 

bow 

The inspectors considered the lack of an overall mu1 ti-discipl inary 

review process by FCF to identify and evaluate the synergistic effect of 

the changes that occurred in the TMI-1 Cycle 10 core design to be a 

weakness in the FCF design process. 

The inspectors also considered the lack of an evaluation of assembly bow 

as a potential contributing mechanism as a weakness in FCF's true cause 

analysis of the DCP. 

3.5 Enaineerina ComDuter Prowarns and Database 

a. Inspection Scope 

fCF's use of engineering compute programs (FCPsj was contingent upon a 

tiered approach to computer program certification. Programs (codes) can 

be fully certified, conditionally certified, interim certified, and 

uncertified. Procedure BWNT 0402-01, "Preparing and Processing BWNT 

- 22 - 

77 
Calculations," Revision 29, June 1, 1994, proscribed the use of 

certified programs in performing calculations, and stated that 

uncert i f i ed programs must have an independent veri f i cat i on of accuracy. 

To evaluate FCF's use of ECPs, the inspectors reviewed the certification 

process for new versions of ECPs and the procedures for error reporting, 

correction, and access control. 

b . Observations and Fi ndi ~gs 

b.1 Software Certification Process 

The inspectors reviewed FCF's recent modification to the certification 

status of NEMC! 7.6 from fully certified to conditionally certified. 

change was due to development of NEMO 7.7, which included notification 

messages regarding the use of NEMO. 

from NEMO 7.6. 

7.6 for this reason. 

The 

The warning messages were excluded 

FCF chose to downgrade the certification status of NEMO 

FCF procedure BWNT 0902-06, "Software Certification," Revision 16, March 

31, 1995, proscribed an affectivity statement for the procedure applied 

to computer programs changing certification status due to errors within 

the code. The procedure requires changes to the certification status to 

be placed in the certification file. However, the inspectors determined 

that the change to NEMO 7.6 was not in the certification file. 

FCF agreed with the inspectors findings and added the change to the 

certificztion file. 

ambiguous because cert i f i cat i on changes for any reason , i ncl udi ng code 

errors, should be recorded in the certification file. FCF also agreed 

that the procedure governing ECP certification was ambiguous, and 

committed to modify the procedure. 

The inspectors concluded that the procedure was 

The inspectors concluded that the ambiguity in the software 

certification procedure and the resulting incorrect information on NEMO 

7.6 were considered weaknesses in FCF's computer code certification 

process. 

b.2 Engineering Computer Program Error Reporting 

The inspectors reviewed calculational files for consistency in ECP error 

reporting procedures. The Inspectors found that FCF had no specific 

procedures to address error reports in cal cul at i onal f i 1 es . Furthermore, the inspectors determined that procedure BWNT 0402-01 did 

not distinguish between the use of conditionally certified and fully 

certified ECPs. However, FCF stated that use of a conditionally 

certified code required justification in the calculation file, 

particularly if error reports exist for the code. 

- 23 - 

78 
The inspectors examined two cases involving the use of the NEM0.4.lRS 

code for TMI-1 Cycle 10. NEMO.4.1RS was a conditionally certified code 

which contained an error in the calculation of certain core average 

values. The first use of this version of NEW was found in BUNT 32- 

1219600-00, "TMI-1 Cycle 10 Flex Nuclear Data," September 17, 1993. The 

calculation file did not include a justification for the use of the 

code, even though an error record existed for the code. FCF agreed that 

this was an oversight and added a reference to the error in the 

calculation file. 

The second use of this version of the NEMO code was documented in BUNT 

32-1218970-00, "TMI-1 Cycle 10 Nuclear data/650 Cycle 9," September 17, 

1983, that also did not contain justification for use of the code. FCF 

also corrected this oversight during the inspection. 

Based on these findings, the inspectors concluded that inconsistency in 

addressing ECP error reports was a weakness in the use of ECPs. 

b.3 Benchmarking CASH0 3 and NEHO 

The applicability of the CASMO 3/NEMO neutronic methodology benchmarking 

to the new FCF reload designs was reviewed by the inspectors. 

topical report BAW 10180-A, "NEMO - Nodal Expansion Method Optimized," 

Revision 1, March 1993, was used as a reference by the inspectors. 

inspectors focused on the potential concern that the new fuel assemblies 

and/or core designs may have differed significantly from the range of 

fuel designs that were benchmarked in the topical report. 

The NEMO 

The 

The inspectors reviewed core loading patterns for TMI-1 Cycles 8 to 11, 

CR3 Cycles 10 and 11, ANO-1 Cycle 13, Trojan Cycles 12 and 13, and 

Sequoyah-1 Cycle 9. The inspectors noted that operating cycle lengths 

have increased over time and that some current reload designs were for 

650 to 670 EFPD (24 month cycle). The longer cycle designs necessitated 

the use of higher U235 enrichments (up to 4.96 w/o) with corresponding 

higher beginning of cycle (BOC) soluble boron concentrations (up to 2300 

ppm at BOC, HZP), or with higher burnable poison loadings. 

inspectors noted that the core designs, now being loaded or soon to be 

loaded, used up to 6 w/o Gd as the burnable poison and that some 

assembly designs used both Gd and the B,C lumped burnable poison. 

The 

Having determined the past and upcoming fuel designs, the inspectors 

focused on reviewing the range of designs that have been benchmarked to 

validate the CASMO 3/NEMO methodology. 

w3s the ability of NEMO to predict core power distributions. 

addition to the benchmark results published in the NEMO topical report, 

the inspectors reviewed power distribution comparisons from the 

following: 

The inspectors's primary concern 

In 

NEMO predictions vs measured incore detector data 

NEMO predictions vs PDQ predictions (two-dimensional (20) quarter 

core diffusion theory results) 

- 24 - 

79 
C. 

(c) 

(d) 

NEW predictions vs SIMULATE-3 predictions (a1 ternate 3D nodal 

code results) 

NEW predictions vs MCNP predictions (20 multi assembly Monte 

Carlo code results) 

Based on this review, the inspectors concluded that NEMO can accurately 

predict core power distributions within the accuracy stated in the NEMO 

topical report. The inspectors also concluded that the current and 

upcoming core designs were within the range of fuel design 

configurations and core parameters that were verified in the topical 

report. Additional evaluations were performed by the inspectors as part 

of the TMI-1 fuel corrosion review (see Section 3.4.b.2 of this report). 

Concl us i ons 

The inspectors concluded that the ambiguity regarding ECP certification 

changes found in procedure BWNT 0902-06, "Software Certification, " 

Revision 16, March 31, 1995, and the resulting incorrect information on 

NEMO version 7.6 were considered weaknesses in FCF procedures. 

The inspectors examined a number of calculation files to determine 

adequacy and consistency in ECP error reporting and found examples where 

ECP error reporting was either inadequate or non-existent. Based on 

these findings, the inspectors concluded that FCF's inconsistency in 

addressing ECP error reports was a weakness in the use of ECPs. 

3.6 

a. 

Enaineerina - Nanufacturinq Interface 

Inspect i on Scope 

The inspectors reviewed specific aspects of the reload analysis/fuel 

fabrication process including the role of the RELATE, the 

engineering/manufacturing interface and the use of the release 

authorization/appl icable documents 1 ist (RA/ADL) . 

b. Observations and Findings 

The inspectors found that the first stage of a reload contract involved 

development of the contract requirements document (CRD) , which contained 

initial licensee requirements for the fuel cycle, including a 

description of work to be performed, the scope of supply, quantity, and 

special features and customer requirements for the contract. After the 

CRD was issued, the preliminary fuel cycle design (PFCD) process 

typically developed three fuel designs, with the final decision left to 

the licensee based on operating requirements. 

fuel cycle design requirements (FCDR) document was prepared by the 

RELATE engineer. 

After customer input, the 

80 

The FCDR contained operating parameters, batch descriptions, control 

component descriptions, and other special and cycle-specific features. 

The inspectors determined that the FCDR can be revised as necessary and 

as more specific data was received from the licensee, such as exact 

- 25 - 
C. 

3.7 

cycle lengths. 

fuel cycle design (FFCD), the RA/ADL was prepared by the fuel design 

control group. 

Concurrent with development of the FCDR and the final 

The inspectors reviewed example RA/ADLs. 

base or contract design by listing the applicable design definition 

documents, part names, and part numbers and was used to release the 

fol 1 owing : 

The ADL defines a specific 

(a) 

(b) 

(c) 

product documentation as part of a procurement package 

documents to establ ish code appl icabi 1 i ty 

documentation for equipment that requires professional engineer 

certi f ication 

The RA was sent to manufacturing to authorize fabrication of components 

specified in the ADL. Fuel manufacturing began as the final fuel cycle 

design was nearing completion. During fuel assembly fabrication, 

communication between manufacturing and engineering can be in the form 

of concurrence requests (CR), deviation reports (DR), transmittal 

records (TR), and design change requests (DCR). 

The inspectors reviewed a number of these records from current core 

design contracts. 

control group. Procedure QC-1434, "Submittal of Manufacturing and 

Quality Documents to Fuel Engineering and Customers," described the 

methods by which manufacturing, quality, and supplier documents were 

transmitted to fuel engineering or to customers for approval (CR) or for 

information (TR) and the means by which such transmittals and approvals 

were documented. Design change requests (DCR) were sent from 

engineering to manufacturing to propose design changes to previously 

released nuclear fuel and core components based on contract specified 

designs. 

These transmittals were maintained by the fuel design 

Concl us i on 

The inspectors concluded that the interface between engineering and 

manufacturing and the use of the RA/ADL were performed in accordance 

with written instructions and satisfactorily completed for those 

activities affecting quality. 

Procurement 

a. Inspection Scope 

Eli 

The inspectors reviewed FCF's activities related to the procurement of 

material and services for use in safety-related products, the 

documentation that FCF maintained supporting the purchases of material, 

equipment, and services used in .he manufactwe c 5 safety-re1 ated 

products, and QA specification 09-1212-05, "Quality Assurance Program 

Requirements for Suppliers," Revision 5, February 11, 1992. 

- 26 - 
b. Observations and Findings 

FCF procedure BWNT-1719-25, "Quality System Surveys & Quality Audits of 

BWFC Suppliers," Revision 4, October 28, 1994, specified the 

requirements for qualifying vendors to supply safety-re1 ated material, 

equipment, and services. BWNT-1719-25 required that FCF perform an onsite quality audit to verify implementation of the applicable quality 

program (for initial approval and triennially thereafter) and also to 

perform an annual evaluation to review the vendor's recent activities 

with FCF. 

status list (SSL). 

The status of vendors was maintained by FCF on the supplier 

The inspectors reviewed several FCF audit reports, supporting placement 

on the SSL, for Suhm Coil Spring Works (BWFC 92-36), FCBC Romans Plant 

(BWFC 94-17), Gray Syracuse Report (BWFC 95-9), and Carpenter Techno1 ogy 

(BWFC 95-7). The audit reports documented the basis for the quality 

audits, results, conclusions, and findings. In addition, the report 

files contained documentation between FCF and the vendors which 

adequately resolved any findings. The inspectors concluded that the 

quality audits had been adequately performed and that FCF had maintained 

adequate documentation. 

The following purchase orders for materials were reviewed by the 

inspectors: 

(a) PO 40028, March 25, 1996, to Zircotube, Paimboeuf, France, for 

(b) PO 21872, January 23, 1995, to Gray Syracuse, Inc., for stainless 

(c) PO 40027, March 25, 1996, to Zircotube for guide tubes 

(d) PO 40051, March 25, 1996, to Cezus, Montreuil-Juigne, France, for 

fuel clad tubing 

steel castings to make end fittings 

end plug bar stock 

(e) PO 27542, July 5, 1995, to Cezus, Rugles, France, for Zirca OY 

spacer grid strip 

(f) PO 36317, January 19, 1996 to Ulbrich Specialty Strip Mill, 

Wallingford, Connecticut, for Inconel spacer gird strip 

(9) PO 4030, March 27, 1996, to Suhm for cruciform holddown spr . ng section manufacturing services for plenum springs 

for nickel alloy sheet for plenum syings 

(h) PO 2849, November 16, 1993, to Inco Alloys International, Inc., 

c . Concl us i ons 

The inspectors determined that specification 09-1212-05 invoked the 

requirements of 10 CFR Part 50, Appendix B, and that 10 CFR Part 21 was 

referenced in each of the POs reviewed. 

the procurements were performed in accordance with written instructions 

and satisfactorily completed for thc.:,e activities affecting quality. 

The inspectors concluded that 

- 27 - 

82 
3.8 Fuel Fabrication 

a. Inspection Scope 

FCF produced PUR fuel bundle and control components for B&W- and 

Westinghouse-designed reactors. They a1 so produced fixed incore 

detectors for B&W-, Combustion Engineering-, and Westinghouse-designed 

reactors. 

Mark-B9, Mark-B10, and Mark-B11 (the B&W designs), and 17x17 for MarkBW, the Westinghouse design. 

The FCF fuel bundle rod matrix design was 15x15 array for 

FCF control components consisted of Mark-B burnable poison rod 

assemblies (BPRAs) , Mark-B control rod assemblies (CRAs) , Mark-B axial 

power shaping rod assemblies (APSRAs), Mark-BW BPRAs (15x15 and 17x17) 

and rod cluster control assemblies (RCCAs). For the Mark-BW17 fuel 

bundles, FCF produced reconstitutable top nozzle assembl ies, 

intermediate spacer grid restraint system, intermediate spacer grid 

assemblies, end spacer grid assemblies, debris resistant bottom nozzles, 

and advanced debri s f i 1 ter bottom nozml es. 

During this inspection, FCF was not fabricating fuel rods or fuel bundle 

assemblies. Therefore, this portion of the inspection was limited to 

evaluating the fabrication of certain fuel bundle components. 

b. Observations and Findings 

b.1 Burnable Poison Rod Assembly 

During this inspection, FCF was fabricating MK-BW type BPRAs for use in 

Westinghouse-designed PWRs. The fabricat ion encompassed manufacturing 

burnable poison rods and thimble plugs, and fastening them to an upper 

structure assembly with appropriate hardware. The inspectors observed 

the fabrication of burnable poison rods and upper structure assemblies, 

and assembly of BPRAs, as well as inspections performed by product 

quality (PQ) personnel and reviewed selective records pertaining to the 

quality of components used to fabricate the BPRAs. The team identified 

no adverse observations in this area. 

b.2 Rod Cluster Control Assemblies 

During tnis inspection, FCF was fabricating RCCAs intended for use in 

Westinghouse-designed reactors. The main components of an RCCA were the 

spider, the control rods and the hardware to attach the control rods to 

the spider. The team observed the activities related to the fabrication 

of the control rods, and the final assembly of the RCCA including 

inspections by PQ, and reviewed records attesting to the quality of 

selected components. 

this area. 

The team identified no adverse observations in 

- 28 - 

83 
b.3 

b.4 

C. 

3.9 

a. 

End Caps 

The inspectors reviewed the end c-ap fabrication process as specified by 

RC 9110-20, "Fabrication of Fuel Rod End Caps." Receiving inspection of 

the end plug rod was performed in accordance with QC-595, "Receiving 

Inspection - Zircaloy Bar Stock and Tubing," Revision 4, October 13, 

1994. Manufacturing machined the end caps to the dimensions required by 

the applicable drawing and inspected 100% of the parts. PQ inspected 

the cays for dimensions and clean1 iness. Accepted product was forwarded 

to material control for storage. 

completeness and forwarded it, along with inspection data, to quality 

verification for retent ion. 

PQ reviewed the route card for 

The inspectors observed that cleaning operations were performed in 

accordance with MA-549, "Mi scell aneous Hardware C1 eani ng, I' Revision 4, 

July 28, 1994. Paragraph 3.6 stated that within the limitations 

specified in the product routing or elsewhere in MA-549, the combination 

and sequence of cleaning methods were left to the discretion of the 

cleaning technician. 

However, the inspectors reviewed the training records of the cleaning 

technicians and determined they had no recorded training regarding the 

selection and sequence of cleaning methods. The inspectors considered 

this a weakness in the FCF manufacturing system. (This issue is 

discussed further in Section 3.12.b.2 of this report.) 

Spacer 6r i ds 

The inspectors reviewed the spacer grid fabrication process as specified 

by RC 8500-11, "MK-BW Zircaloy Intermediate Spacer Grid Assembly 

Fabrication." Receiving inspection of the zircaloy strip was performed 

in accordance with QC-586, "Receiving Inspection - Zirconium Alloy Sheet 

or Strip," Revision 3, August 14, 1995. Manufacturing received released 

spacer grid strip material, recorded the applicable information for 

traceability on the route card (RC) and assembled the spacer grid 

strips. The inspectors considered the manufacturing and inspection 

activities suitable for the production of grid spacers for safety 

related applications. 

Concl usi ons 

Other'than the weakness identified in the cleaning of end plugs, the 

inspectors considered the manufacturing and inspection activities 

suitable for the production of the safety-related items. 

Cal i brat i on 

Inspection Scope 

The inspectors reviewed the FCF program for ensuring that equipment was 

properly controlled and maintained in calibration. FCF maintained a 

gage control laboratory staffed by a gage control technician who was 

- 29 - 

84 
responsible for control1 ing the status and maintaining the records 

associated with equipment calibration. FCF calibrated most of the 

equipment used in the facility with the exception of certain specialty 

machines, such as the Baldwin tensile tester and the standards used to 

perform the calibrations, such as gage blocks. 

b. Observations and Findings 

FCF maintained +he calibration records for the approximately 2,500 items 

in a hard copy filing system arranged by date of calibration (month or 

month/day). 

this system and was able to provide the inspectors with the supporting 

documentation for all calibrations reviewed. The inspector's 

calibration reviews included the gage envelope dial indicator (QC14-17) , water channel gage standard fixture (QC-1949) , micrometer inspection 

gage (QC-1578), guide tube alignment gage (QC-4), lower end fitting 

casting template (QC-1579), lower end fitting casting inspection gage 

(QC-1578), and lower end fitting true position functional gage (QC-53). 

FCF was able to provide documentation supporting traceability of the 

calibrations reviewed to the appropriate standards including the 

National Institute of Standards and Technology (NIST). 

The gage control technician was efficient in the use of 

FCF procedure QC-1405, "Control of Measuring & Test Equipment," Revision 

5, January 31, 1995, paragraph 4.5 stated, in part, that calibration 

services were classified as commercial grade and the measuring and test 

equipment were classified as commercial grade items. However, when the 

equipment was used to verify the capability of safety-related items, the 

equipment had a safety-related function to verify the parameters defined 

for acceptability of the safety-related item. Accordingly, FCF took 

actions to qualify the comercial grade service provided by calibration 

suppl iers, to ensure sui tabi 1 i ty for the intended safety-re1 ated 

application, by performing an audit of the suppliers technical and 

quality program. FCF used MIL-STD 45662A, "Calibration System 

Requirements," as the basis for the audits. 

After successful completion of the audit, the calibration suppliers were 

listed on the SSL. 

supporting placement on the SSL, for AA Jansson, Inc., (BWFC 95-l), 

American Electronics Laboratory (BWFC 95-18), and Gage Laboratory (BWFC 

93-35), which indicated that tr!e audits were adequately performed and 

documented. In addition, the inspectors reviewed POs to calibration 

suppliers for AA Jansson, Inc. (PO 36241, January 12, 1996), American 

Electronic Lab (PO 38981, March 5, 1996), and Gage Lab (PO 30301, August 

17, 1995), and determined that they contained adequate technical and 

quality requirements to assure performance of the requested service. 

The inspectors concluded that FCF had taken adequate actions to verify 

and document the quality of the calibration services used to support 

saf2ty-re1 ated work. 

The inspectors reviewed the FCF audit reports, 

- 30 - 

85 
C. 

3.10 

During an inspection of production line equipment, the inspectors noted 

that the Baldwin tensile tester, identified as FCF QC-519, exhibited a 

"Satec" calibration sticker which stated that the machine had been 

calibrated on March 8, 1995, and was due for calibration on March 8, 

1996. There was no FCF calibration sticker present on the machine. The 

inspectors noted that the machine was not administratively out of 

calibration due to procedure QC 800, "Gage Control," paragraph 5.2, 

which allowed for the next calibration to be performed at any time 

during the calendar month that the previous calibration expired such 

that the calibration on the Baldwin machine was acceptable for FCF work 

through March 31, 1996. 

Discussion with FCF personnel indicated that the Baldwin machine was 

intended to be used in its present form for only an additional two 

months and at that time it was to be modified. Based on consideration 

for the short period of intended use, FCF had extended the calibration 

interval to eighteen months and had documented the extended calibration 

interval in QC 800. 

twelve months to eighteen months was ASTM E-4-94, Section 20, "Time 

Interval Between Verifications" which stated, in part, that it was 

recommended that testing machines be verified annually or more 

frequently if required and in no case shall the time interval between 

verifications exceed 18 months. 

FCF indicated that the basis for the extension from 

The inspectors determined that FCF had not established a technical basis 

for the calibration interval extension. The inspectors also noted that 

since the previous March 8, 1995, calibration had been performed by 

Satec, FCF may consider Satec's input when establishing the technical 

basis. 

FCF contacted Satec who provided a letter dated March 27, 1996, that 

stated that the decision to extend the current verification interval 

rested with FCF, that the machine could probably be used without a 

problem, but that FCF should have the machine's calibration reverified 

as soon as possible. 

letter did not provide a technical basis for extension of the 

cal i brat i on interval . 

However, the inspectors determined that the Satec 

Concl us i ons 

The inspectors concluded, based on the review of the FCF documentation 

and the Satec letter, that FCF had not established a technical basis for 

using the Baldwin machine past its original calibration date (and its 

administrative extension of March 31, 1996). This was identified as a 

weakness in the implementation of the FCF calibration program. 

Fi el d Services 

Field Services was divided into three functional groups; engineering, 

outage operations and equipment operations. The Field Services quality 

assurance activities were governed by the same topical report that 

governed QA activities in other operations at FCF, document 56-1177617- 

- 31 - 

86 
03, "Quality Assurance Program for B&W Fuel Company," May 30, 1994. 

Procedure MA-580, "Specialized Tooling and Equipment," Revision 0, 

October 12, 1995, set forth the guidelines and responsibilities for the 

design, implementation, documentation and control of specialized tooling 

and equipment bearing on the outcome of manufactured product quality. 

On the basis of its review, the inspectors concluded that the QA program 

for field service activities was adequate. 

3.11 Oualitv Action Requests 

The inspectors reviewed the quality action requests (QARs) program and 

the status of several open QARs and concluded that FCF personnel were 

sufficiently responsive to QARs. 

3.12 Training 

a. Inspection Scope 

The inspectors evaluated the FCF training program in both the 

engineering and fabrication areas. 

b. Observation and Findings 

b.1 Engineering 

The inspectors evaluated the adequacy of training in the engineering and 

services units primarily through interviews of management and staff, 

including interviews with unit managers in the fuel management and 

operations analysis, thermal and performance analysis and engineering 

analysis services units. The inspectors considered training to be made 

up of the following two general areas: 

(a) direct, job-related technical training 

(b) non-technical or administrative training 

The inspectors also evaluated training for new employees and training in 

problem solving techniques. Procedure BWNT-1702-22, "Employee 

Training," Revision 16, October 10, 1994, provided requirements for 

training. The procedure stated, in part, that all initial job and ongoing quai ity-related training shall be documented on a personnel 

training report (PTR). 

in the procedure as training on new or revised QA program manuals, the 

records management program manual, working instructions and new or 

revised qual ity-related policies, procedures, and forms. 

In general, the inspectors found that no formal requirements exist for 

direct technical training, such as the use of engineering computer 

codes. 

On-going qual ity-related training was described 

Training on administrative procedures was the only training that 

- 32 - 

87 
was officially documented. The inspectors examined available training 

records for one engineering unit. A1 though substantial records existed 

on administrative procedure training and software training, only one 

record was found for training directly applicable to job performance. 

b.2 Fuel Fabrication 

FCF had divided the training of employees into two distinct categories, 

inspection personnel and non-inspection personnel. The inspectors found 

that procedures BWNT-1702-22, "Employee Training," Revision 16, October 

10, 1994, applied to all FCF employees and that QC-1440, "Qualification 

of Inspection Personnel ,*I Revision 9, May 31, 1995, applied to personnel 

performing inspections to verify conformance to specified requirements 

for the purpose of acceptability. 

The inspector reviewed BWNT-1702-22 that stated that employee training 

was the responsibility of the management of each organization and that 

each supervisor should evaluate the training needs of the employees 

within the group on a continuing basis. All initial job and on-going 

quality related training was to be documented on a PTR. The specific 

training required by BWNT-1702-22 included initial job training to be 

accomplished prior to beginning work and documented on a PJR. 

going quality related training was to be provided by the applicable 

manager. 

The onThe inspectors reviewed the training records for a selected group of 

employees from various areas including QC, PQ, final assembly, gage 

control , metallurgical laboratory, assembly room, machining, and 

manufacturing. FCF provided records for the personnel performing 

inspection activities which had documentation of the initial and ongoing quality related training. 

inspection personnel training was in accordance with the requirements of 

QC-1440 and adequately documented. 

The inspectors concluded that the 

However, for non-inspection personnel , FCF could provide no records 

documenting either initial training or on-going quality training. FCF 

management indicated that this training had occurred but there had 

previously been no requirement to document the completion of the 

training. 

C. Concl us i on 

Through review of FCF procedures and interviews with engineering 

management and staff, the inspectors found that no formal requirements 

exist for direct technical training, such as the use of engineering 

computer codes. Training on administrative procedures is the only 

training that was officially documented. 

lack of formal requirements for tecinical training and the limited 

incorporation of technical training to be a weakness. 

The inspectors considered this 

- 33 - 
The inspectors did not identify any examples of production activities 

having been affected by the lack of documented training. 

inspectors concluded that the lack of documentation for the training of 

non-inspection personnel was a weakness in the implementation of the QA 

program. 

However, the 

3.13 Continuous Imorovement Proaram 

The inspectors also interviewed the manager of the FCF continuous 

improvement program (CIP) , part of the total qual i ty management program 

at FCF. Employees trained in CIP generally lead or were a member on a 

qual i ty improvement teams (QIT) . Currently, 17 qual i ty improvement 

teams have been formed. 

program which included identifying, defining and analyzing a problem and 

determining and implementing a solution. Some topics of recent QITs 

include eliminating sources of fuel failures and improvements in fuel 

rod welding. The final report on QIT B5 on fuel integrity was examined 

as an example. The Fuel Integrity QIT evaluated approximately 200 ideas 

submitted as the root cause of the unknown fuel failures. Formation of 

the inspectors, its processes for finding the root cause, the results of 

that effort, and recommendations for corrective action were included in 

the report. The program appears to be an effective method for improving 

product qual i ty . 

The CIP program was based on a seven step 

3.14 Entrance and Exit Reetinlps 

In the entrance meeting on March 18, 1996, the NRC inspectors met with 

members of FCF management and staff, and discussed the scope of the 

inspection. 

proprietary information, as well as those of FCF. In addition the 

inspectors establ ished contact persons within the management and staff 

of the applicable FCF organizations. 

The team also reviewed its responsibilities for handling 

89 

During the exit meeting with FCF management and staff, on April 17, 

1996, the inspectors discussed its findings and concerns, as well as 

FCF’s weaknesses. 

- 34 - 
Figure 1 

TMI-1 Cycle-10 Arrangement of the "1" Loading Patterns 

1 234 5 6 789 10 11 12 13 14 15 

A 

6 

C 

D 

E 

F 

6 

H 

K 

L 

U 

N 

0 

P 

R 
PARTIAL LIST OF PERSONS CMACTED 

Andrews, B. 

Armentrout, C. 

Bohart , J. 

Carr, C. W. 

Coleman, T. 

Cud1 in, 3. 

Deveney , R. C . Engel ke, W. 

Ford, 3. T. 

Gardner, R. L. 

Hanson, G.E. 

Harlow, N. 

Hobson, G.H. 

Lamanna, 1. 

Matheson, J. E. 

Mayberry, J. R. 

McPhatter, F. 

Meyer, G. 

Vice President, Fuel Engineering 

Lead, Qual i ty Audits and Programs 

Pres i dent 

Vice President, Manufacturing and Service 

Vice President 

Unit Manager, Analysis Services 

Benchmarking/Methodol ogy (Supervi sor Nuclear Techno1 ogy) 

Manager, Process Quality 

Manager, Fuel Manufacturing 

Manager, Qual i ty Benchmarking/Methodol ogy (Manager Fuel Engineering) 

Technical Specialist Benchmarki ng/Methodol ogy (Principal Engineer) 

Manager, Chemistry Engineer 

Manager, Design and Devel opment 

Manager, Product Qual i ty 

Pri nci pal Engi neer 

Manager, Thermal and Performance Analysis 

ITEM OPENED AND CLOSED 

ODened 

none 

C1 osed 

99900001/90-01-01 NON 

99900001/90-01-02 NON 

99900001/91-01-01 NON visual inspection of tubes with felt plugs 

99900001/91-01-02 NON 

99900001/91-01-03 NON 

99900001/91-01-04 URI 

burr on lower end fitting 

radiographs as qual i ty records 

issuance of COR and CVAR 

adequate guidance to operators and inspectors 

root cause for the failure to detect felt plugs left 

in tubes 

- 36 - 

91 
ACROUVIIS USED 

ADL 

APSRA 

BOC 

BPRAs 

B&W 

BWFC 

BUST 

CDR 

CFR 

CFT 

CIP 

CNFP 

COLR 

CR 

CR3 

CRA 

CRD 

CVAR 

DCP 

DCR 

A 

DNB 

DNBR 

DR 

DRN 

ECCS 

ECP 

EFPD 

FCDR 

FC F 

FFCD 

FOAK 

FSAR 

FT I 

Gd 

GDC 

GPUN 

HFP 

HZP 

LBVs 

LHR 

Li 

LOCA 

MTC 

NIST 

NRC 

PCT 

PFCD 

ANO- 1 

Applicable Documents List 

Arkansas Nuclear Unit 1 

Axial Power Shaping Rod Assembly 

Beginning Of Cycle 

Burnable Poison Rod Assembl ies 

Babcock and Wilcox Company 

B&W Fuel Company 

Borated Water Storage Tank 

Component Di screpancy Report 

Code of Federal Regulation 

Core Flood Tank 

Continuous Improvement Process 

Commercial Nuclear Fuel Plant 

Core Operating Limits Report 

Concurrence Request 

Crystal River Unit 3 

Control Rod Assembly 

Contract Requirements Document 

Contract Variation Approval Report 

Distinctive Crud Pattern 

Design Change Request 

Delta (Different i a1 ) 

Departure From Nucleate Boi 1 ing 

Departure From Nucleate Boiling Ratio 

Deviation Reports 

Document Release Notice 

Emergency Core Cooling System 

Engineering Computer Programs 

Effective Full Power Days 

Fuel Cycle Design Requirements 

Framatome Cogema Fuel s 

Final Fuel Cycle Design Fi rs t-Of-A-Ki nd 

Final Safety Analysis Report 

Framatome Technologies, Inc. 

Gadol i n i um 

General Design Criterion 

GPU Nucl ear Corporation 

Hot-Full -Power 

Hot-Zero-Power 

Licensing Basis Values 

Linear Heat Rate 

Li thi um 

Loss-of-Coolant Accident 

Moderator Temperature Coefficient 

National Institute of Standards and Technology 

U. S. Nuclear Regul atory Commi ssion 

Peak Cladding Temperature 

Preliminary Fuel Cycle Design 

- 37 - 

92 
P#l 

w 

PPrn 

PQ 

PSC 

PTR 

PUR 

QA 

QAR 

QC 

QIT 

RA/ADL 

RALS 

RC 

RCCA 

RCS 

RELATE 

RPS 

SBLOCA 

SRP 

SSL 

3D 

TMI-1 

TR 

TVA 

20 

w/o 

Program Manager 

Purchase Order 

Parts Per Mi 11 ion 

Product Qual i ty 

Prel iminary Report of Safety Concern 

Personnel Training Report 

Pressurized-Water Reactor 

Qual i ty Assurance 

Qual i ty Act i on Request 

Qual i ty Control 

Quality Improvement Team 

Release Authorization/Appl icable Documents List 

Reload Analysis and Licensing Services 

Route Card 

Rod C1 uster Control Assembly 

Reactor Cool ant System 

Reload Licensing Analysis Task Engineer 

Reactor Protection System 

Small Break Loss-of-Cool ant Accident 

Standard Review P1 an 

Suppl ier Status List 

Three-Dimensi onal 

Three Mile Island Unit 1 

Transmittal Records 

Tennessee Valley Authority 

Two-Dimensional 

Weight Percent 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 205q-1 

November 4. 1996 

Mr. Lawrence M. Froman, Manager 

Panalarm Business Unit 

Process and Analytical Instruments Division 

AMETEK, Inc. 

7401 North Hamlin Avenue 

Skokie, IL 60076 

SUBJECT: NRC INSPECTION REPORT 9?901303/96-01, NOTICE OF VIOLATION, AND 

NOTICE OF NONCONFORMANCE 

Dear Mr. Froman: 

On September 12, 1996, the U.S. Nuclear Regulatory Commission (NRC) completed 

an inspection at your Panalarm Business Snit facility. 

presents the results of that inspection. 

The enclosed report 

During this inspection, the NRC inspector found that one of your activities 

appeared to be in violation of NRC requirements. Specifically, your procedure 

for identifying, evaluating, and reporting deviations from the requirements of 

safety-re1 ated purchase orders confused the terms "deviation" and "defect"; it 

also failed to address significant changes in 10 CFR Part 21 reporting 

requirements and procedures since the procedure was last revised more than six 

years ago. 

The violation is cited in the enclosed Nztice 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 enclo 

Please note that you are required to resDond to this letter 

preparing your response. The NRC wiil use your response, 

determine whether further enforcement act ion is necessary 

with regulatory requirements. 

In addition, the NRC inspector found that the implementat 

assurance Droaram failed to meet certain NRC reauirements 

id NOV when 

in part, to 

to ensure compliance 

on of your quality 

," imposed on you by 

your customers. 

organizational independence and verification functions of your qual ity 

assurance group as required by 10 CFR Part SO., Appendix 8. 

nonconfwmances were identified in this regard. 

Specifically, your procedures failed to adequately define the 

Three 

94 

These nonconformances are cited in the enclosed Notice of Nonconformance 

(NON), and the circumstances surrounding them are described in detail in the 

enclosed report. 

should follow the instructions specified in the enclosed NON when preparing 

your response. 

You are requested to respond to the nonconformances and 
Hr. Froman -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 (PDR) . 

Sincerely , 

Robert M. Gallo, Chief 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Docket No. 99901303 

95 
NOTICE OF VIOLATION 

Panalarm Division, METEK, Inc. 

Skokie, Illinois 

Docket No. : 99901303 

During an NRC inspection conducted on September 9 through 12, 1996, a 

violation of NRC requirements was identified. 

Statement of Pol icy and Procedure for NRC Enforcement Actions, " NUREG-1600, 

the violation is listed below: 

In xcordance with the "General 

10 CFR 21.21, "Notification of failure to comply or existence of a defect and 

its evaluation," requires, in part, that each corporation subject to the 

regulations adopt appropriate procedures to ensure the evaluation and proper 

reporting of deviations and failures to comply. 

Contrary to the above, Qual i ty Control Procedure QC-90-100, 'I( lOCFR21) 

Notification Procedure," Revision 0, dated March 6, 1990, failed to address 

the identification or evaluation of deviations. The procedure incorrectly 

confused the terms "deviation" and "defect", with the result that personnel 

were not alerted to the need to identify deviations from safety-related 

purchase order requirements, and to evaluate the deviations to determine if 

they could become defects. 

significant changes that were incorporated into 10 CFR Part 21 since 1990 

concerning reporting requirements and the content of procedures. 

Procedure QC-90-100 also failed to reflect 

( 9990 1303 / 96-0 1-0 1 ) 

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

Pursuant to the provisions of 10 CFR 2.201, Panalarm Division of AMETEK, Inc., 

is hereby required to submit a written statsiltent 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 4th day of November, 1996. 

96 

Enclosure 1 
NOTICE OF NONCONFORMANCE 

Panalarm Division, AMETEK, Inc. 

Skokie, Illinois 

Docket No.: 99901303 

Based on the results of an inspection conducted on September 9 through 12, 

1996, it appears that certain of your activities were not conducted in 

accordance w i t h NRC requ i remen t s . 

Criterion I of Appendix B to 10 CFR Part 50, "Organization," requires, in 

part, that the quality assurance functions shall verify that activities 

affecting safety have been correctly performed. The authority of persons 

performing activities affecting safety-re lated functions shall be clearly 

established and delineated in writing. 

assurance functions shall have sufficient authority and organizational freedom 

to identify qual i ty problems. 

The persons performing quality 

A. Contrary to the above, Paragraph 12.6 of Section 12, "Test Control," of 

the "AMETEK, Inc., Panalarm Division Quality Assurance Manual ,'I Document 

900181, Revision 10, dated December 8, 1993, specifies that documented 

test results are approved by engineering and/or quality assurance. 

Criterion I does not allow substitution of engineering for quality 

assurance review of final acceptance test results. (99901303/96-01-02) 

6. Contrary to the above, the organization chart on page 3-2 of the 

"AMETEK, Inc., Panal arm Division qual i ty Assurance Manual, 'I Document 

900181, Revision 10, dated December 8, 1993, does not reflect the 

current organizational status of qual ity assurance. Furthermore, 

procedures such as Qual i ty Control Procedure QC-76-47, "Training of Test 

and Inspection Personnel," dated April 25, 1986, referred to the QC 

supervisor and QC department, whereas the quality assurance manager 

stated thdt reference to his position was intended. (99901303/96-01-03) 

C. Contrary to the above, the Panalarm inspector performing the final 

acceptance testing for skfety-related temperature switch modules 

reported to a production supervisor, so that there was no independent 

quality assurance verification of the testing. (99901303/96-01-04) 

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) a description of steps that have been or 

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

been or will be taken to prevent recurrence; and (3) the dates your corrective 

actions and preventive measures were or will be completed. 

Dated at Rockville, Maryland 

this 4th day of November, 1996. 

-2- 

98 
Report No: 

Organization : 

Contact: 

Nuclear Industry 

Activity: 

DATES : 

Inspector: 

APPROVED BY: 

U. S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

99901303/96-01 

Panalarm Business Unit 

Process and Analytical Instruments Division 

AMETEK, Tnr 

7401 North Ham1 in Avenue 

Skokie, Illinois 60076 

A.H. Demerer, Quality Assurance Manager 

(847) 509-7196 

Annunciator systems including temperature 

moni tori ng modules 

September 9-12, 1996 

Richard C. Wilson, Senior Reactor Engineer 

Gregory C. Cwalina, Chief 

Vendor In spec t i on Sect i on 

Special Inspection Branch 

Division of Inspection and Support Programs 

99 

Enclosure 3 
1 INSPECTION SUUMRY 

During this inspection, the NRC inspector reviewed the implementation of 

selected portions of the AHETEK, Inc., Panalarm Division quality assurance 

(QA) program, and reviewed activities associated with the supply of safetygrade replacement parts for annunciator systems supplied by Panalarm for 

nuclear power plants in the 1970s and early 1980s. 

In the early 1970s the Panal it company produced Partalarm annunciator systcllis. 

The company name was changed to Riley, and subsequent owners were US Filter 

and Ashland Oil, prior to acquisition by AMETEK. Panalarm has approximately 

100 employees. Nuclear activity amounts to about 0.1% of sales. 

The audit 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 

Reaulations (10 CFR Part 50) 

0 10 CFR Part 21, "Reporting of Defects and Noncompliance" 

During this inspection, a violation of NRC requirements was identified and is 

discussed in Section 3.1 of this report. 

During this inspection, three instances where Panalarm failed to conform to 

NRC requirements imposed upon them by NRC licensees were identified. These 

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

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

This was the first NRC inspection of Panalarm. 

3 

3.1 

INSPECTION FINDINGS AND OTHER COMMENTS 

10 CFR Part 21 Proqram 

The inspector reviewed Panal arm's procedure for reporting in accordance 

with 10 CFR Part 21: Quality Control procedure QC-90-100, "(lOCFR21) 

Notification Procedure," Revision 0, dated March 6, 1990. The procedure 

focused on evaluating a defect to determine if it was a reportable 

condition; it stated that if the defect was determined to not be a 

repo~able condition, then no further action would be taken. 

Procedure QC-90-100 incorrectly confused the terms "deviation" (which is 

not used in the procedure) and "defectf' (which is incorrectly used). As 

a result, Panalarm personnel were not alerted to the need to identify 

deviations from purchase order requirements, and to evaluate the 

deviations to determine if they colild become defects. Further, the 

evaluation of defects as used in the procedure is not required, nor 

would it serve any useful purpose. The procedure also did not address 

significant changes to Part 21 in the years since the procedure was 

2 

100 
written, regarding reporting requirements and content of procedures 

a result, the procedure failed to comply with 10 CFR Part 21. 

Failure to have procedures that address the reporting and procedura 

content requirements of 10 CFR Part 21 constitutes Violation 

99901303/96-01-01. 

3.2 Oual i tv Assurance Proaram 

a. JnsDect ion SCOD~ 

As 

The inspector selectively reviewed the "AMETEK, Inc., Panalarm Division 

Quality Assurance Manual, " Document 900181, Revision 10, dated December 

8, 1993, and the "AMETEK Panalarm Division Quality Control Procedures 

Manual," Document 900004, dated May 2, 1986, which described the 

Panalarm quality assurance (QA) program. During the inspection the 

inspector asked about a quality control (QC) procedure that was not 

included in the Procedures Manual, and the QA manager determined that an 

obsolete copy of the manual had inadvertently been provided. The 

inspector then reviewed the table of contents for the January 28, 1994, 

revision of the QC Procedures Manual, which identified 30 additional QC 

procedures. 

commercial grade item dedication, had already been identified and 

reviewed during the inspection. 

three others. 

Two of these, covering 10 CFR Part 21 reporting and 

The inspector selected and reviewed 

b. Observations and Findinas 

Section 1 of the QA manual stated that the program is based on the 

American National Standards Institute standard ANSI N45.2-1971, and 

Section 2 stated that products supplied on purchase orders requiring 

compliance with Appendix B to 10 CFR Part 50 will be processed using the 

procedures described and referenced in the QA manual. Revision 0 of the 

manual was dated May 5, 1977. 

Panalarm management advised the inspector that Panalarm was considering 

changing the QA program to IS0 format and was also considering ending 

their Appendix B program. Neither decision had been made by the 

conclusion of the inspection. 

The inspector had two concerns with Panalarm's QA program. 

relates to QA verification of the final acceptance testing of Model 86 

modules, which is the principal basis for considering them to be safetygrade, as described in Section 3.3 below. In Section 12, "Test 

Control," of the QA manual, paragraph 12.6 stated that documented test 

results are approved by engineering andlor QA. Engineering review of 

the test results is a means of ensuring quality. 

was little independent QA review of other activities affecting quality, 

such a review of the final acceptance testing is necessary to satisfy 

Criterion I of Appendix 8 to 10 CFR Part 50, "3rganization," which 

requires the independent QA organization to verify that activities 

affecting safety-re1 ated functions have been correctly performed. 

The first 

However, because there 

3 

101 
C. 

3.3 

a. 

b. 

The inspector also determined that Papalarm procedures did not reflect 

the current Organization structure. The organization chart dated July 

1996 that was provided to the inspector differed from the chart included 

in Section 3, "Organization", in QA Manual Revision 10 dated December 8, 

1993. The QA manual chart shows the QA function reporting to the 

operations manager, with no alternate path specified; the July 1996 

chart shows the QA manager reporting to the production manager, but with 

a dashed 1 ine representing an administrative 1 ink directly to the 

division vice president and manager. 

confusing references to QA and QC positions. 

QC-76-47, "Training of Test and Inspection Personnel ," dated April 25, 

1986, referred to the QC Supervisor and the QC department, but the QA 

manager stated that reference to his position was intended. 

clarification is significant because Panalarm's production department 

has QC personnel who lack ?be organizational independence required by 

Criterion I of Appendix 8 to 10 CFR Part 50. 

, 

Several procedures a1 so contained 

For example, QC Procedure 

This 

Conclusions 

Panal arm's fai 1 ure to ThermocouDl e Monitor requi re QA veri f i cat 1 on of 

final acceptance testing con;t;iutes Nonconformance No. 99901303/96-01- 

02. Panalarm's failure to provide clear procedural r'eferences to the 

independent QA function where required constitutes Nonconformance 

NO. 99901303/96-01-03. 

Model 86 ThermocouDle Monitor 

InsDection ScoDe 

Prior to the inspection, the QA manager identified the Temp-Matic (or 

TempMatic--Panal arm 1 i terature uses both terms) Model 86 thermocouple 

monitor as the only safety-related component currently supplied. The 

inspector reviewed test and dedication procedures and reports, and other 

pertinent documentation, and observed a final acceptance test 

demonstration. 

Observations and Findinas 

The Model 86 analog module accepts a thermocouple input, compares it 

with an adjustable setpoint, and provides status light and contact 

outputs. Variations include a differential unit accepting two 

thermocouple inputs, a "half-gain" version, and a burnout protection 

feature. The module consists of two small printed circuit boards with 

several interconnecting wires, joined by standoffs and bolts to form an 

assembly about the size of a pack of cigarettes. 

case, and is intended for insertion into an annunciator grid structure 

in the control room. 

temperature switch. 

annunciator systems. It is sup,~lied in small quantities as a 

rep1 acement part. 

The module has no 

licensees commonly refer to the module as a 

The Model 86 is not part of current Panalarm 

4 

loa 
The inspector reviewed three 1973-4 qualification type test reports and 

verified that they identified the test specimens clearly enough to 

perrnit determining the similarity of current units. The inspector also 

reviewed subsequent engineering change notices that documented design 

reviews. 

changes of the Model 86 modules. 

These documents appeared to be adequate for tracing design 

Panalarm completed a dedication procedure for the Model 86 temperature 

modules in 1992 that provided the fallowing: 

0 Document 900396, "Model 868 Temp-Mat i c EPRI Contormi ty , I' 

Revision 0, dated November 5, 1992 (currently in Revision 3 dated 

January 30, 1995), identified each part in the Model 86 versions 

as critical or non-critical and tabulated critical 

characteristics and their verification methods for the critical 

parts. 

Quality Control Procedure QC-92-110, "Commercial Grade Item 

Dedication for Safety Related Assemblies - Quality Control 

Procedures for Dedicated Items," Revision 0, dated July 30, 1992, 

specified batch receipt inspection and controlled handling 

practices for critical parts; it also required training for 

personnel performing these activities, and for testing personnel. 

Notes referencing these two new documents were added to the two 

assembly level drawings for Model 86 modules, 86800-P-2 and 

86800-V-1. 

The NRC inspector found that the Model 86 parts dedication activities 

did not address possible heterogeneity of commercial grade parts lots, 

and no audits of commercial grade parts suppliers were performed. 

However, the receipt inspection did verify visual and performance 

characteristics for many parts, which were then subject to controlled 

handliny. The principal basis for considering the completed Model 86 

modules to be safety grade was the final acceptance testing, which was 

relied on for dedication of some parts as well as performance of the 

assembl ed modul e. 

The acceptance test procedure written for GE contracts--Document No. 

108-8022-900035, "Final and Acceptance Test Procedures. for Model-86 

Tempmatic Modules for General Electric [GE] Nuclear Energy Division 

Jobs," Revision 16, dated December 29, 1989, was more comprehensive than 

the previous procedure, Document No. 900024, "Calibration and Final Test 

Procedures for Temp-Matic Model 86," Revision 2, dated July 11, 1975. 

The new procedure repeats more of the in-process calibration points; 

provides more detail for model variations; includes a check for input 

loop resistance; tests the optional burnout protection; and requires 24 

hours post-test operation. Although the new procedure's title would 

appear to restrict its use to POs from GE, this is apparently not the 

case. 

for dedicating Model 86 switches. This document in turn requires testing in accordance with the revised procedure to verify certain critical 

5 

103 

Dedication procedure QC-92-110 requires use of procedure 900396 
characteristics at the module level. As discussed below, an IES 

Utilities audit report stated that they would continue to invoke 

QC-92-110 and 900396 in POs. 

Although no safety-related work was in progress during the inspecti 

Panalarm demonstrated final and acceptance testing for a Model 86 m,Jle 

using Procedure 108-8022-900035. 

concerns with the procedure: 

distingi'ish between test values for standard ad half-gain modules; 

Figure 2 specified an incorrect wire type for type J thermocouples; and 

the procedure did not specify that high and low range, but not setpoint, 

potentiometers are sealed with glyptol. In each case the Panalarm 

inspector was aware of the correct procedure, and explained it. The NRC 

inspector did not consider these concerns significant; Panalarm agreed 

to correct them. 

, 

The NRC inspector identified a few 

steps 6 and 7 of Part A did not 

The NRC inspector noted that the Panalarm inspector performing the final 

acceptance test reported to a production supervisor, and there was no 

oversight by the QA organization. 

with Criterion I of Appendix B, "Organization", which requires the QA 

organization to verify that activities affecting safety-related 

functions have been correctly performed. 

to immediately notify affected personnel that either the QA manager or 

the QC technician reporting to him must witness the final acceptance 

testing of safety-related equipment. 

procedural change depends on Panalarm's decision concerning continued 

supply of safety-related equipment. 

This practice was not in conformance 

The Panal arm QA manager agreed 

The schedule for a written 

The inspector found no purchase orders for non-GE procurement of 

Model 86 modules. 

identified by the QA manager did nor constitute a complete set of recent 

safety-re1 ated orders. Subsequentl;t, an additional PO from a 1 i censee 

was traced through a file cabinet containing certificates of conformance 

to a separate set of PO files, but time did not permit pursuit of 

possible additional safety-related POs through this path. The 

additional PO was for the Entergy Operations procurement of isolators 

descri bed bel ow. 

This was partially because the POs initially 

Since the Model 86 dedication program was performed in response to GE 

requirements, and the acceptance test procedure for GE orders was more 

extensive than the older Panalarm procedure, it would be necessary to 

examine specific 1 icensee procurement requirements to determine whether 

direct licensee procurements of Model 86 modules would be acceptable. 

As discussed below, an audit report by IES Utilities stated that their 

safety-related POs would continue to invoke Panal arm procedures that 

require the use of the GE acceptance test procedure, but no related 

procurement records were available to the NRC inspector within the 

available time 1 imits. 

c. Conclusions 

6 

104 
3.4 

a. 

b. 

The inspector concluded that Panalarm’s QA program for Model 86 modules 

supplied to GE was acceptable. The acceptability of direct licensee 

procurements would depend on the specific requirements of POs. 

Panalarm’s failure to perform independent verification of final 

acceptance testing constitutes Nonconformance No. 99901303/96-01-04. 

Purchase Orders 

The inspector reviewed three POs from GE and two from Entergy 

Operations, Inc. 

as described below. 

The POs covered a variety of equipment and services, 

a: GE Nuclear Energy PO No. 52896020591 dated April 19, 1996, ordered 

three Model 86-BVTFF-EGCUST temperature switches. 

the switches were safety-related, invoked 10 CFR Part 21, and required 

the items to be either purchased safety-grade, dedicated, or evaluated 

as performing no safety-re1 ated function. The switches were del i vered 

to GE San Jose, California, for use at the Grand Gulf power station. 

Panalarm provided acceptance test data sheets which identified the test 

instruments, and certification of conformance to contract requirements 

which identified customer drawing no. 164C5687P302, Revision 16. The 

data sheets were signed as attested to by the QA manager two days after 

the production inspector’s signoff. Except for this latter instance of 

Nonconformance 99901303/96-01-04, the NRC inspector found the 

documentation for the PO to be acceptable. 

The PO stated that 

GE Nuclear Energy PO No. 205-95L289 dated March 17, 1995, ordered two 

temperature switches, GE drawing no. 16465687P320. The switches were 

delivered to GE San Jose for use at the Nine Mile Point unit 2 (NMP2) 

power plant. 

former quality control manager on April 27, 1995; there is no evidence 

that he actually witnessed the testing, although that was the same day 

as the production inspector’s signoff. 

concerns with the documentation for this PO. 

In this case the data sheets were attested to by the 

The inspector had no other 

The inspector also briefly reviewed GE PO No. 205-95L600 dated July 13, 

1995. 

from NMP2, and required scrapping the switch after submittal of the 

report. The PO stated that, when the switch was placed in service on 

January 7, 1991, a plant operator touched the switch and it spuriously 

alarmed. The PO identified the switch as safety-related, and invoked 

10 CFR Part 21. Panalarm provided a certification of meeting the PO 

requirements. 

tested satisfactorily, and test poi,:ts were verified; spurious alarms 

were noted only at the edges of the deadband, which is typical of all 

units. An acceptance test data sheet was also provided, with 

attestation by the QA manager three days after the production 

inspector’s signoff. The NRC inspector noted that Panalarm’s GE test 

This PO called for a failure analysis of a temperature switch 

A factory service repair report stated that the unit 

7 
procedure specified that all setpoints should be set for approximately 

mid-scale trip before shipping. 

The inspector was aware of three LERs, but no 10 CFR Part 21 reports, 

submitted by NMP2 in recent years involving Panalarm temperature switch 

modules. LER 92-001 dated February 3, 1992, addressed a temperature 

switch that generated a spurious high temperature alarm; it stated that 

from April 1987 to February 1992, NMP2 had 76 work requests for the 

repai r/repl acement of Panal arm temperature sui tches, with dri ft in and 

out of the alarm state as a predominant failure mode. 

number for the LER switch was 164C5687P001, whereas the part number for 

the switch returned to Panalarm for evaluation was 164C5687P301; the LER 

did not give a serial number, but evidently the switch addressed in the 

LER was not the one returned to Panalarm. Another NMP2 LER, 92-024, 

reported a temperature switch with a trip relay coil that fa'led open, 

causing a downscale trip output. 

rep1 acement of a1 1 Panal arm temperature switches with another 

manufacturer's products. NMP2 LER 94-04 dated January 17, 1994, 

referred to replacement of at least 65 Riley temperature switches, but 

the LER did not identify the replacement equipment. 

The GE p'art 

Both of these LERs indicated plans for 

Enteray: Entergy Operations, Inc. PO No. MP519933 dated July 19, 1995, 

covered two GG81147001 optical isolators for the Grand Gulf nuclear 

station. 

for original equipment qualified to IEEE 323-74 and 344-75 by Wyle test 

report 4407411, and referenced the original Riley job no. 607-70024; it 

also required electrical tests "as acceptable by Riley Co." The PO 

stated that the equipment performs a safety-related function, invoked 

10 CFR Part 21, and required the seller to comply with his QA program as 

approved by the purchaser. The VRC inspector observed an Entergy letter 

to Panalarm dated February 23, 1994, which stated that based upon review 

of Revision 10 of the Panalarm Qk manual dated December 8, 1993, the 

Panalarm QA program was considered acceptable to Entergy. The letter 

also called for Entergy review of any program revisions prior to 

implementation. 

audit, although he was not required to have any. 

The PO stated that the parts shall be qualified replacements 

The Panalarm QA manager had no evidence of an Entergy 

Panalarm provided a certificate of compliance with PO requirements. 

test report for each isolator showed input and output voltage levels for 

ten channels measured per procedure QC-77-53 dated November 23, 1981. 

A 

Entergy PO No. WP061473 dated November 23, 1994, covered 20 type 

LP92268005 printed circuit boards for the Waterford 3 plant. 

stated that the boards were for use on an annunciator panel originally 

furnished under Ebasco/Riley PO UNY-403540 dated January 30, 1976, 

through Supplement #8 dated July 26, 1979. 

that a 1981 Ebasco audit of Riley found them acceptable for Appendix B 

listing.) The PO required the item manufacturer's certification that 

there were no changes in design, material , manufacturing, or 

interchangeability between January 30, 1976, and the date of manufacture 

of the replacement boards. Alternately, a list of changes could be 

provided for Entergy's approval. 

The PO 

(The NRC inspector found 

The PO invoked the latest version of 

8 
Panalarm’s QA manual, and the QA program which has been approved by 

Entergy’s supplier audit supervisor. 

provisions of 10 CFR Part 21 apply to the item included on the order. 

The PO called for certification of conformance t o a l l PO requirements, 

and required certification of an engineering equivalency evaluation (but 

not submittal of the evaluation itself) for any substitute parts, with 

revised technical manual pages or product information. 

The PO also stated that the 

Panalarm ce: 1 :fied that no chan3es had been made that affected 

qualification to IEEE 323-1974 and IEEE 344-1975. A list of drawing 

revisions was provided, with engineering change notices attached that 

identified several parts changes. 

report, but there was no evidence of testing. 

Panalarm supplied an inspection 

The NRC inspector reviewed an Entergy letter dated February 23, 1994. 

The letter stated that based on review of Revision 10 of Panalarm’s QA 

manual, the QA program was considered acceptable for safety-related POs. 

The letter did not mention any audit, and the Panalarm QA manager had no 

evidence of an Entergy QA audit, or of Entergy surveillance of any 

activity on the PO. The PO did not specifically invoke Appendix 6 or 

state that the equipment was safety-related. 

perform a safety-related procurement, the NRC inspector did not observe 

documentation at Panalarm that supported that objective. 

If Entergy did intend to 

c. Conclusions 

The inspector concluded that the Model 86 modules supplied to GE 

conformed to Appendix B of 10 CFR Part 50. 

The optical isolators supplied to Entergy were nc;t demonstrated to be 

safety-grade by the Entergy and Panal arm documentation reviewed. 

Although Entergy’s PO for printed circuit boards invoked 10 CFR Part 21 

and Panal arm’s QA manual, it was not clear that the PO covered safetyrelated equipment. Absent evidence of an audit or surveillance, there 

was also no basis to support Method 2 or 3 dedication of the boards for 

safety-re1 ated use. 

3.5 Customer audits 

a. ScoDe 

The inspector reviewed reports of audits by GE, Wolf Creek Nuclear 

Operating Corporation, and IES Utilities, and a Commonwealth Edison 

strrvei 11 ance report. 

b. Observations and Fi ndi nas 

GE perfmned an audit of Panalarm ori March 7 and 8, 1996. 

resolution of the audit findings, GE intended to retain Panalarm as an 

approved suppl ier of safety-re1 ated temperature switches and re1 ated 

commercial grade items. The Panalarm QA manager stated that at the time 

of the NRC inspection, both GE corrective action requests remained open; 

Upon 

9 

107 
they involved Panalarm’s failure to conduct triennial internal and 

external audits. All five observations were closed. 

Yolf Creek scheduled an audit of Panalarm on May 23-24, 1996. 

intent of the audit was to extend the Appendix B scope of Panalarm to 

include inverters and isolators, and to address Panalarm’s failure to 

notify of a mandatory inspection hold point on a safety-related PO (this 

finding was closed by a September 3, 1996, letter). While at Panalarm, 

the Wolf Creek representative determined that programmatic refinements 

would be necessary to maintain an Appendix B program, although he 

considered the procedures adequate to support a commerci a1 grade 

program. Panalarm did not plan to revise the QA program or 

documentation pending decisions concerning IS0 and Appendix B 

commitments, so the Wolf Creek representative changed his audit to 

become a surveillance o,f activities on an existing PO. 

not extend the safety-related scope of supply. The NRC inspector 

considered the Wolf Creek representative’s actions to be prudent. 

The 

Wolf Creek did 

IES Utilities, for the Duane Arnold Energy Center (DAEC), conducted an 

audit of Panalarm on June 13-15, 1995, as a triennial followup to a 1992 

audit. The audit had three findings and seven observations, the last of 

which were closed by an IES letter dated February 7, 1996. Based un the 

audit, Panalarm was retained as an approved Appendix B supplier of 

Model 868 switches, including their repair, for the DAEC. The NRC 

inspector noted that the IES audit report and related correspondence 

were distyibuted to several other licensees. Page 1 of the audit report 

states that it was performed to fulfil DAEC requirements, and there was 

no indication that it was a joint utility group audit. The audit report 

also stated that Panalarm documents 900396 Revision 3 and QC-92-110 

Revision 0 would continue to be invoked on DAEC POs, and specified that 

subtier supplier documentation would have to be verified by testing, 

source inspection, and/or audit/survey. 

Commonwealth Edison (ComEd) conducted a survey of Panalarm on April 6, 

1992. The announcement letter stated that CcmEd intended to perform 

Method 2 dedication. CornEd’s May 21, 1992, letter stated that Panalarm 

was listed as an approved manufacturer of commercial grade items for 

Series 10, 50, and 70 annunciators and parts. The Panalarm QA manager 

had no subsequent audit information regarding ComEd. 

The NRC inspector noted that some of the audits by Panalarm customers 

raised concerns with Panalarm’s 10 CfR Part 21 program, yet after the 

concerns were resolved the Panalarm Part 21 procedure was still 

inadequate as described in Section 3.1. 

c. Conclusions 

The inspector concluded that the customer audits of Panal arm were 

effective, apart from the findings .3f this inspection report. 

3.5 Entrance and Exit Meetinqs 

10 

108 
In the entrance meeting on September 9, 1996, the NRC inspector 

discussed the scope of the inspection, outlined the areas to be 

inspected, and established interfaces with Panalarm management. 

exit meeting on September 12, 1996, the inspector discussed his findings 

and concerns. 

In the 

LIST OF PERSONS CONTACTED 

Lawrence M. Froman, Division Vice President and Manager Arl in H. Demerer , Qual i ty Assurance Manager 

David Hinerd, Production Manager 

Lorenzo Garcia, Production QC Inspector 

Lou Granato, QA/QC Technician 

11 

109 
ITEMS OPENED, CLOSED, AND DISCUSSED 

Qpened 

9990 1303/96-01-0 1 VI0 inadequate Part 21 procedure 

99901303/96-01-02 NON inadequate provision for verifying 

99903303/96-01-03 NON inadequate organizational definitions 

9990 1 303 /96- 0 1 -04 

safety-re1 ated activities 

in QA manual 

NON inadequate inspect ion independence 

C1 osed 

Discussed 

None 

12 

110 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. -1 

Oecexnber 31, 1396 

Ms. F.J. Harvey, Acting President 

Energy Systems Business Unit 

Westinghouse El cctri c Corporation 

P.O. Box 355 

Pittsburgh, FA 15230 

SUBJECT: NRC INSPECTION REPORT 99901307/96-01 

Dear Ms. Harvey: 

On October 4, 1996, the U.S. Nuclear Regulatory Commission (NRC) completed 

an inspection of Westinghouse Energy Systems Business Unit (ESBU) facilities 

in Monroeville and Cheswick, Pennsylvania. The enclosed inspection report 

presents the results of that inspection. 

The inspection reviewed the implementation of your qual i ty assurance program 

in supplying safety-related equipment, software, and services to NRC-1 icensed 

facilities, and reviewed your program (and its implementation) established 

pursuant to Part 21, "Reporting Defects and Noncompliance," of Title 10 of the 

Code of Federal Requlations (10 CFR Part 21). 

During this inspection, the NRC inspector found that one of your activities 

appeared to be in violation of NRC requirements. 

procedures adopted pursuant to 10 CFR Part 21 contained minor weaknesses and 

inconsistencies, that, taken in the aggregate, may render them inadequate to 

meet the requirements of §21.21(a) of 10 CFR Part 21. 

NRC Enforcement Policy as promulgated in NUREG-1600, this is considered a 

minor violation and therefore, no Notice of Violation will be issued. 

Although the NRC expects that ESBU will take prompt and appropriate corrective 

action, no response to this item is required. 

Specifically, your system of 

In accordance with the 

The NRC inspector found no instances in which ESBU failed to comply with other 

requirements of 10 CFR Part 21. However, one unresolved item was identified 

involving Corrective Action Reports of the ESBU Process Control Division. The 

question of adequacy of the justification for closeout of the reports and 

hence, the potential existence of deviations or failures to comply requiring 

evaluation in accordance with 10 CFR Part 21 remained unresolved at the end of 

the inspection. 

identified in Section 3.1.b.3 of the enclosed report. 

You are requested to respond to the unresolved item 

111 

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). 
Ms. Harvey -2- 

We appreciate the cooperation of your staff during this inspection. If you 

have any questions, please contact Mr. Stephen Alexander at (301) 415-2995. 

Si ncerel y , 

mb d~ nief 

Special Inspection Branch 

Di vi si on of Inspection and Support Programs 

Office of Nucl ear Reactor Regul at i on 

cc/wencl : 

Mr. Nicholas Liparulo, Manager 

Regulatory and Engineering Networks 

Docket Number: 99901307 
Report No. 

Organization: 

Contact : 

Nuclear Industry 

Activity: 

Dates: 

Inspectors: 

Approved By: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

9990 1307/96-01 

Westinghouse Electric Corporation 

Energy Systems Business Unit 

P.O. Box 355 Pittsburgh, PA 15230 .. 

H.A. Sepp, Manager, 

Regulatory and Licensing Initiatives 

Nuclear Safety-Re1 ated Equipment, Components, 

Rep1 acement Components and Parts, Software, and 

Services for the Commercial Nuclear Power Industry 

September 23 through October 4, 1996 

Stephen D. Alexander, Reactor Engineer 

Ralph R. Landry, Senior Reactor Engineer 

Gregory C. Cwalina, Chief 

Vendor Inspection Section 

Special Inspection Branch 

Division of Inspection and Support Programs 

Enclosure 1 

113 
1 INSPECTION SUMARY 

During this inspection, the NRC inspectors reviewed the implementation of 

selected portions of the quality assurance program of Westinghouse Energy Systems Business Unit (ESBU) for supplying safety-related equipment, software, 

and services to NRC-licensed facilities. 

program (and its implementation) established pursuant to Part 21, "Reporting 

of Defects and Noncompliance," of Title 10 of the Code of federal Reaulations 

(10 CFR Part 21). 

The inspectors also reviewed specific technical concerns with regard to 

certain considerations in the design of some safety analysis computer models 

that may be deficient and that may lead to nonconservative results, that were 

not identified in software error ronnrts. Compliance with 10 CFR 50.46 

reporting requirements were also within the scope of this inspection. 

The review focused on the ESBU 

The inspection bases were: 

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

Fuel Reprocessing Plants," to 10 CFR Part 50 

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

10 CFR Part X, Appendix K, "ECCS Evaluation Models" 

10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems 

for light Water Nuclear Power Reactors" 

The NRC inspectors found that the ESBU system of procedures (including those 

of its major divisions) adopted pursuant to 10 CFR Part 21 contained various 

weaknesses and inconsistencies that, taken in the aggregate, may render them 

inadequate to meet the requirements of 921.21(a). 

Enforcement Pol icy as promulgated in NUREG-1600, this was considered a minor 

violation and no Notice of Violation was issued. 

In accordance with the NRC 

Although the NRC inspector found no conclusive evidence of instances in which 

ESBU failed to comply with other requirements of 10 CFR Part 21, two instances 

were identified involving Corrective Action Reports of the ESBU Process 

Control Division. The question of adequacy of the justification for closeout 

of the reports and hence, the potential for the existence of deviations or 

failures to comply requiring evaluation in accordance with 10 CFR Part 21 

remained unresolved at the end of the inspection. This unresolved item is 

discussed in Paragraph 3.1. b.3 of this report. 

The inspectors found no instances in which ESBU was not in compliance with 

10 CFR 50.46 or 10 CFR Part 21 with regard to its system of tracking changes 

in predicted reactor core parameters (e.g., peak cladding temperature (PCT)) 

under modelled design basis accident conditions and reporting them to affected 

licensees. 

2 

114 
2 STATUS OF PREVIOUS INSPECTION FINDINGS 

No previous findings were reviewed during this inspection. 

3 INSPECTION FINDINGS AMD OTHER CMHENTS 

3.1 10 CFR Part 21 Procrram 

a. Part 21 Program Review Scope 

The inspector reviewed the procedures adopted pursuant to 10 CFR Part 21 by 

ESBU (ESBU-21.0 and its predecessors, ESBU-19.0 and OPR-19.0) and those of 

selected major ESBU divisions (called Level I1 procedures, Level I being the 

ESBU Quality Systems Manual) including the: 

e Nuclear Services Division (NSD), Monroeville, PA 

i Prccess Control Division (PCD), Ohara Township, PA 

Electro-Mechanical Division (EMD) , Cheswick, PA 

e Systems and Major Projects Division (SMPD), Monroeville, PA 

e Commerci a1 Nuclear Fuel s Di vi si on (CNFD) , Col umbi a, SC 

The inspector noted that the Part 21-related procedures of the Advanced 

Technology Business Area, Operating Plant Business Area, and the Westinghouse 

Pensacola, Florida, Division (steam generators and pressurizers) feed into 

ESBU-21 .O as we1 1. 

The inspector also reviewed lower tier (called Level 111) Part 21-related 

procedures for selected departments within the major divisions including those 

of Rep1 cement Component Services (RCS,. of NSD and Nuclear Safety Analysis 

(NSA) Department of SMPD. 

In addition, the inspector reviewed Level I1 and Level I11 procedures used by 

the various ESBU divisions and departments for implementation of Criterion XV, 

"Control of Nonconforming Material ,'I and Criterion XVI, "Corrective Action," 

of 10 CFR Part 50, Appendix By to determine how effective those procedures are 

and how well they are implemented to identify those nonconformances that 

constitute deviations or failures to comply in basic components that have been 

supplied or offered for use at NRC-licensed facilities, thus requiring 

evaluation under Part 21 using the ESBU-21.0 process. 

To assess implementation, the inspector reviewed the records of Part 21 

evaluations (called Potential Issue (PI) evaluations by ESBU-21.0) performed 

by the ESBU Safety Review Comittee for 1991-1995, with emphasis on those that 

did not result in NRC notification. 

Material Review Reports and Corrective Action Reports (CARs) of EMD; Defective 

Item Notices (DINS) , Design Engineering Orders and Design Engineering Order 

Notifications (DEOs and DEONs) , and CARs of PCD; and Software Error Reports of 

Then the inspector reviewed selected 

3 

115 
the Nuclear Safety Analysis (NSA) Department of SMPD. 

reviewed to determine which material nonconformances and computer software 

errors constituted deviations or failures to comply as defined in $21.3, which 

of these were identified as such and which were evaluated (or should have been 

evaluated) in accordance with 10 CFR Part 21. 

These records were 

b. Part 21 Program Review Observations 

b.1 Part 21 Program Procedures 

ESBU has implemented 10 CFR Part 21 requirements, and the related criteria of 

10 CFR Part 50, Appendix B, by a hierarchy of procedures on three levels: 

Level I: The ESBU Quality Systems Manual (QSM) 

Level 11: 

Part 21 and 10 CFR Part 50, Appendix B, Criteria XV and XVI 

ESBU and major division QA implementing procedures for 10 CFR 

Level 111: Departmental level implementing procedures that would aid in 

recognition and identification of conditions adverse to safety as defined in 

ESBU-21.0 

The principal Level I1 procedure for overall ESBU Part 21 compliance is 

ESBU-21 .O, "Identification and Reporting of Conditions Adverse to Safety." 

ESBU 21.0 (as well as its predecessors, ESBU-19.0, and OPR-19.0) is applicable 

to, and is used directly by most ESBU divisions, including PCD, NSD, SMPD, and 

EMD . 

b.2 Part 21 Program Adequacy 

During the review of the Level I1 and 111 procedures mentioned above (except 

for those of CNFD) of ESBU and selected divisions and departments with 

emphasis on the various revisions and versions of ESBU-21.0 back to 1993, the 

inspector determined that Revision 1 of ESBU-22.0, dated September 20, 1996, 

(as well as previous revisions reviewed), could not be fully relied upon to 

ensure that all deviations and failures to comply would be evaluated and 

reported in accordance with $21.21 because of the following weaknesses and 

inconsistencies: 

(1) The Westinghouse-coined combined category of "conditions adverse to 

safety" (CASs), while useful and complete (including both deviations and 

failures to comply as well as departures from Westinghouse specifications or 

requirements) was not used consistently throughout the procedure such that 

failures to comply and some deviations could be excluded from consideration. 

(2) The director or responsible officer designated by the procedure to receive 

reports of defects or failures to comply associated with substantial safety 

hazards per §21.21(a) (3) was not consistent throughout the procedure. 

Section V of ESBU-21.0, under "ESBU Safety Review Committee Chairman 

Responsibilities," required that "recomnendations" [the committees findings] 

be reported to the "ESBU Vice President and General Manager, or his 

representative," a position that no longer existed at the time of the 

4 

116 
inspection. However, the procedure designaLed the ESBU President as the 

officer responsible for reporting of defects or failures to comply associated 

with substantial safety hazards to the NRC pursuant to 521.21(c). 

position of ESBU President was not permanently filled at the time of the 

inspection, although the Chief Operating Officer of the Westinghouse 

Industries and Technology Group was acting as ESBU President. 

President, in a September 9, 1996, memorandum (WIN: 272-4914), had delegated 

the responsibility for receiving 921.21(a)(3) notifications to the General 

Manager of ESBU Operations . 

The 

The Acting ESBU 

Other observations regarding ESBU-21 .O were as follows: 

a 

a 

0 

a 

0 

The description of Part 21 on Page 1 of 13 of ESBU-21.0, in listing 

those things with which a basic component may fail to comply, omitted 

the Atomic Energy Act of 1554, as amended, and licenses of the NRC. 

Among the definitions of Section 11, on page 4 of 13, in Paragraph 2 

under "Basic Component," the list of things with which a basic component 

may fail to comply omitted the Atomic Energy Act of 1954, as amended, 

and rules of the NRC. 

Discovery was not defined as in 921.3, i.e., the completion of 

documentation of a deviation or failure to comply with the potential for 

creating a substantial safety hazard. Rather, the procedure simply 

describes discovery as when a PI is opened. 

60-day evaluation period when a PI is opened as opposed to clearly 

requiring the evaluation to be completed within 60 days of Part 21- 

defined discovery as required by 921.21(a)(l) or an interim report to be 

submitted to the NRC within 60 days of Part 21-defined discovery as 

required by 921.21(a)(2). 

ESBU-21.0 provided the vague requirement to the ESBU Safety Review 

Committee (SRC) to "review referred items re1 ative to the requirements 

of NRC regulations" instead of delineating the specific review 

requirements of Part 21, i.e., to determine if the deviation (CAS) being 

evaluated is a defect (that is, could it (a) create a substantial 

hazard, or (b) lead to exceeding a technical specification safety 

limit); or if the failure to comply (CAS) being evaluated could be 

associated with a substantial safety hazard. 

The procedure starts the 

The requirement for interim reporting on Page 11 of 13, Section VI, 

Paragraph 2(b) under "Notification Guidelines," cites .lo CFR 21.21(b) 

which is incorrect. Interim reports are required by 521.21 (a) (2). 

Appendix A to this report lists other Level I1 and I11 divisional and 

departmental procedures reviewed that were pertinent to the recognition and 

identification of deficiencies as deviations or failures to comply (i .e. , conditions adverse to safety as defined in ESBU-23.0) that should be submitted 

to the ESBU SRC for evaluation/further evaluation per 921.2l(a)(l). Some 

specific comments are included. 

5 

117 
b.3 Part 21 Program Implernentation/Compl iance 

The inspector reviewed selected records of evaluations of documented 

deviations and failures to comply, called "potential issues" (PIS) by ESBU21.0, with emphasis on those that did not result in NRC notification per 

§21.21(c) because the ESBU SRC had determined that the deviations were not 

defects or the failures to comply were not associated with substantial safety 

hazards. 

(which was provided) beyond that in the file to determine its acceptability. 

None were identified that should have been reported to the NRC. 

The inspector reviewed selected records of the various divisions and 

departments within ESBU to assess the effectiveness of their procedures and 

programs in the recognition and identification (among the various types of 

deficiencies, error reports, and nonconformances, that are tracked) of 

deviations and failures to comply as defined by §21.3 that would need to be 

evaluated per the ESBU-21.0 process in compliance with §21.21(a)(l). 

The inspector found only one PI that required additional information 

0 PCD Defective Item Notices (DINs): The inspector identified no DINs 

that should have been recognized as deviations or failures to comply as 

defined in 521.3 (or CASs as defined in ESBU-21.0) in delivered basic 

components, but were not. 

PCD Corrective Action Requests (CARs): The inspector identified two 

CARs, 92-064 and 93-002, with questionable justification for closeout. 

The CARs' dealt with the failure of certain purchased components to meet 

all test requirements. The components had been used in equipment 

supplied by PCD as basic components. PCD's disposition of the CARs was 

that the components were satisfactory as shipped. However, there was 

insufficient evidence in the records available for review for the 

inspector to reach that conclusion. PCD agreed to follow up on the CARs 

in question and provide additional information. Should the 

justification ultimately be deemed inadequate, the potential exists for 

unrecognized (undiscovered) , and hence uneval uated, devi ations or 

failures to comply. This item will require review of additional 

information being developed by the vendor. Designated Unresolved Item 

99901307/96-01-0 1 

0 PCD Development Engineering Order Notices (DEONs) : The inspector 

identified no DEONs that should have been recognized as deviations or 

failures to comply in delivered basic components and were not. 

0 EM0 Material Review Reports (MRRs): The inspector identified no MRRs at 

EMD that should have been recognized as deviations or failures to comply 

in delivered basic components and were not. 

0 NSA Software Error Reports: The NSA Department of SMPD handles errors 

in software codes in accordance with NSA's procedure WP-4.39.3, Revision 

0, August 31, 1996, formerly DP-3.7.4 (Revision 3, January, 31, 1993, 

reviewed) and then WP-3.7.4 (Revision 5, August 14, 1995, reviewed), 

"Software Error Reporting and Resolution." 

safety analysis computer codes are used by certain utilities directly 

In a few cases, some of the 

6 

118 
(e.g. the LOCBART and NOTRUMP loss-of-cool ant-accident (LOCA) codes are 

used by Virginia Power (forme.rly VEPCO)). In these instances, the code 

itself, and not the analysis, is the delivered basic component. In 

accordance with ESBU procedures, (and as evidenced in the records) NSA 

notifies these utilities promptly of the error as soon as an error 

report is opened because they are direct users of the codes. Therefore, 

the user notification meets the requirements of 521.21(b) with regard to 

identified code errors regardless of the results of further analysis of 

the impdct of those errors (and their correction and reanalysis) on 

predicted core accident parameters. 

The inspector found that although it had taken Westinghouse as long as 

13 months (one instance in 1992) to close out an error report, the 

determination of the error/cause, correction of the error (writing new 

code), verification and Val idation, and rerunning a71 affected plant 

analyses using the corrected code to determine the impact of the error 

(i .e., nonconservatisms) quantitatively (or qualitatively) are 

legitimate discovery activities, That is, it is part of the discovery 

process, for which Part 21 has no prescribed time limit, to conduct 

these activities to determine if the existing analyses (which are the 

delivered basic components in most cases, not the codes themselves, fail 

to comply (e.g:, with 10 CFR 50.46, 10 CFR Part 50, Appendix K, or 

technical specifications) or contain a deviation (departure from a 

technical procurement specification). Note that the inspector reviewed 

selected 1 icensee procurement documents (contracts with ESBU for 

analytic services) and found none in which any of the software errors 

reviewed by the inspector could be construed as a deviation from the 

largely non-technically prescriptive language in the contracts. 

Either an actual deviation (i .e. , an analyzed negative impact of a 

software error) or failure to comply, should they have the potential to 

create a substantial safety hazard or contribute to exceeding a safety 

limit, must be evaluated per 10 CFR 21.21(a)(l) or must be reported to 

affected licensees or purchasers per 521.21(b). for example, if an 

error in the NOTRUMP code, used in conjunction with LOCBART, should be 

identified, ESBU would need to determine what effect or impact such an 

error, when corrected, would have on, for example, peak cladding 

temperature (PCT). 

any existing plant nuclear safety analyses, that may have been performed 

using the hypothetically faulty or nonconservative codes or computer 

models, constitute or contain deviations or failures to comply as 

defined in 921.3. 

Only then could ESBU reasonably determine whether 

In addition, the inspector noted a few instances in which the changes in 

PCT resulting from a code error (determined after correcting the error 

and running preliminary analyses to asses the potential error magnitude) 

were large enough for ESBU to open a PI, and evaluate it under 

§21.21(a)(l) per ESBU-21.0 or its predecessors. 

The inspector also found that there have been improvements in the 

interface between the error reporting and resolution procedures and the 

Part 21 procedure. Also, there was an improving trend in closing out 

7 

119 
I 

error reports promptly due to increased management attention and 

followup. The inspector identified no software error reports that 

should have been treated as deviations or failures to comply in 

delivered basic components and were not. 

c. 

The NRC inspector concluded that the ESBU system of procedures (including 

those of its major divisions) adopted pursuant to 10 CFR Part 21, §21.21(a), 

contained various weaknesses and inconsistencies that , taken in the aggregate, 

may render them inadequate to meet the requirements of §21.21(a). That is the 

procedures could not be relied upon to ensure that (1) all deviations and 

failures to comply would be adequately evaluated to identify defects and 

failures to comply associated with substantial safety hazards [§21.2l(a)(l)], 

or that defects or failures to comply associated with substantial safety 

hazards would in all cases be reported within five working days to a person 

who qualifies as a director or responsible officer as defined in 921.3 

[§21.21(a)(3) J. The weaknesses were discussed with the cognizant vendor staff 

during the inspection and summarized at the exit meetings. In accordance with 

the NRC Enforcement Policy as promulgated in NUREG-1600, this was considered a 

minor violation and hence no Notice of Violation was issued. This violation 

notwithstanding, however, the inspector did not identify any instances in 

which 10 CFR Part 21 was not otherwise complied with. 

Part 21 Program Review Conclusions 

The inspector further concl uded that the NSA practices regarding hand1 i ng of 

software error reports met the requirements of 521.21 whether the error is 

reported to users immediately or is determined to be (at least potentially) a 

deviation or failure 'to comply requiring evaluation per §21.21(a)(l). 

3.2 The LOCA Enaineerina Review Council 

a. Scope of LERC Review 

The inspector3 reviewed the function and operation of the LOCA Engineering 

Review Council (LERC), and the consistency in modelling among analysts to 

assure that model changes are applied uniformly. The function of the LERC was 

examined through review of the minutes of LERC meetings and the resolution of 

issues brought before the LERC with followup interviews of cognizant staff. 

The LOCA code guidelines review was to cover the manner in which the codes are 

used and applied to insure that a consistent modelling procedure was used from 

pl ant-to-pl ant. The inspector reviewed the vendor-to-1 i censee reporting 

procedures for model changes, and resulting PCT changes, to ensure that ESBU 

procedures and practices supported 1 icensees in meeting the 10 CFR 50.46 

annual and 30-day reporting requi rement s . 

b. LERC Observat i ons 

In July 1992, NSA formed the LERC to examine perceived errors and proposals 

for application of new methodologies, design techniques, and analysis 

procedures to assure that they are reasonable, appropriate and in conformance 

with NRC regulations. The LERC was made up of senior technical staff members 

with sufficient background and knowledge of the history of code development 

8 

120 
and applications. 

proposed changes and improvements, while a1 so considering 1 icensing, customer 

relations, and regulatory concerns. 

replacement for the quality assurance technical review requirement of 10 CFR 

Part 50, Appendix B. After review of LERC meeting minutes file, the inspector 

examined the following selected issues in detail : 

At the March 27, 1996, meeting, the LERC considered two issues. The first was 

a concern regarding handling the integral fuel burnable absorber (IFBA) rod 

design. A large number of analytical calculations were being performed when 

IFBA rods were to be used. Four possible solutions were discussed to either 

maintain the status quo, or to seek relief from the NRC from the results of 

certain previously submitted analyses. 

load would require NRC approval. 

current procedures, in spite of the large demand on analysts' time. 

The LERC was intended to review the technical aspects of 

The LERC was not intended to be a 

The methods of lessening the analytic 

The decision was to continue with the 

The second issue the LERC considered in the March 27, 1996, meeting was a 

concern regarding the enthalpy transfer methodology used between the SATAN and 

LOCTA computer codes. At the time of this LERC meeting (and also at the time 

of the inspection), ESBU used the node-centered method. The recommendation 

being considered by the LERC was to use a more modern, so-called "donor-cell" 

enthalpy transfer methodology. 

SATAN would result in a reduction in PCT of 50 to 200°F. SATAN, as then 

configured, used only four core nodes. While use of the less restrictive 

enthalpy transfer would lower the PCT, a new core nodalization sensitivity 

with more core nodes would negate any PCT margin gains. 

was decided to be maintained. 

the study was installed in SATAN as an option, with the default being the 

node-centered enthalpy transfer method. 

enthalpy transfer was the documentation provided to the NRC. The pertinent 

ESBU WCAPs (Westinghouse document designation) refer to the node-centered 

methodology, while some unofficial presentation materials discuss donor-cell 

methods. The conclusion was that the information of record was consistent 

with the methodology in standard use. 

Coding the donor-cell method and using it with 

Again, the status quo 

The donor-cell methodology that was coded for 

Related to the concern over the 

The decision to maintain the status quo with regard to fuel rod analysis 

procedures does not raise a safety concern and is at the discretion of the 

vendor. The enthalpy transfer methodology and core nodalization used in the 

current analyses are consistent with those reviewed and approved by the NRC. 

Use of the donor-cell methodology would require review and approval by the 

NRC. At that time, core nodalization would be reconsidered. 

During the August 17, 1996, meeting, the LERC considered the way in which the 

LOCBART code models the fuel assembly grids; specifically, whether or not the 

code was correctly solving the heat conduction/mass balance for conditions of 

insufficient liquid droplets. 

prepared, and one computer run indicated a change in PCT of +20"F for a plant 

in which PCT occurs late in the LOCA sequence. In that case there was more 

than 200°F margin to the 2,200"F limit of 10 CFR 50.46. The LERC concluded 

that this constituted a model improvement rather than a code error since it 

reflected a newer approach to problem solution rather than incorrect 

modelling. 

A new solution technique for the model was 

The LERC recommended that implementation of this solution 

9 

121 
technique be discretionary until further model changes, such as axial offset, 

were implemented as a code update. 

The inspector determined that the LERC disposition of this issue was 

reasonable and justifiable in the case of implementation of solution update 

techniques. 

cognizant ESBU staff that implementation of the updated model would not always 

result in an increase in PCT. In the case of early PCT plants, a reduction in 

PCT could occur. i,, ,he case of some of the plants in which PCT occurs late 

in the LOCA sequence, an increase in PCT of as much as 20°F could occur. No 

cases were found that violated the limits of 10 CFR 50.46. 

It was shown during the inspector's discussions with the 

In the May 26, 1995, meeting, the LERC considered questions regarding the 

issue of safety injection versus rn;17 in the broken loop for 4-loop plants. 

The concern was that the downcomer delivery was dependent on the break 

location, i.e., bottoin, top, or side of the pipe. The discussion resulted in 

a 4-to-3 LERC vote split and a dissenting opinion. The opinions of the LERC 

minority were appropriately addressed, and resulted in a revision of the 

accepted analysis process. 

Ultimately, analyses were performed with injection locations at the top, side, 

bottom, and 45-degree angle. The bottom location was found to be the limiting 

case. The dissenting opinion was withdrawn. 

The analysis approach,was submitted to the NRC for review and approval as part 

of the submittal dealing with condensation effects credit during safety 

injection (the so-called "COSI" submittal). The pertinent WCAPs are: 

0 

WCAP-11767, "COSI SI/Steam Condensation Experiment Analysis," March 

1988. 

WCAP-10054-P, Addendum 2, Revision 1, "Addendum to the Westinghouse 

Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety 

Injection into the Broken Loop and COSI Condensation Model," October 

1995. Use of the methodology was reviewed and approved by the NRC. 

c. Conclusions Regarding the LERC 

The inspector concluded that, for the activities reviewed, the LERC functioned 

well as a sounding board for issues, concerns and questions regarding code 

models and methods of application. In the limited number of issues reviewed 

by the inspector, concerns were described, discussed, and a satisfactory 

sol ut ion or resol uti on adopted. 

3.3 LOCA Code Use Guide1 ines 

Use of the current generation of Westinghouse LOCA codes is defined by the 

company's code guidelines. 

a plant database (IMP) from which a steady state input deck emerges specific 

to the applicable code, such as NOTRUMP, LOCTA, SATAN, plus a template deck. 

Very limited freedom is accorded the analyst in modifying the input stream. 

Specific data sets, such as a pumped flow table, can be altered to permit use 

The analyst uses a preprocessor (SPADES) to access 

10 
of new data, or to examine possible effects of plant changes. the plant input model are reviewed by an independent analyst as part of the 

quality assurance program. 

errors and work-arounds until final approved corrections have been made and 

incorporated. 

and consistency in code use was reasonable and acceptable. 

Changes made in 

The code guidelines contain notification of code 

The inspector determined that the method of ensuring continuity 

3.4 10 CFR 50.46 ReDorting 

a. Scope 

The inspectors reviewed ESBU loss-of-cool ant accident (LOCA) analysis computer code procedures. Licensee reporting requirements for LOCA analyses must be in 

accordance with 10 CFR 50.46, and 10 CFR Part 50, Appendix K. Specifically, 

changes in an analysis, whether due to code model changes, error corrections, 

or plant input changes, resulting in a total change in calculated peak cladding temperature (PCT) of 50°F (using the absolute value of the changes) 

or more must be reported to the NRC within thirty (30) days. resultins in less than a 50°F total calculated PCT change are to be reported annual 1 y . 

Changes 

b. Observat i ons 

ESBU requires the assigned analysts for the various plants for which ESBu 

performs analyses to keep a running tally of changes in PCT, for example, to 

know the margin to 2200°F PCT, to know when the absolute value of PCT changes 

accumulates to >50"F for 30-day reports pursuant to 10 CFR 50.46, and to 

compile all changes for the 50.46 annual reports. 

Since there are approximately forty (40) analysts working on the different 

Westinghouse NSSS equipped plants, the inspector reviewed the manner in which 

changes involving PCT are contmlled and applied. that NSA consistently prepared qual i ty assurance documents and PCT summary sheets far each plant. 

resulted in QA PCT tabulations listing all affected plants and their 

respective results. 

and final PCT for both large-break and small-break LOCAs for each individual 

customer. In addition, NSA reports all changes immediately to any customer 

who uses Westinghouse analysis codes for licensing-basis analyses. 

The inspector determined 

Changes made in analyses affecting more than one plant 

Then NSA prepares reports summarizing PCT, change in PCT, 

The inspector determined that ESBU transmits code changes resulting in PCT 

changes of tens of degrees Fahrenheit to customers promptly in case the 

customer has made changes ESBU is not aware of that would result in a total 

change in PCT of 50°F or more; in which case, the 30-day reporting requirement 

would be imposed. Also, NSA reviews the database used to track PCT margin for 

all affected plants (called the "IMP" database) prior to each use to be sure 

plant changes have been properly applied and that there are no errors in the 

database. 

Finally, NSA prepares a closeout record for resultant PCT changes providing 

root cause for the changes, potential nonconformance, corrective action, and 

conclusions. 

11 

123 
c. Concl us i ons 

The procedures used to insure timely reporting of PCT changes resulting from 

analyses of small and large break LOCAs are sufficient to permit compliance 

with the reporting requirements of 10 CFR 50.46. 

effectively implemented to ensure accuracy and quality assurance in the 

reporting . 

Procedures have been 

PARTIAL LIST OF PERSONS CONTACTED 

Jerry Malley, Engineer, AID, SMPD 

Larry Walker, Manager Aux. Equip Engr., SMPD 

Larry Kamenicky, Manager, ESBU Quality Systems-I 

R.O. Kechner, Manager, ESBU Quality Systems-I1 

Rocco A. Asselta, EMD Principal Engr (QA) 

Gene R. Strussion, Sr. Engr., NSD, RCS 

Norm Mueller, SMPD, Mgr. I&C Tech & Analysis 

Richard B. Miller, NSD, RLI, Fellow Engr. 

Steve Tritch, NSD General Mgr. 

Philip T. McManus, Process Control Division, T.Q. Engr. 

David N. Alsing, ESBU Quality Systems I11 

Curt F. Ciocca, Sr. Engr., SMPD, NSA 

Sumit Ray, CNFD-CE 

John S. Galembush, NSD, REN, Senior Engr. 

A1 Casadei, CNFD-CE 

Meena Mutyala, Director, ESBU Quality Systems 

David K. Allison, CNFD, Dev. Prog. 

Aldo R. Govi, ESBU/QS, Supervising Engineer 

Nick Liparulo, Manager, REN, NSD 

H.A. Sepp, Manager, RLI, REN, NSD 

William McElroy, Qual & Reliability Engr., PCD 

Frank Rizzi, QA Eng., RCS, NSD 

Tom Cornale, Manager of Quality & Safety Engineering, EMD 

Tim Dunn, Principle Engineer, RCP Engineering 

Dan Garner, NSD, NA 

Mark Kachmar, NSD, NA 

Steve Rupprecht, NSD, NA 

B.J. Metro, Sr. Engineer, SMPD/SE/ICAT 

R.3. Sero, General Manager, SMPD 

ITEMS OPENED, CLOSED, AND DISCUSSED 

pi scussed: 

99901307/96-01-01 URI Questionable Justification for Closeout of PCD 

CARS 92-064 and 93-002 

124 

12 
APPENDIX A 

ESBU Level I1 QA Procedures of Interest Applicable to NSD, SMPD, ABTA: 

Implementing Criterion XV: 

ESBU-13.1, "Field Deviation Report (FDR)," Revision 1, August 31, 1996 

WP-13.2, "Control of Nonconformances, It Revision 0, August 31, 1996 

WP-13.3. "Deviation Notices," Revision 0, August 31, 1996 

Implementing Criterion XVI: 

ESBU-14.1, "ESBU-Significant Quality Issues," Revision 3, September 20, 1996 

ESBU-14.2, "Corrective and Preventive Action," Revision 1, September 20, 1996 

PCD Level I1 QA Procedures of Interest: 

Implementing Criterion XV: 

DP 13-001, "Control of Nonconformances," Revision 0, April 1, 1996, 0 Duplicates, but does not refer to DP 13-003. 

Does not refer to 13-002 for a DIN tag. 0 General : "Reference and foll ow ESBU-19.0. It 

0 In addition, Paragraph 2.d states: "Determine if the nonconformance is a 

0 

condition adverse to safety per ESBU procedure entitled "Identification 

and Reporting of Conditions Adverse to Safety. 

DP 13-002, "DIN Tags," Revision 0, April 1, 1996 

DP 13-003, "Nonconforming Material Review and Disposition," Revision 0, April 

1, 1996 

DP 13-004, "Deviation Notices," Revision 0, April 1, 1996 0 2.a., refers deviation notice DN to ESBU SRC per ESBU-19.0 

Implementing Criterion XVI: 

DP 14-001, "Corrective Action RequestsjReports," under No. 4 Review condition 

(subject of CAR) directed the reviewer to determine if the issue represents a 

"potential substantial safety hazard." It then referred the reviewer to ESBU19.0 if the CAR issue is "reportable" instead of referring the reviewer to 

ESBU-22.0 if the CAR issue could be a condition adverse to safety. 

The special PCD QA procedure for implementing Part 21 and Criteria XV and XVI 

for the type of software (mostly for EPROM) applicable to the type of 

equipment produced by PCD was DP 04-105, "Software Error Reporting and 

Resolution." The inspector reviewed the current revision of DP 04-105, 

Revision 0, April 1, 1996. 

The PCD procedure that covers how e;-rors and deficiencies are corrected and 

initiates engineering review and design drawing changes that prompt referral 

13 

125 
to ESBU-21.0 when required is DP 04-004, "DEON." The inspector reviewed the 

current revision of the DEON (Development Engineering Order Notification) 

procedure, Revision 0, dated April 1, 1996. 

EMD Level I1 QA Procedures of Interest: 

Implementing Criterion XV: 

IDP-Ql, Revision 4, March 15, 1996, did not refer to ESBU-21.0 nor did it 

provide for determining if any of identified nonconformances ever have been or 

could be conditions adverse to safety, Although, among the records reviewed 

at EMD, the inspector found none that should have been treated as conditions 

adverse to safety. IDP-Q1 also does not refer to PAI-403 or IDP-Q17 

Implementing Criterion XVI: 

PAI-403, "Significant Quality Problems," Revision 6, March 31, 1994. Items 

are to be processed per IDP-Q2. 

IDP-02, revision dated February 2, 1990, out of date, says process per OPR19.0 

IDP-Q17, Revision 0, August 6, 1996, SQP Log 

NSD Level I11 QA/Part 21 Procedures of Interest: 

The Engineering Techno1 ogy Department was formerly a part of NSD. 

Technology Instruction Manuals ET-A-1 .O, "Glossary" (Revision 1, October 1, 

1993) , and ET-8-1 .O, "Identification, Evaluation and Closeout of Safety 

Concerns" (Revision 1, October 1, 1993), discuss deviations, but not failures 

to comply. The organization(s) to which these two procedures were meant to be 

applicable have long since changed, and other procedures now perform their 

functions, but they remained effective at the time of the inspection. 

The Level 111, departmental procedure in use by the Replacement Component 

Services Department of NSD for QA/Criterion XV implementation at the time of 

the inspection was RCS-415, "Control of Material Deficiency Reports," Revision 

3, June 30, 1994. 

deficiency described in the material deficiency report (MDR) has ever, or 

could ever have, constituted a condition adverse to safety, as defined in 

ESBU-21.0 in a BC delivered to NRC-licensed facility. RCS-415 used the term 

"substantial safety hazard" regarding reporting, but this term effectively 

excludes those deviations from evaluation which are potential defects by 

virtue of potentially causing the exceeding of a technical spec fication 

safety 1 imi t. 

Engineering 

RCS-415 did not provide for determination of whether the 

RCS uses the ESBU Level I1 QA procedures applicable to NSD for 

of Criterion XVI (i.e., ESBU 14.1 and 14.2) 

mpl ementati on 
SMPD Level I11 QA/Part 21 Procedures of Interest: 

NSA's procedure UP-4.19.3, Revision 0, August 31, 1996, formerly DP-3.7.4 

(Revision 3, January, 31, 1993, reviewed) and then WP-3.7.4 (Revision 5, 

August 14, 1995, reviewed), "Software Error Reporting and Resolution.'' This 

procedure did not allow error report closeout to be scheduled more that six 

months hence, but did allow extensions. Without close management attention, 

this has the potential for allowing excessive amounts of time for discovery. 

In cases in which the error could have a significant negative impact on PCT, 

for example, that is, a large PCT penalty relative to existing licensing-basis 

analyses, it could delay the timely opening of a PI and commencement of the 

§21.21 (a) (1) evaluation. 

15 

127 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-0001 

October 16, 1996 

Mr. Howard E. Houserman 

General Manager 

Zetec, Incorporated 

1370 NU Mall Street 

Post Office Box 140 

Jssaquah, WA 98027-0140 

SUBJECT : NRC INSPECTION REPORT 99901037/96-01 

Dear Mr. Houserman: 

On July 18, 1996, the U.S. Nuclear Regulatory Commission (NRC) completed an 

inspection at Zetec, Incorporated (Zetec) . results of that inspection. 

The enclosed report presents the 

The NRC inspection team evaluated tbe measures that you have established for 

control1 ing your Company’s Eddynet@ computer software, which is widely used 

for nondestructive examination, acquisition, and analysis of nuclear power 

plant steam generator tubing, and concluded overall, that Zetec has a 

satisfactorily controlled software development and software Val idation 

program. 

Eddynet software, such as employee involvement, employee communication, 

proactive solicitation and disposition of potential issues from your software 

customers, and an overall sense of ownership and pride noted in your 

ernpl oyees’ demeanor. 

The team observed strengths in Zetec’s control associated with its 

However, some weaknesses were identified, as discussed herein, regarding 

Zetec’s control of its Eddynet software relative to commercial industry 

acceptable guidance. Additionally, the inspectors determined that the 

implementation of one aspect of your quality assurance program failed to meet 

certain NRC requirements contractually imposed on you by your customers. 

Specifically, one of your activities relative to the quality of safety-related 

activities was not delineated in written instructions or procedures as 

required by 10 CFR Part 50, Appendix B. 

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 nonconformance and to follow the instructions specified in the enclosed 

NON when preparing your response. 

This nonconformance is cited in the 

’ Ewt is a registered trwieimrt of Zetec, inc. 

128 
H. E. Houseman -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 (PDR). 

Sincerely, 

original signed by 

Robert M. Gallo, Chief 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Docket No. 99901037 

Enclosures : Inspection Report 99901037/96-01 

cc: Mr. Stephen H. von Fuchs 

Quality Assurance Manager 

Zetec, Incorporated 

Post Office Box 140 

Issaquah , Washington 98027-0140 

129 
NOTICE OF NONCONFORMANCE 

Zetec , Incorporated 

Issaquah , Washington 

Docket No.: 99901037 

Based on the results of an inspection conducted July 15 through 18, 1996, it 

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

the requirements of the U. S. Nuclear Regulatory Comnission. 

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

10 CFR Part 50 requires that activities affecting quality be prescribed by 

documented instructions, procedures, or drawings of a type appropriate to the 

circumstances and shall include appropriate quantitative or qualitative 

acceptance criteria for determining that important activities have been 

satisfactorily accomplished. 

Section 2, "Qual i ty Assurance Program," of Zetec's Qual ity Assurance Program 

Manual , Revision 14, January 1995, states in part, that "every procedure and 

all revisions are reviewed and approved by Quality Assurance ... This review 

ensures that the standards of acceptability for all features and requirements 

are clear. . . . 'I 

Contrary to the above, Zetec did not establish, review or approve procedures 

to control and ensure repeatable results regarding the development of safetyrelated eddy current computer data screening techniques used by data analysts 

for their standard of acceptability. (99901037/96-01-01) 

Please provide a written statement or explanation to the U.S. Nuclear 

Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555- 

0001, with a copy to the Chief, Special Inspection Branch, Division of 

Inspection and Support Programs, Office of Nuclear Reactor Regulation, within 

30 days of the date of the letter transmitting this Notice of Nonconformance. 

This reply should be clearly marked as a "Reply to a Notice of Nonconformance" 

and should include the following for each nonconformance: (1) a description of 

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

description of steps that have been or will be taken to prevent recurrence of 

these items; and (3) the dates your corrective actions and preventive measures 

were or will be completed. 

Dated at Rockville, Maryland, 

this 16th day of October, 1996 

130 Enclosure 1 
Report No: 

Organization: 

Contact : 

Nuclear Industry 

Activity: 

Dates: 

Inspectors : 

Approved by: 

U. S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

9990 1037/96-01 

Zetec, Incorporated 

Issaquah, Washington 

Stephen H. von Fuchs, Q 

(206) 392-5316 

lity A Iran Man 

Manufacturer and supplier of steam generator tubing, 

eddy current inspection instruments, probes, software, 

calibration standards, manipulators, training and 

services for the commercial nuclear power industry. 

July 15-18, 1996 

Joseph 3. Petrosino, Team Leader 

John M. Gallagher, Electrical Engineer 

John K. Ganiere, Electrical Engineer 

Michael J. Morgan, Reactor Engineer 

Phillip Rush, Materials Engineer 

Gregory C. Cwalina, Chief 

Vendor Inspection Section 

Special Inspection Branch 

Division of Inspection and Support Programs 

131 

Enclosure 2 
1 INSPECTION SUWRY 

During this inspectioq, the NRC inspectors (team) reviewed the implementation 

of selected portions of the Zetec, Incorporated (Zetec) quality program, which 

was expressed in the Zetec Quality Program Manual (2-QA, Revision 14, January 

1995), and specified controls to encompass the requirements contained in the 

criteria of 10 CFR Part 50, Appendix B, American National Standards InstituteAmerican Society of Quality Control (ANSI/ASQC) 49001 and International 

Orgacization for Standardization (ISO) Standard-9001. Through a review of NRC 

licensee's purchase orders (POs) and discussions with Zetec staff, the team 

found that NRC licensees impose Appendix 6, "Quality Assurance Criteria for 

Nuclear Power Plants and Fuel Reprocessing Plants," to Part 50 of Title 10 of 

the Code of Fede ral Retmlations (10 CFR Part 50) and other unique nuclear 

requirements on Zetec for its nondestructive examination services (NDE), NDE 

components. 

Conversely, Zetec's Eddynet@' computer software and associated testing probes 

were found classified by Zetec and procured by NRC licensees and NDE testing 

customers as commercial grade items (CGl's), as defined in 921.3 of 10 CFR 

Part 21. 

probes as CGI's, Zetec's quality manual policy stated that its manual defines 

corporate quality policies and requirements and applied to all Zetec products 

and services. 

computer software program, probe design, manufacture, and supply as CGI's, it 

used some elements of its 2-QA quality program for control of these 

components. 

. analyst activities, EC test calibration standards, and other related 

Even though Zetec considers its Eddynet software and associated 

The team found that although Zetec classified its Eddynet 

Therefore, in addition to the below inspection bases, the team also considered 

industry acceptable guidance when it evaluated Zetec's program controls such 

as, the Institute of Electrical and Electronics Engineers, Incorporated (IEEE) 

Standard 7-4.3.2-1993, "IEEE Standard Criteria for Digital Computers in Safety 

Systems of Nuclear Power Generating Stations," as endorsed by NRC Regulatory 

Guide 1.152, "Criteria for Digital Computers in Safety Systems of Nuclear 

Power Plants," and Electric Power Research Institute (EPRI) NP-6201, "PWR 

Steam Generator Examination Guide1 ines." The inspection bases were: 

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

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

Reaulations (10 CFR Part 50) 

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

Overall, the NRC inspection team found that the measures for your Eddynet 

software development and control were generally sound, with the exception of a 

few weaknesses that are discussed herein. The team observed strengths in 

Zetec's program, such as employee involvement, employee communication, and 

solicitation and disposition of potential fssues from your software customers. 

* 

Eddynet io a registered tradenrerk of Zetec, Inc. 

2 

132 
The team also observed that your employees demonstrated an overall sense of 

ownership and pride in their jobs. The team also believed that the strengths 

noted contributed to the soundness of Zetec's program and the adequacy of the 

eddy current (EC) software. 

During this inspection, one instance where Zetec, Incorporated failed to 

conform to NRC requirements imposed upon them by NRC licensees was identified. 

This nonconformance is discussed in Section 3.7 of this report. 

2 STATUS OF PREVIOJS INSPECTION FINDINGS 

2.1 Violation 99901 037/85-01-01 (CLO SED1 

NRC Inspection Report 99901037/85-01 stated that, contrary to Section 21.6 of 

10 CFR Part 21, copies of Section 206 of the Energy Reorganization Act and 

Procedure ZAG-16 were not posted at the Zetec facility. 

The team looked at Zetec's current posting and reviewed its content regarding 

the regulations. 

requirements of 10 CFR Part 21. 

The team noted that the posted documents addressed the 

2.2 Violation 99901037/85-01-02 tCLOSED1 

NRC Inspection Report 99901037/85-01 stated that, contrary to Section 21.21 of 

10 CFR Part 21, Zetec did not communicate the requirements of 10 CFR Part 21 

to Westinghouse Specialty Metals Divi.;ion, which supplied Inconel tubing to 

Zetec in August 1982. 

This issue concerned a single piece of Inconel tubing which had been 

fabricated into a calibration standard by Zetec for one licensee's steam 

generator tubing EC inspection. This matter was adequately resolved and 

corrective action was taken in 1985 when the finding was identified, 

2 3 

Nonconformances 85-01-03 through 85-01-13 are considered closed because 

adequate corrective action was taken and the applicable licensees were made 

aware of the issues when the inspection report was issued. Additionally, in a 

letter dated June 9, 1986, regarding Inspection Report 99901037/85-01, the NRC 

staff stated that it had reviewed Zetec's replies and found them responsive to 

the concerns raised in the 1985 notice of nonconformance. 

Noncon f ormances 99901037/85-0 1-03 t hrouqh 85-0 1 - 13 (CLOSED 1 

3 INSPECTION FINDINGS AND OTHER COMMENTS 

3.1 10 CFR Part 21 Proqram 

The team reviewed Procedure ZAG-16, "Reporting of Safety Hazards," which Zetec 

adopted pursuant to 10 CFR Part 21. 

procedure generally met the intent of fart 21, it lacited some clarity and 

specificity in certain areas. For exmple, the procedure was not very clear 

about the vendor's responsibilities under §21.21(b) of Part 21 and it did not 

define some key words, such as eva'luation and discovery. The team discussed 

The team noted that although the 

3 

133 
potential changes to ZAG-16 with the QA Manager, which would appropriately 

clarify Zetec's procedure. 

modifying the procedure as necessary within approximately 90 days from the 

receipt of this report. 

The QA Manager committed to reviewing and 

No other concerns were noted in this area. 

3.2 Quality Assurance Proaram 

The inspection team selectively reviewed Zetec's Quality Manual Z-QA, 

Revision 14, dated Januarj 1995, and found that it addressed 10 CFR Part 50, 

Appendix 8, IS0 9001, and ANSI/ASQC 49001. 

sections which correlated to the different requirements and regulations. 

team noted that document 2-QA stated that: "The scope of the quality assurance 

program plan includes the total operation of Zetec, Incorporated. The 

measures are to be implemented to the extent necessary to assure that 

services, equipment and other items supplied to or performed for customers by 

Zetec shall conform to the specified quality levels of applicable codes, 

standards, regulatory criteria and purchase order specifications.. . .I' 

Further, the team noted that the cover page of Zetec's Quality Manual stated, 

"This manual defines Corporate Quality Policies and Requirements and applies 

to all Zetec products and services." 

Z-QA was divided into separate 

The 

The team reviewed a sample of the licensee's purchase orders (POs) and 

verified that Appendix B of 10 CFR Part 50 and other unique nuclear 

requirements had been imposed on Zetec for its nondestructive examination 

(NDE) services, EC test calibration standards, and related components. The 

team also found that Zetec's probe products and Eddynet software were 

typically procured by NRC licensees and NDE testing customers as commercially 

available products; consequently, Zetec's quality program would encompass 

Zetec's control of Eddynet and associated probes under its IS0 9001 aspects. 

From its review of Quality Manual 2-QA and observations of Zetec's operation, 

the team concluded that the implementation of Zetec's program and 

establishment of its manual appeared to be adequate. 

3.3 

The team evaluated Zetec's software development validation and configuration 

control process to determine whether Zetec was achieving and maintaining a 

consistent level of quality in its Eddynet software products. The Eddynet 

software is used in the NDE eddy current inspection of safety-related steam 

generator tubing and plays a critical role in the generation of the eddy 

current inspection data. 

Eddynet@ Software DeveloDment. Val idation and Configuration Control 

Zetec has a single manager, the Software Development Supervisor, in charge of 

the entire software development program. This person is responsible for all 

aspects of the development and configuration control of the Eddynet software. 

The Software Development Group consists of six programmers. Three of the 

programers have had extensive field experience in performing EC acquisition 

and analysis, and four of the six are considered to be experienced computer 

language C programmers. An associated group, the Software Support Group, a1 so 

reports to the Software Development Supervisor and is responsible for 

developing the software validation check1 ists. 

4 

134 
The Eddynet 95 software product is written in computer language C for 

application in either of two different Hewlett Packard workstations: Series 

300/400 (HP-UX Version 9.03 operating system); or Series 700 (HP-UX Version 

9.05 operating system). Eddynet 95 comprises several independent programs , such as acquisition programs, analysis programs, and inspection management 

programs. The team noted that the Eddynet series and associated firmware are 

offered as commercial products. Although ZeLec is registered as an IS0 9001 

supplier, the team was told by Zetec staff that its software development does 

not conform to the guidelines of IS0 9000-3, "Guidelines for the Application 

of IS0 9001 For the Development, Supply and Maintenance of Software." 

3.3.1 Software Requirement Specification 

The Software Development Supervisor told the team that Zetec had not 

established any formal software requirement specification for the performance 

expectations of the Eddynet software. The original Eddynet software was 

written in an effort to develop a digital acquisition system and consisted of 

relatively straightforward algorithms, As a result, the original development 

process did not follow a formal software development life-cycle approach which 

would encompass a software requirement specification. 

with the Zetec software staff, the team understood that since it first 

developed its Eddynet software, Zetec has continued to enhance Eddynet 

software with strong emphasis on the man-machine interface characteristics of 

its software program. 

ensuring that the software product is validated before release. 

From its discussion 

As a result, Zetec had focused much of its attention on 

The team discussed with other NRC staff its concern that no documented 

software requirement specification existed. Upon further consideration of 

Zetec's utilization of its available resources for the development and 

validation of the software, the team concluded that the lack of a detailed 

specification was more a question of development efficiency than quality. 

The team observed that the majority of software changes which had been 

incorporated into new versions of Eddynet 95 were related to man-machine 

interface enhancements. 

Change Request (SCR) Program which is descrjbed in Section 3.5 of this report. 

No other concerns were noted in this area. 

These changes were derived from Zetec's Software 

3.3.2 Software Verification and' Val idation 

The team reviewed Zetec's software verification and validation process. 

team reviewed the methodology that Zetec had established to validate the 

Eddynet software program after Zetec made changes to the program. 

following three Zetec work instructions were reviewed to evaluate their 

adequacy and appropriateness: 

The 

The 

e 

e 

e 

New Version Product Disk Validation 

Version Update Disk Validation 

frototype Disk Validation 

5 

135 
The team concluded that the major purpose of these work instructions was to 

ensure that new version releases, version update releases, and prototype 

releases of the Eddynet software product performed correctly. Additionally, 

the team found that these procedures appeared to adequately control the 

applicable portions of the Zetec program for which they were written and were 

appropriate to the circumstances. 

The team found that the Software Support Group had developed a software 

checkout book that contained Eddynet validation checklists used by the Zetec 

group responsible for the Val idation process. The software checkou,t book 

consists of multiple checkout groups and subgroups which contain specific 

validation checklists for each of the programs within Eddynet. The Software 

Support Group develops and maintains these Val idation checkl ists. A version 

checkout coordinator from the Saftware Support Group was designated as being 

responsible for the overall coordination of the validation process. The 

validation is performed by employees within Zetec’s Product Support Group and, 

to a limited extent, by the Software Support Group. Zetec indicated that its 

validator personnel had extensive field experience with Eddynet. 

The team concluded that there was sufficient independence between the software 

devel opers , the Software Support Group who devel op the Val i dat i on checkl i st s , and the Zetec group that performs the validation process. 

The software group stated that the intent of the validation process was to 

exercise all possible program functions and, therefore, verify the 

functionality of each of the programs in the software. 

functionality problems encountered by Zetec during its validation process were 

recorded on version validation sheets. The problems were then compiled by the 

version checkout coordinator and were reviewed and handled by the Software 

Development Supervisor. 

had been made, Zetec subjected its software to a second complete validation 

process checkout. That is, Zetec repeats all of the actions that were done 

for the first validation process. 

The team found that 

The team observed that after the appropriate changes 

On the basis of conversations with the Software Support Group, the team 

concluded that the checkout process is continually revised to account for new 

features added to the Eddynet program, and the subgroup checklists are updated 

to ensure that all new features of the software are exercised. In addition, 

based on procedural instructions and management instructions, it is the 

understanding of the individual validators that they are responsible to 

exercise all functions within the program, even if each function is not 

specified in the subgroup checkl ist. 

A1 though the team found that the current Val idators are adequately performing 

their job functions, Zetec had not established documented controls or release 

signature requirements for ensuring that the Val idation checkout subgroup 

lists are fully updated to include functional tests for all new features or 

changes associated with a new version or v?rsion update before sending the 

checkout subgroups to the validation team. As a result, there is a potential 

that new features of the software may not be functionally tested before they 

are delivered to customers for safety-related applications. 

classified this as a weakness in the validation process. 

The inspectors 

6 

136 
The team noted that, in addition to the validation of the functional 

performance of the program, the validation process also served as feedback 

from the validators about the acceptability of new features. The program 

descriptions associated with software changes are not considered complete 

until the validation has been finished. 

evaluate and assess the man-machine interface characteristics and 

functional i ty associated with each software change. Overall, Zetec's method 

of performing validation appeared to be effective because of the added 

strength which comes from using personnel who have Eddynet field experience. 

The experience aspect dlso compensates somewhat for Zetec's lack of a 

documented software requirement specification, as discussed in Section 3.3.1 

of this report. No other concerns were noted in this area. 

This allows the validation team to 

3.3.3 Review of Eddynet* 95, Version 2.0, Validation 

The team reviewed the checkout group and subgroups completed for the 

validation of Eddynet 95, Version 2.0, HP-UX Version 9.03 (300/400 Series). 

The team observed that all items within the subgroups appeared to be checked. 

The team a1 so noted that the employees performing the Val idat i on often changed 

terminology in the subgroup checklists and added comments within the margins 

of the checklists. In addition, checklist instructions were modified if the 

information in the instruction was different from the information regarding 

the program functionality. 

instruction that was changed because it incorrectly referred to a "button" 

rather than to a "pull-down menu." 

were not recorded on the validation sheets since they did not affect the 

functionality of the software. 

For instance, the team found one check1 ist 

In most cases, these changes or comments 

Further review of this area revealed that Zetec had not established any 

documented controls for ensuring that the written comments or changes made in 

the validation checklists are reviewed and handled by either a version 

checkout coordinator or by the Software Development Supervisor. 

during discussions with Zetec staff, the team also identified this as a 

weakness because of the potentially inconsistent results that could occur 

during the Val idation process. 

Therefore, 

The team also reviewed the master version validation sheet and verified that 

all problems encountered during the validation were reviewed and handled by 

the Software Development Supervisor. 

validation of Eddynet 95, Version 2.0, was performed in accordance with the 

work instruction "New Version Product Disk Validation," and appeared to be 

effectively controlled. 

In general, the team concluded that the 

No other concerns were identified in this area. 

3.3.4 Review of Eddynet* 95, Version 3.0, Validation 

The team witnessed Zetec's Val idation activities associated with the 

inspection planning program of Eddynet 95, Version 3.0. 

analyst in the Product Support Group W~S performinq the validation in 

accordance with the appropriate subgroup checklist. She data analyst 

confirmed to the team that the intent of the validation was to exercise all 

program functions to ensure that the software performs as designed. 

achieve this, validation personnel executed the steps written in the subgroup 

A data manager/data 

To 

7 

137 
checklist. 

validators generally exercise more functions than are prescribed in the 

checkout list in an effort to verify the functionality of all aspects of the 

program and to check for the potential impact of incorrect key manipulations. 

The data analyst told the team that all new functions are not always 

incorporated into the checkout list they receive. As a result, the analysts 

may add supplemental checks to test new functions at their discretion. 

analyst who encounters a new function that is not on the checklist may contact 

the Software Support Group for an explanation of the new function. As 

described in Section 3.3.2 of this report, the team concluded that Zetec 

appeared to lack adequate controls to ensure that all new features of the 

software were incorporated into the validation subgroup check1 ist; however, it 

was viewed as a strength that the analysts may add supplemental checks to test 

new functions . 

The team found through its discussion and review that the 

An 

The team determined that new versions of the Eddynet 95 software are validated 

on both Series 300/400 and Series 700 Hewlett Packard workstations. 

witnessing the validation of Version 3.0, the team noted that the validation 

on the workstations was performed essentially in parallel by separate 

validators. The team believed that this parallel validation process is a 

valuable asset to Zetec’s process control and allowed for crosschecks and 

real-time on-line discussions between validators regarding problems associated 

with the software or with the validation subgroup checklists. This could also 

be considered as a diverse implementation of the software programs and, as 

such, an additional means for finding software errors. 

In 

3.3.5 Impact of Source Code Change 

The data analyst and Software Development Group indicated that changes to the 

source code that do not affect the high-level functionality of the program or 

cause a visible impact on the operator’s workstation are transparent to the 

validator. As such, the team concluded that the independent Val idation 

process is not structured to assure that software errors which are transparent 

to the user are detected. Specifically, there is no formalized independent 

check of the source code that allows for detection of these types of 

programming errors. 

This practice is not in accordance with acceptable nuclear and software 

development community-accepted quality assurance (PA) procedural guidance and 

practices that govern the development and validation of software, and would be 

a considered as a noncompliance if Appendix B to 10 CFR Part 50 were 

applicable. The inspectors concluded that Zetec lacked criteria to allow for 

the determination of source code changes that require an independent check, 

and a process for performing the independent check of the source code. 

3.3.6 Eddynete Software Configuration Control 

The team reviewed Zetec’s processes for controlling and reporting changes to 

software, including a review of procedures for revising Eddynet software 

programs. 

change notice (SCN) system. 

Zetec indicated that software changes were controlled by a software 

Each software change was documented in an SCN 

8 

138 
form that described the changes, the proposed activities, and the personnel 

responsibilities for changing the software. The team reviewed QA Procedure 

QAP-8, "Control of Production Software," Revision 11. This procedure 

controlled the revision, Val idation, and release of production software such 

as Eddynet 95, and gave specific details about SCN processing. 

The team also reviewed Zetec's SCN log and dssociated SCNs. 

concluded that the software revisions appeared to be processed in accordance 

with QAP-8. 

specific Eddynet version. 

software revisions and noted they were validated before they were released. 

The team 

Each SCN was uniquely identified and could be traced to a 

The team also reviewed a sample population of 

Zetec staff stated to the team that the software changes associated with a new 

version of Eddynet are typically described in a Zetec "announcement bullet in" 

that is sent to customers when new versions are released. The team reviewed 

the release announcement for Eddynet 95, Version 2.0, and confirmed that the 

information in it pertained to significant software changes or enhancements. 

The team concluded that Zetec maintains a good program for tracking and 

controlling software changes and for notifying customers of the changes and 

enhancements associated with new versions of Eddynet. 

3.3.7 Development, Val idation, and Configuration Control 

As described above, the team reviewed Zetec's processes for software 

development , software Val idation, and software configuration control. The 

team concluded overall that Zetec has an effectively implemented and 

control led software development and software val idat i on program; however, some 

weaknesses were apparent. 

in detail with Zetec's staff: 

The team discussed the following two observations 

The team noted that high-qual i ty man-machine interface characteristics 

with respect to effective use of software products in field environment 

is a principal attribute of the Eddynet software. The development and 

Val idation processes support this attribute because they are performed 

by Software Development and Product Support personnel who have extensive 

field experience with Eddynet. 

experience requirements for programmers and validation personnel are not 

control 1 ed by QA procedures. 

However, the team also noted that field 

The team concluded that there are no formal software requirement 

specifications for software changes. 

on real-time on-line discussions in the performance of both software 

development and validation job tasks to maintain a high level of 

agreement during the development and Val idation activities. 

believed that this can result in undocumented bases for important 

decisions regarding the final software product. 

Instead, there i s a heavy re1 i ance 

The team 

3.4 Identifvinta and ReDortinq Software Pzficiencies 

139 

The team asked Zetec about how it addressed program errors in software that 

could potentially affect the quality of EC data being manipulated or analyzed. 

Zetec indicated that significant software deficiencies attract immediate 

9 
attention as they are processed in accordance with QA Procedure QAP-9, 

“Software Change Request,” Revision 0, and are reported to customers. Zetec 

provided three examples of how it notified customers about software problems 

or limitations. The team concluded that Zetec appeared to have addressed the 

problems promptly and efficiently. Although 10 CFR Part 21 is not applicable 

to Zetec’s Eddynet software, the Zetec process appeared to meet the intent of 

921.21(b) of 10 CFR Part 21. 

3.5 Software Chanse Reauests 

The team reviewed QA Procedure QAP-9, “Software Change Request.” This 

procedure defines the process for documenting, reviewing, and implementing 

SCRs. 

personnel to report problems associated with Eddynet or to make 

recommendations for software enhancements. 

Zetec developed the SCR form for the use of customers and Zetec field 

The Software Development Supervisor reviews all SCRs and is responsible for 

deciding if an SCR will be approved (that is, if action will be taken to 

address the customer’s request) and for deciding the schedule for implementing 

the change or enhancement. 

this decision may require speaking with the originator of the request to get 

additional information or to modify the request, (to reduce the impact on 

other parts of the program). 

The Software Development Supervisor noted that 

Upon approving an SCR, the Software Development Supervisor assigns the task to 

a programmer. The Software Development Supervisor discusses the details of 

the change with the assigned programmer and, to implement the software change, 

relies on the programmer’s experience and judgment. Preceding the validation 

phase, the Software Development Supervisor reviews the new or modified code 

developed by the programmer to ensure that it meets its intended function. 

The team noted that the majority of the programmers and validators appear to 

have strong field experience with Eddynet. 

software requirements specification for software changes, formal discussions 

and reviews take place throughout the change process to ensure that software 

changes are handled appropriately. 

validated on two separate operating systems. 

Although there is no formal 

In addition, the software changes are 

The team reviewed Zetec’s process for handling SCRs by tracking several SCRs 

that had been initiated by Framatome Technologies, Inc. Specifically, the 

team reviewed Zetec’s SCR log to ensure that the SCRs were reviewed upon 

receipt, assigned an SCR number, and appropriately dispositioned. The team 

also confirmed that SCNs were implemented as appropriate and that the 

corresponding software changes were incorporated into new versions of Eddynet. 

In general, the’team concluded that Zetec reviewed and addressed the SCRs 

promptly and in accordance with QAP-9. 

Overall, the team concluded that the SCR process appears to be effective in 

identifying man-machine interface problems and customer requests for software 

enhancements. 

most of these SCRs request software enhancements rather than report software 

Zetec receives a large number of SCRs. The team confirmed that 

10 

140 
deficiencies. 

have a good program for tracking and addressing them promptly and 

appropriately. 

Despite the large numbers of SCRs it receives, Zetec appears to 

3.6 Data Ana 1 vsi s Procedures and Acaui s ition Tech niaues 

The team reviewed the vendor's EC data analysis, "Eddy Current Data Analysis 

Procedure," FSP 301-EVALY Revision 7, July 7, 1994, and an acquisition 

procedure, "Eddynet Eddy Current Acquisition Procedure, " FSP-301-EN, Revision 

3, October 20, 1994. The team evaluated Zetec's analysis procedures. and site 

specific procedures to determine to what extent industry data analysis 

practices discussed in such industry documents as EPRI NP-6201 had been 

incorporated into Zetec's program. 

addressed general personnel , documentation, equipment , and procedural 

requirements for the analysis of EC inspection data. 

Revision 3, addressed required actions needed to set up some of the hardware 

and software for acquiring EC inspection data. 

The team found that Procedure FSP 301-EVAL 

Procedure FSP 301-EN, 

The team found that the guidelines were clearly written. 

analysis guidelines contained such strengths as specific requirements on lead 

analyst responsibilities, analyst independence, and noise and rejectable data 

criteria. By interviewing NDE technicians, the team confirmed that Zetec 

analysts were familiar with the details of the procedure. 

In addition, the 

The team found that site-specific guidel ines for examining steam generator 

tubing are typically developed and re /'sed through discussions between the 

utility and the NDE vendors involved in the steam generator tube inspections. 

The team noted that a typical site-specific examination guideline did not 

contain all of the attributes found in FSP 301-EVAL. Zetec indicated that the 

NDE technicians generally focused on revising the finer details of the 

proposed site-specific guidelines rather than on the broader aspects of the 

procedure. 

The team concluded that Zetec's EC analysis and acquisition procedures 

contained sufficient detai 1 ed instruct ions for conducting inspection-re1 ated 

activities. 

however, these strengths were not always being considered by NDE technicians 

for incorporation into si te-specific examination guidel ines. 

The team noted several strengths in the analysis guidel ines; 

3.7 ComDuter-Assisted Data Analysis 

The team reviewed the vendor's controls for computer data screening (CDS) 

techniques used in the analysis of EC inspection data. The review focused on 

assessing whether Zetec had been adequately managing its activities involving 

the use of CDS. The team found that Zetec is often contracted to supply EC 

data analysis using a CDS algorithm. These computer algorithms use specific 

criteria which are inputted by a Zetec data analyst before a particular 

utility customer S/G tube EC inspectic? to identify erpected EC indications. 

The individual responsibility is significant for tne data analyst because they 

must develop these criteria based on expected modes of tube degradation for 

each different utility S/G tube EC examination. 

Zetec did not have a written or formal type of procedure for controlling or 

However, the team noted that 

11 

14 1 
establishing the specific process that is to be used by its data analysts for 

developing these criteria. As a result, the potential exists that screening 

criteria developed by one analyst may be significantly different from that of 

another analyst. To establish the parameters for identifying indications in 

the EC data, a data analyst must independently develop the screening logic. 

Before a Zetec analyst starts EC inspection at a particular nuclear power 

plant site, the CDS is typically subjected to a site-specific examination, 

which it must pass in order to be used in an inspection. 

The team’s review of this activity identified that when developing CDS 

criteria, analysts will typically address known degradation modes. However, 

because of the different experience levels of the analysts, a variance in the 

effectiveness of the developed criteria can occur. This can allow types of 

degradation previously undetected by a particular analyst to be missed by the 

CDS algorithm. The team identified to the Zetec staff types of parameters, or 

ranges, that may need to be delineated within procedures or instructions to 

ensure compliance with the requirements, and to ensure that the widest 

spectrum of degradation modes will be identified. 

required by Zetec’s program. 

These aspects were not 

The team also recoqnized that Zetec’s data analysts have a broad knowledge of 

sms 

ysts 

degradation mechanisms and may adapt well to any new degradat 

that appear during inspettions. The team noted that although 

have not been working with written procedures or instructions 

written guidance and instruction for the development of these 

was using only experienced analysts for developing its CDS cr 

on mechan 

Zetec ana 

regard i ng 

criteria, 

teria. 

Zetec 

The team determined that the potential existed that the lack of such 

procedures could lead to differing performance capabilities for CDS techniques 

under actual inspection conditions and as a result, Zetec’s CDS program could 

fail to identify potentially defective tubes. Therefore, although Section 

2.5, “Procedures,” of Zetec’s Z-QA manual indicated that every procedure and 

all revisions are reviewed and approved by quality assurance to ensure that 

standards of acceptability for all requirements and features are clear, no 

requirements for CDS criteria development were found. 

lack of written guidance to govern the development of CDS criteria as 

Nonconformance 99901037/96-01-01. 

The team identified the 

3.8 Eddy Current Probe Fabrication 

The team observed the EC probe manufacture and interviewed assembly-1 ine 

personnel involved with the fabrication of the EC inspection probes. The team 

reviewed several of Zetec’s procedures used to manufacture and inspect the EC 

probes. For example, Zetec’s “Fabrication Procedure for Motorized Rotating 3 

Coil Probe Head,” Revision 0, dated March 13, 1996, specified the sequence for 

constructing a certain type of EC probe, and the procedure clearly defined the 

steps and essential variables of the fabrication process and required periodic 

simp1 c checks to verify coi 1 /probe perforrance. 

The team also observed that the assembly-line personnel involved in coil 

construction were knowledgeable about the use of the various testers used to 

verify the performance of coi 1 s. The team reviewed an acceptance-re1 ated 

12 

142 
procedure, QAP-21, "ULC Check-Out Procedure," Revision 2, and noted that the 

procedure contained steps for visual inspections (including physical 

measurements of probe dimensions), a physical inspection of the overall 

integrity of the probe, and a functional inspection requirement to verify the 

electrical response of the probe. 

personnel used consistent reference standards for verifying probe performance 

and appeared knowledgeable in their areas of responsibility. 

3.9 EC Personnel Trainina. Oual ifications. and ResDonsibilities 

The team reviewed Zetec's Procedure QAP-101, "Personnel Qualification & 

Certification," Revision 4, and discussed the responsibilities of Zetec's 

Field Services (FC) Group with Zetec's Supervisor of Product Support and 

Technical Development. The team assessed training documentation and reviewed 

certification examinations and discussed the specifics of Zetec's 

certification and qualification program with Zetec's Senior Training Manager. 

The team noted that Zetec EC personnel are qualified and certified to Zetec's 

written Procedure QAP-101 and that Procedure QAP-101 addressed the 

recommendations contained in ASNT SNT-TC-lA, "Personnel QualificationCertification in Nondestructive Testing," 1984. The team also reviewed 

various Level I, 11, and I11 certification examinations and noted that the 

examinations appeared to be thorough and we1 1 -designed, and adequately test 

knowledge of basic theory and EC application. The team noted that Zetec's 

certification and qualification records appeared to generally conform to the 

requirements of ASME Code Section XI (1989) and met the intent of 

Criterion IX, of 10 CFR Part 50, Appendix B. 

The team observed that Zetec's inspection 

The team found that Zetec's analyst training program experience requirements 

may vary because they are dependent upon the currently available job 

assignment and the collective amount of specific site-testing and OJT 

available for the recommended nine-month hands-on period. 

current practice could have the potential to limit job variety, exposure to 

various steam generator designs, and differences in testing practices. The 

team also noted that Zetec's program lacked specificity relative to 

del ineating types of experience that would be required for individual 

qualification. 

Therefore, Zetec's 

During the team's review of this area, it also noted that Zetec's supplemental 

Level I steam generator tube inspection training requirements, that is, 

training performed after Level I certification for on-site, on-the-jobtraining (OJT), were typically followed by routine assignment to Zetec's Field 

Services Group. 

showed that some candidates accumulated more than 1,000 hours of site-specific 

trainind, while other records only reflected 400-500 hours of similar 

training. Consequently, Zetec's program could allow a 500-600 hour disparity 

between different Level I1 NDE examiner training hours. The team noted, in 

the records which it reviewed, that an individual's particular level of 

involvement with a specific team and unique job preparation, such as in-house 

mock-up work, EC equipment setups, and simulated data acquisition run-through, 

was typically the primary contributor to an individual's specific hours used 

for their experience and qualification. 

Individual training records that were reviewed in this area 

13 

143 
I A1 though a concern regarding analyst experience requirements was noted, the 

team concluded that Zetec's current cornmi tment to the ASNT recommendations and 

its current program meets the intent- of ASNT-SNT-TC-1A by providing a 

methodology that appears to appropriately qual i fy and certify its eddy current 

testing personnel. 

I 

I 

3.10 Jnternal Audits 

Internal auditing takes place within Zetec's 2-QA quality program. Section 

XVIII, "Audits" of Z-QA, addresses the requirements contained in 

10 CFR Part 50 Appendix B and IS0 9001. 

house auditing of the EC function, safety-related/ASME Sections I11 and XI 

Code activities, on at least a "once per each calendar year" basis. 

verified that internal third-party assessments of Zetec's QA Group is 

performed by product services personnel at least once per calendar year. 

This document calls for a formal inThe team 

On the basis of its review of documentation and discussions with Zetec's QA 

Manager, the team concluded that Zetec's internal QA audits were adequate and 

met both the intent and scope of requirements in Criterion XVIII, "Audits," of 

10 CFR Part 50 Appendix B. 

The team also reviewed Zetec's Product Service Group's internal audits of QA 

for 1994 and 1995. On the basis of its review of documentation and 

discussions with the Product Services Manager, the team concluded that Zetec's 

audits of QA were adequate and that these audits met both the intent and scope 

of Criterion XVIII of 10 CFR Part 50 Appendix 8 requirements. 

At the entrance meeting on July 15, 1996, the NRC team leader discussed the 

scope of the inspection, outlined the areas to be inspected, and established 

contacts to be used during this inspection with Zetec management. 

meeting on July 18, 1996, the team discussed its findings and concerns. 

At the exit 

PERSONNEL CONTACTED 

Howard E. Houserman 

Jimmy Cri ttenden 

Doug Handy 

Laura McDonald 

Mike 0' Laugh1 in, 

Bob Vollmer 

Stephen H. von Fuchs 

Richard War1 i ck 

kern W i 1 son 

Troy Wol 1 er 

14 

144 

General Manager 

Product Support 

Electronic Supervisor 

Qual i ty Assurance 

Validation-Data Manager/Data Analyst 

Product Support Manager 

Qual i ty Assurance Manager 

Software Development Supervisor 

Software Administrative Assistant 

Software Support Specialist 
Selected Generic Correspondence on the Adequacy of 

Vendor Audits and the Quality of Vendor Products 

Ident i f i er Title 

Information Notice 96-62 Potential Failure of the Instantaneous Trip 

Function of General Electric RMS-9 Programmers 

Information Notice 96-68 Incorrect Effective Diaphragm Area Values in 

Vendor Manual Result in Potential Failure of 

Pneumatic Diaphragm Actuators 

Information Notice 96-70 Year 2000 Effect on Computer System Software 

145 
URC FORM 335 2-891 

lRCM 1102, 

t201,3202 

US. NUCLEAR REGULATORY COMMISSION 

BIBLIOGRAPHIC DATA SHEET 

(See instructions on the reverse] 

p. TITLE AND SUBTITLE 

Licensee Contractor and Vendor Inspection 

Status Report 

Quarterly Report 

October - December, 1996 

i. AUTHORW 

1. REPORT NUMBER 

(Assigned bv NRC. Add Vol., Supp., Rev., and Addendum Numbers, if any.) 

NUREG-0040 

Vol. 20, No. 4 

3. DATE REPORT PUBLISHED 

MONTH YEAR 

March 1997 

%. FIN OR GRANT NUMBER 

5. TYPE OF REPORT 

2uarterly 

7. PERIOD COVERED IlncIusiveDatesi 

ktober - December 19% 

1, PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC. provide Division, OfficeorRegfon, U.S. Nuclear Regulatory Commission, andrnai1ingaddress;if contractor, provide 

name and mailing address) 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

U.S. Nuclear Regulatory Comnission 

Washington, D.C. 20555-0001 

I. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, type “‘Sameasabove’:ifcontractor,provide NRC Division, Office or Region, U.S Nuclear Regulatory Commission, 

and mailing address1 

Same as above. 

0. SUPPLEMENTARY NOTES 

1. ABSTRACT Izm worn‘s or less) 

This periodical covers the results of inspections performed by the NRC’s Special 

Inspection Branch, Vendor Inspection Section, that have been distributed to the 

inspected organizations during the period from October - December 1996. 

13. AVAlLABl LlTY STATEMENT 2. KEY WORDS/DESCRIPTORS IList wordsorphrases that willassistresearchersin locating thereport.1 

Vendor Inspection 

Unlimited 

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I IThisPagel 

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(This Report) I 

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I I 

RC FORM 335 (2-89)