NUREG-0040 

Vol. 19, No. 3 

Licensee Contractor 

and Vendor Inspection 

Status Report 

Quarterly Report 

July - September 1995 

U.S. Nuclear Regulatory Commission 

Office of Nuclear Reactor Regulation 
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NUREG-0040 

Vol. 19, No. 3 

Licensee Contractor 

and Vendor Inspection 

Status Report 

Quarterly Report 

July - September 1995 

Manuscript Completed: December 1995 

Date Published: December 1995 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 
ABSTRACT 

This periodical covers the results of inspections performed by the NRC’s 

Special Inspection Branch, Vendor Inspection Section, that have been 

distributed to the inspected organizations during the period from July 1995 

through September 1995. 

iii 
DISCLAIMER 

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

States Government Neither the United States Government nor any agency thereof, nor 

any of their employees, make any warranty, express or impkd, or assumes any legal liabitity or mpomibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disdOse& 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, mauufacturer, or othenvise does not necessarily constitute or 

imply its endorsement, recommendation, or favoring by the United States Government or 

any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof. 
Porfions of this document may be iUegible 

in eiectronic image prodPct& Images are 

produced from the best a-e original 

document 
CONTENTS; 

PAGE 

Abstract ................................... :. ............................ i i i 

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

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

Continental Disc Corporation 

Liberty, Missouri 

Di vesco, Inc . Jackson , Mi ss i ss i ppi 

Eaton Corporation. 

Danbury, Connecticut 

General Electric Nucl ear Energy 

San Jose, Cal i forni a 

ITT Barton 

City of Industry, California 

Li berty Techno1 ogi es 

Conshohocken, Pennsylvania 

Mid-South Nuclear 

Birmingham, A1 abama 

Power Distribution Services 

West Chester, Ohio 

Rosemount Nuclear Instruments 

Eden Prairie, Minnesota 

Target-Rock Corporation 

Farmingdale, New York 

(99901287/95-01) ............... 2 

(99901117/95-01) ............... 20 

(99901290/95-01) ............... 34 

(99900403/95-01) ............... 44 

(99900113/95-01) ............... 53 

(99901225/95-01) ............... 62 

(99901270/95-01) ............... 71 

(99901286/95-01) ............... 82 

(99900271/95-02) ............... 98 

(99900060/95-01) .............. 125 

Westinghouse Electric Corporation (99900005/95-01) .............. 139 

Pittsburgh, Pennsylvania 

Select Generic Correspondence on the Adequacy of Vendor 

Audits and the Quality of Vendor Products ............... 214 

V 

INTRODUCTION 

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

licensing and inspection program is that licensees are responsible for the 

proper construction and safe and efficient operation of their nuclear power 

plants. 

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 requirements 

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

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 inspection effort is performed by the 

industry within the framework of quality verification programs. 

The Federal government and nuclear industry have established 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 suppliers, contractors and vendors also have 

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

codes, and standards. 

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

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

(AE) firms, suppliers of products and serviices, independent testing 

laboratories performing equipment qualification tests, and holders of NRC 

construction permits and operating licenses in vendor-related areas. 

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

problems are determined and appropriate corrective actions are developed. The 

inspect ions a1 so review vendors to veri fy conformance wi th appl i cab1 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, licensee-vendor interface, environmental qualification of equipment, 

and review of equipment problems found during operation and their corrective 

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

the inspected organization is required to take appropriate corrective action 

and to institute preventive measures to preclude recurrence. 

imp1 ications are found, NRC ensures that affected 1 icensees are informed 

through vendor reporting or by NRC generic correspondence such as information 

notices and bulletins. 

When generic 

vi i 
Thi s quarterly report contains copi es of a1 1 vendor inspection reports i ssued 

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 list selected bulletins, generic 

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

correspondence involving vendor issues. 

Each vendor inspection 

viii 
I NS P ECT I ON REI PORTS 
. ***** . 

JNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-OOOl 

September 14, 1995 

Mr. Kenneth R. Shaw, President 

Continental Disc Corporation 

3160 West Heartland Drive 

Liberty, MO 64068 

SUBJECT: NRC INSPECTION NO. 99901287/95-01 

Dear Mr. Shaw: 

This letter addresses the U.S. Nuclear Regulatory Commission (NRC) inspection 

of your facility at Liberty, Missouri, conducted by Messrs. Bill Rogers and 

Robert Pettis of this office on June 27 through 29, 1995, and the discussions 

of their findings with you at the conclusion of the inspection. The 

inspection was conducted to evaluate your quality assurance program and its 

implementation in selected areas such as control of purchased material 

and services, supplier audits, manufacturing control and a review of your 

program for implementing Part 21, "Reporting Defects and Noncompliance," of 

Title 10 of the Code of Federal Resulations. 

Areas examined during the NRC inspection and our findings are discussed in the 

enclosed inspection report. 

procedures and representative records, discussion, and observations by the 

inspectors. 

This inspection consisted of an examination of 

A1 though your qual ity assurance program implementation was generally 

satisfactory, the inspection identified that it did not meet applicable NRC 

requirements in the areas of control of purchased material, equipment, and 

services. 

are identified in the enclosures to this letter. 

The specific findings and references to the pertinent requirements 

Please provide us within 30 days from the date of this letter a written 

statement in accordance with the instructions specified in the enclosed Notice 

of Nonconformance. 

show good cause for us to do so. 

We will consider extending the response time if you can 

The responses requested by this letter and the enclosed Notice are not subject 

to the clearance procedures of the Office of Management and Budget as required 

by the Paperwork Reduction Act of 1980, Public Law No. 96-511. In accordance 

with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter and 

the enclosed inspection report will be placed in the NRC Public Document Room. 

2 
Mr. Shaw -2- 

If there are any questions concerning this inspection we will be pleased to 

discuss them with you. 

Sincerely, 

Speci a1 Inspect ion Branch 

Division of Inspection and 

Office of Nuclear Reactor Regulation 

Support Programs 

Docket No. 99901287 

Enclosures: 1. Notice of Nonconformance 

2. Inspection Report 99901287/95-01 
NOTICE OF NONCONFOMCE 

Continental Disc Corporation 

Liberty, Missouri 

Docket No.: 99901287 

Based on the results of an NRC inspection conducted at the Liberty, Missouri, 

facility of Continental Disc Corporation on June 27-29, 1995, it appears that 

certain of your activities were not conducted in accordance with NRC 

requirements. 

I. Criterion V of Appendix B to 10 CFR Part 50, "Instructions, Procedures, 

and Drawings," states, in part, 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 instructions, procedures, or drawings. 

Contrary to the above, Continental Disc Corporation Quality Assurance 

Manual, General Revision 1, issued April 1, 1990, did not address the 

quality requirements of non-pressure boundary parts, exempt from 

American Society of Mechanical Engineers (ASME) Section 111, Division I, 

NCA-1275, "Rupture Disc Devices," used in the manufacture of nuclear 

safety-related rupture discs, an activity affecting quality. (95-01-01) 

11. Criterion VI1 of Appendix B to 10 CFR Part 50, "Control of Purchased 

Material, Equipment and Services," states, in part, that measures shall 

be established to assure that purchased material, equipment, and 

services conform to the procurement documents and include provisions for 

source evaluation and selection, objective evidence of quality furnished 

by the contractor or subcontractor, inspection at the contractor or 

subcontractor source, and examination of products upon delivery. 

Contrary to the above, Quality Control Instruction No. 1014, Revision C, 

dated March 20, 1995, stated in paragraph 8.2 of Section 8.0, "Supplier 

Approval Criteria," that suppliers could be qualified on tkz sinjle 

basis of "completion of a Quality System Questionnaire. I' Coctinevtal 

Disc Corporation's method of qualification for certain Suppliers was 

limited to the supplier completing a Quality System Questionnaire which 

did not provide objective evidence of quality to demonstrate that 

purchased materi a1 conformed to procurement documents. 

Consequently, as documented by the review of the qualification 

activities for Teledyne Rodney, Metal Goods, Castle Metals, Joseph T. 

Ryerson & Son, and Coulter Steel & Forge, Continental Disc Corporation 

did not adequately qualify suppliers furnishing materials or services 

used in the manufacture of nuclear safety-related rupture discs and 

rupture disc holders. Additionally, as documented by review of the 

qualification activities for Teledyne Rodney, Metal Goods, and Sherry 

Laboratories, Continental Disc Corporation accepted material test 

reports from suppliers furnishing material and test services without a 

basis for accepting such documentation and subsequently certified to its 

customers that the material complied with the purchase order 

requirements. (95-01-02) 

Enclosure 1 

4 
111. Criterion V of Appendix B to 10 CFR Part 50, "Instructions, Procedures, 

and Drawings," states, in part, 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 instructions, procedures, or drawings. 

Criterion VI1 of Appendix B to 10 CFR Part 50, "Control of Purchased 

Haterial, Equipment and Services," states, in part, that measures shall 

be establ ished to assure that purchased material, equipment, and 

services conform to the procurement documents. 

Paragraph 4.5 of Section 4.0, "Purchasing Documents, " of the Continental 

Disc Corporation Qual ity Assurance Manual stated that when regulatory 

requirements, design bases or other requirements necessary to assure 

adequate qual i ty are incorporated in a Continental Di sc Corporation 

customer's purchase order, these requirements shall be referenced in the 

purchase order to Continental Disc Corporation suppliers and apply to 

suppl ier and sub-t i er suppl i er performance. 

Continental Disc Corporation Quality Control Instruction No. 1003, 

"Purchasing and Supplier Quality Assurance Pol icy," Revision F, dated 

March 20, 1995, stated in Paragraph 6.7 of Section 6.0, "Supplier 

Quality Policy," that suppliers shall flow down all purchase order 

requirements to any authorized sub-tier suppliers. 

Contrary to the above, as documented by review of CDC purchase orders 

No. 49912 to Teledyne Rodney, dated May 18, 1995; No. 49874 to Castle 

Metals, dated May 16, 1995; No. 47600 to Joseph T. Ryerson, dated 

October 26, 1994; No. 47107 to Coulter Steel & Forge, dated 

September 20, 1994; and No. 49909 to Metal Goods, dated May 18, 1995; 

Continental Disc Corporation did not impose customer required 10 CFR 

Part 50, Appendix B, quality requirements in purchase orders to its 

suppliers furnishing materials or services used in the manufacture of 

nuclear safety-re1 ated rupture discs and rupture disc holders. 

(95-01-03) 

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

Regulatory Comnission, 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: 

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

that have been or will be taken to prevent recurrence; and (3) the dates your 

corrective actions and preventive measures were or wi 1 1 be compl eted. 

(1) a description of the steps that have 

Eated at Rockville, Maryland 

this 14th day of September, 1995 

2 

5 
REPORT NO.: 

ORGANIZATION: 

ORGANIZATIONAl 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF INSPECTION AND SUPPORT PROGRAMS 

99901287/95-01 

Continental Disc Corporation 

3160 Heartland Drive 

Liberty, Missouri 64068 

Dean Dachenhausen, Director 

Quality Assurance 

Continental Disc Corporation provides rupture discs 

and rupture disc holders to the nuclear industry. 

INSPECTION DATES: June 27 - 29, 1995 

TEAM LEADER: 

Bill H. Roclerd 

Vendor Inspection Section 

Special Inspection Branch 

OTHER INSPECTORS: Robert L. Pettis, Jr. 

Vendor Inspection Section 

Special InspeLtion Branch 

REVIEWED BY: 

Gregory@.&wal ina, Chief Date 

Vendor Inspection Section 

Special Inspection Branch 

APPROVED BY: 

Robert M. Gallo, Chief, 

Special Inspection Branch 

Enclosure 2 

6 
1 

Duri 

SUMMARY OF INSPECTION FINDINGS 

9 this insDection. the NRC inspe t rs evaluat d Continental Dis 

Corpoiation's (CDC) implementation * of quality assurance measures for those 

activities which directly affected the qual fty and performance capability of 

their product. These activities included the control of purchased material 

and services, audits, and production control. The team also reviewed CDC's 

program for implementing Part 21, "Reporting 3efects and Noncompliance," of 

Title 10 of the Code of Federal Rewlations (10 CFR Part 21). 

1.1 Violations 

1.1.1 

suppliers specify that the provision of 10 CFR Part 21 apply when applicable, 

CDC Quality Control Instruction No. 1003, "Purchasing and Supplier Quality 

Assurance Policy," Revision F, dated March 20, 1995, (which required that 

suppliers shall flow down all purchase order requirements to any authorized 

sub-tier suppliers), and Quality Control Instruction No. 1020, "Nuclear Safety 

Related Materials," Revision F, dated February 6, 1990, (which required that 

compliance with 10 CFR Part 21 shall be mandatory on all shop orders and all 

purchase orders to suppliers when imposed by the customer purchase order), CDC 

did not impose the requirements of 10 CFR Part 21 i n i t s purchase orders to 

suppliers furnishing material or services used in the manufacture of nuclear 

safety-re1 ated rupture discs and rupture di sc holders to f i 11 customer 

purchase orders from licensees specifying that the requirements of 10 CFR 

Part 21 applied. (Non-Cited violation) 

1.2 Nonconformances 

Contrary to 10 CFR 21.31, which requires that purchase orders to 

1.2.1 Contrary to Criterion V of Appendix B to 10 CFR Part 50, "Instructions, 

Procedures, and Drawings, 'I the CDC Qual i ty Assurance Manual, General Revision 

1, issued April 1, 1990, did not address the quality requirements of nonpressure boundary parts, exempt from American Society of Mechanical Engineers 

(ASME) Section 111, Division 1, NCA-1275, and used in the mdnufaCLure of 

nuclear safety-related rupture discs. (95-01-01) 

1.2.2 Contrary to Criterion VI1 of Appendix B to 10 CFR Part 50, "Control of 

Purchased Material, Equipment and Services," and Sections 7.2 and 7.8 of the 

CDC Quality Assurance Manual, CDC did not adequately qualify suppl iers 

furnishing materials or services used in the manufacture of nuclear safetyrelated rupture discs and rupture disc holders. 

qualification was limited to the supplier completing a Quality System 

Questionnaire which did not provide sufficient objective evidence to 

demonstrate that the supplier is effectively implementing its quality program. 

CDC's method of supplier 

Additionally, CDC did not independently verify material test reports and 

certificates submitted to them from these suppliers. 

supplied by CDC to its customers purchasing nuclear safety-related rupture 

discs and rupture disc holders. 

Such documentation was 

(95-01-02) 

2 

7 
1.2.3 Contrary to Criterion V of Appendix B to 10 CFR Part 50, "Instructions, 

Procedures, and Drawings;" Criterion VI1 of Appendix B to 10 CFR Part 50, 

"Control of Purchased Material, Equipment and Services;" Paragraphs 4.5 and 

4.6 of Section 4.0, "Purchasing Documents," of the CDC Quality Assurance 

Manual; Paragraph 6.7 of Section 6.0, "Supplier Quality Policy of CDC Quality 

Control Instruction No. 1003, Revision F, dated March 20, 1995; and Paragraph 

5.1 of CDC Quality Control Instruction No. 1020, Revision F, dated February 6, 

1990; CDC did not impose customer required 10 CFR Part 50, Appendix B, quality 

requirements in purchase orders to its suppliers furnishing materials or 

services used in the manufacture of nuclear safety-related rupture discs and 

rupture disc holders. (95-01-03) 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

This was the first NRC inspection of CDC. 

3 INSPECTION FINDINGS AND OBSERVATIONS 

3.1 Entrance and Exit Meetinas 

During the entrance meeting on June 27, 1995, the NRC inspectors discussed the 

inspection scope and developed general information about CDC's products and 

activities. During the exit meeting on January 29, 1995, the NRC inspectors 

discussed their findings and observations with CDC management. 

3.2 Backaround 

CDC supplies rupture discs and rupture discs holders to the nuclear industry 

as well as for many general industrial applications. Rupture discs, used to 

provide instantaneous relief of overpressure conditions, are manufactured in a 

variety of designs of various metals and combinations of metals and teflon 

liners. The rupture discs can also be designed to resist vacuum, be nonfragmenting, and provide indication when rupture has occurred. 

3.3 10 CFR Part 21 Proaram 

The inspectors reviewed CDC's 10 CFR Part 21 program including procedures and 

implementation. 

had suppl ied and consequently had not performed any evaluations in accordance 

with their 10 CFR Part 21 program. 

The inspectors reviewed CDC's 10 CFR Part 21 implementing procedure, Quality 

Control Instruction No. 1013, "Compliance With 10-CFR-21," Revision A, dated 

March 30, 1995. 

several instances were the terms deviation and defect were inappropriately 

interchanged. 

Part 21, and other minor procedural discrepancies were discussed with CDC who 

indicated that the procedure would be modified. 

CDC had not identified any deviations in products that they 

The procedure was adequately written with the exception of 

The use of these terms, and their definitions in 10 CFR 

3 

8 
The inspectors reviewed CDC's posting as required by 10 CFR 21.6 and 

determined it to be in accordance with the regulation. 

Section 206 or the Energy Reorganization Act and a notice which described the 

regulations and applicable procedures, including the name of the individual to 

whom the reports could be made, and where the regulations and procedures could 

be examined. 

CDC had posted 

The inspectors identified one area where implementation of the CDC 10 CFR 

Part 21 program was not adequate. 

that suppliers flow down all purchase order requirements to any authorized 

sub-tier suppliers and that compliance with 10 CFR Part 21 shall be mandatory 

on all shop orders and ali purchase orders to suppliers when imposed by the 

customer purchase order. 

did not comply with these requirements. This issue is discussed in detail in 

section 3.11 of this report. 

CDC Quality Control Instructions required 

The inspectors identified several examples where CDC 

3.4 Internal Audits 

Internal audits were required to performed by Section 18, "Audits," of the CDC 

Qual ity Assurance Manual. The inspectors reviewed the implementing procedure 

Quality Control Instruction No. 1001, "Qual ity Assurance Internal Audit," 

Revision H, dated December 2, 1994. The procedure required that internal 

audits be performed annually with the frequency to be shortened if determined 

necessary based on previous results, corrective actions, nonconformances or 

customer feedback. 

to prepare the audit plans, train auditors, assign audit personnel, and 

evaluate the results. 

internal audits. 

the department where the finding was identified. 

The procedure required the Director of Quality Assurance 

An audit plan and audit checklist were used for the 

Corrective actions were to be provided by the managers of 

The inspectors reviewed the reports for the internal audits which had been 

performed December 20-21, 1994, June 20-21, 1994, November 22-23, 1993, and 

December 16-21, 1992. 

specific audits which included material control, material traceability, 

manufacturing, document control, design control, and work in process and did 

not identify any findings. 

documented a thorough process which was in accordance with the requirements of 

the imp1 ement i ng procedure. 

The audit reports specified the areas reviewed for the 

The inspectors concluded that the audit reports 

3.5 RUDtUre Disc Productim Process 

The inspectors reviewed the applicable procedures, and their implementation, 

for selected portions of the rupture disc manufacturing process. 

discs come in a variety of combinations including metal rupture discs, teflon 

seals, and those with backpressure supports. The inspectors reviewed Qual ity 

Control Instructions, Manufacturing Procedures and Production Operating 

Procedures applicable to the production of rupture discs and disc holders and 

observed portions of the production process. 

Rupture 

4 

9 
Quality Control Instruction No. 1023, "Control of Rupture Disc Production 

Processes," Revision A, dated June 4, 1991, described the control and 

documentation procedures appl i cabl e to rupture di scs , the conditions under 

which rupture disc manufacturing and testing were performed, and the 

documentation maintained to provide evidence of compliance with the CDC 

qual i ty assurance program. The procedure i ncl uded product ion personnel 

training and qualification, work instructions, material purchases, material 

controls, production flow, lot identification control, shipping department 

control , final inspection, packing, delivery, and documentation. 

Manufacturing Procedures (MP) were used to provide instruction on the specific 

steps of rupture disc manufacturing. MPs included instructions on the order 

of precedence of referenced documents (shop order having the highest 

precedence), testing of rupture discs using teflon liners, and preparing for 

the final burst tests. The inspectors discussed the use of MPs with COC 

personnel and observed the manufacture of several rupture discs. 

Production Operating Procedures (POP) provided instruction on additional 

activities related to the production of rupture discs such as parts forming, 

parts etching, and teflon forming. For example, POP 1000, Initial Issue, 

dated January 26, 1995, provided specific quality requirements on the process 

of teflon forming such as no burn holes or pin holes were allowed in the 

teflon seals. 

forming and the subsequent check for pin holes. 

The inspectors observed a successful demonstration of teflon 

CDC used a rupture disc burst test to demonstrate the quality of each lot of 

rupture discs manufactured. 

on the number of rupture discs in the order. 

No. 2000, "Final Lot Rupture Disc Burst Test Procedure," Revision E, dated 

February 3, 1995, required a sample of all lots manufactured to be burst 

tested (a destructive test) to verify that the rupture discs manufactured in 

the lot would perform as designed. Lot size was based on Mil-Std-1OSE and CDC 

MP 2003 and required a minimum sample size of two burst tests for any order or 

lot manufactured. 

would manufacture three rupture discs and burst test two of the three. 

addition, if ordered quantities were broken down into smaller lots for control 

purposes the burst test sample size would be based on the smaller lots. The 

sample to be burst tested was selected by a burst test witness at random from 

the rupture discs or rupture disc components and were to include all rupture 

disc components. The inspectors observed the successful burst tests of 

several rupture discs and also witnessed a failure of rupture disc material 

(premature rupture) where the materi a1 was rejected and appropriately 

dispositioned. 

CDC made rupture discs of varying lot sizes based 

Quality Control Instruction 

For example, if the order required one ruplu.I- disc, CDC 

In 

The inspectors concl uded, based on a review of the appl i cabl e procedures, 

discussion with personnel, and observations of portions of the production 

process, that the CDC quality assurance program was well implemented in the 

production area and that production testing was adequate to ensure the quality 

of the delivered rupture discs. 

5 

10 
3.6 Customer Audits of CDC 

The NRC inspection team reviewed CDC document "Customer On-si te Qual ity Audits/ Approvals," dated June 12, 1995. 'The document listed all customer 

audits performed at CDC, including the audit criteria, audit date and results. 

CDC had been recently audited by the following customers: 

June 3, 1992, Astro Nuclear/Dynamics on June 14, 1994, Public Service Electric 

& Gas on June 7, 1995, Commonwealth Edison on November 15, 1991, and 

Philadelphia Electric on February 10, 1992. 

Entergy on 

3.7 Customer Purchase Orders to CDC 

The NRC inspection team reviewed selected customer purchase orders to CDC 

which specified the requirements of 10 CFR Part 50, Appendix 8, and 10 CFR 

Part 21 for nuclear safety-related rupture discs and holders since 1990. 

purpose of the purchase order review was to determine if CDC had properly 

implemented its quality assurance program, particularly in the area of raw 

material supplier qualification. The inspectors reviewed the following 

purchase orders: 

The 

3.7.1 Commonwealth Edison Company 

Commonwealth Edison Company (CECo) ordered six, safety-related, 6-inch rupture 

discs on purchase order No. 4T8711, dated March 21, 1994, for its LaSalle 

Station No. 1. The purchase order requested certified material test reports 

(CMTR) and a Certificate of Compliance to the purchase order requirements 

which required compliance to 10 CFR Part 50, Appendix By 10 CFR Part 21 and 

ASME Section 111, Division .l, Class 2, 1974 Edition. The purchase order also 

stated in Statement 0013 that when subtier vendors are utilized, the 

appropriate qual i ty assurance program requirements shall be incorporated in 

the procurement documents. 

CDC processed the order using material furnished by Teledyne F., '7- y (CDC 

purchase ordv NQ 41291) and an unidentified supplier (CDC purchase order No. 

33535) which supplied an Inconel 600 vacuum support. The vacuum support 

material was sent to Metlab Testing Services, Inc., which verified on 

October 2, 1991, that the chemical analysis conformed to Inconel 600. CDC 

passed on the Metlab chemical analysis certificate (No. 91-6179) and a 

Material Test Certificate from Teledyne Rodney (No. 18141) under a CDC 

Certificate of Conformance which certified that the materials were furnished 

in strict accordance with the requirements and applicable specifications of 

the purchase order. 

3.7.2 Niagara Mohawk Power Corporation 

Niagara Mohawk Power Corporation (NMPC) ordered sixty, safety-related l-inch 

rupture discs of various pressure and temperature ratings on purchase order 

No. 13874, dated October 14, 1993, for its Nine Mile Point, Unit 2, nuclear 

plant. 

certificate of compliance to the purchase order requirements which required 

The purchase order requested CMTRs Far the disc material and a 

6 
compliance to 10 CFR Part 50, Appendix B, 10 CFR Part 21, ASME Code Section 

111, Division 1, Class 2, 1980 Edition through the Winter 1980 addenda and for 

CDC to extend to sub-suppliers all appropriate technical and quality 

requirements. 

CDC processed the order using material furnished by Teledyne Rodney (CDC 

purchase order not identified). 

(No. 10706) from Teledyne Rodney under a CDC Certificate of Conformance which 

certified that the materials were furnished in strict accordance with the 

requirements and applicable specifications of the purchase order. 

CDC passed on a Material Test Certificate 

3.7.3 PECO Energy Company - Peach Bottom 

PECO Energy Company (PECO) ordered eight, 16-inch, safety-re1 ated rupture 

discs on purchase order No. 8W230572, dated August 9, 1994, for its Peach 

Bottom Atomic Power Station. 

cert i f i cate of compl iance to the purchase order requi rements . Additional 

purchase order requirements included CDC’s compliance to 10 CFR Part 21, 

furnish materials in accordance with CDC’s Quality Assurance Manual , Revision 

1, dated April 1, 1990, ASME Code Section 111, qivision 1, Class 2, 1986 

fdition through the Winter 1987 addenda and extend to lower tier suppliers all 

applicable quality assurance program requirements. 

The purchase order requested CMTRs and a 

CDC processed the order using material furnished by Teledyne Rodney (CDC 

purchase order No. 42200) and Castle Metals, Inc. (CMI)(CDC purchase order not 

identified). CDC passed on a Material Test Certificate (No. 29099) from 

Teledyne Rodney and a Certificate of Test from CMI under a CDC Certificate of 

Conformance which certified that the materials were furnished in strict 

accordance with the requirements and applicable specifications of the purchase 

order. 

3.7.4 PECO Energy Company - Limerick 

PECQ ordered a 16-inch, safety-related rupture disc on purchase order Na. 

LS237.030, dated February 2, 1994, for its Limerick Nuclear Generating Station. 

The purchase order requested CMTRs and a certificate of compliance to the 

purchase order requirements which included Compliance to 10 CFR Part SO, 

Appendix 8, 10 CFR Part 21, ASME Code Section 111, Division 1, Class 2, 1986 

Edition through the Winter 1987 addenda and to extend to 1 ower tier suppl i ers 

a1 1 appl icabl e qual i ty assurance program requirements. 

CDC processed the order using material furnished by Teledyne Rodney (CDC 

purchase order No. 41207) and Joseph T. Ryerson & Son (CDC purchase order No. 

43240). CDC passed on a Material Test Certificate (No. 26736) from Teledyne 

Rodney and a Hetallurgical Test Report (No. 50302) from North American under a 

CDC Certificate of Conformance which certified that the materials were 

furnished in strict accordance with the requirements and applicable 

specifications ot the purchasE order. 

7 

12 
3.7.5 Georgia Power Company 

Georgia Power Company (GPC) ordered two, safety-related 16-inch rupture discs 

on purchase order No. 60179000000, dated September 1, 1994, for its Edwin I. 

Hatch Nuclear Plant. The purchase order requested CMTRs and a certificate of 

compliance to the purchase order requirements which included compliance to 10 

CFR Part 50, Appendix By 10 CFR Part 21, ASM Code Section 111, Division 1, 

Class 2, 1980 Edition through the Summer 1981 addenda and provide access to 

lower tier suppliers for quality assurance inspection or audit by GPC. CDC 

processed the order using material furnished by Teledyne Rodney (CDC purchase 

order No. 42000), CMI (CDC purchase order No. 46244) and Metal Goods (CDC 

purchase order No. 46331). 

CDC passed on two Material Test Certificates (Nos. 27140 and 30547) from 

Teledyne Rodney and a Certificate of Test from Allegheny Ludlum Steel (through 

CMI) and a Test Report (No. 50302) from J&L. Specialty Products Corporation 

(through Metal Goods) under a CDC Certificate of Conformance which certified 

that the materials were furnished in strict accordance with the requirements 

and applicable specifications of the purchase order. 

3.7.6 Public Service Electric & Gas Company 

Public Service Electric & Gas Company (PSE&G) ordered four, 12-inch, safetyrelated rupture discs on purchase order No. P2-0699501, dated June 9, 1994. 

The purchase order requested CMTRs and a certificate of compliance to the 

purchase order requirements which included compliance to 10 CFR Part 50, 

Appendix B, 10 CFR Part 21 and ASME Code Section 111, Division 1, Class 2, 

1977 Edition. 

CDC processed the order using materizl furnished by Teledyne Rodney (CDC 

purchase order No. 41291) and CMI (CDC purchase order No. 42341). CDC passed 

on a Material Test Certificate (No. 16866) from Teledyne Rodney and a CMTR 

from Inco Alloys International (through CMI) under a CDC Certificate of 

Conformance which cevtified that the materials were furnished in strict 

accordance with the requirements and applicable specifications of the purchase 

order. 

3.8 Classification of ComDonents 

ASME Code Section 111, Division 1, NCA-1275, "Rupture Disc Devices," stated 

that the rupture disc holder was the only portion of the rupture disc 

considered part of the pressure boundary and therefore was the only component 

designated "Code materi a1 I' and subject to material control s. The CDC Qual i ty 

Assurance Manual did not address the quality requirements of exempt, nonpressure boundary parts. 

not considered a pressure boundary part under ASME Code, CDC had not 

classified the rupture disc material as safety-related. The inspection team 

pointed out that although the ASME Code exempts the disc from consideration as 

Code material, such non-pressure boundary safety-related items must be 

processed and controlled in accordance with the applicable requirements of 10 

CDC stated that since the rupture disc itself was 

8 

13 
CFR Part 50, Appendix B. Although the rupture disc material is designed to 

fail (in a specified and predictable manner) materials are often chosen for 

use due to other considerations (such as resistance to corrosion) and could 

therefore have importance to the overall safety function of the rupture disc 

assembly. In addition, the rupture disc material is often specified by the 

customer on the purchase order and is therefore a technical requirement 

requiring verification and documentation. CDC's failure to properly classify 

material and components and accordingly the failure to take appropriate 

actions to verify that the material and components have been manufactured and 

controlled under an acceptable quality assurance program which has been 

properly implemented has been identified as Nonconformance 95-01-01. 

NRC Information Notice (IN) 88-95, "Inadequate Procurement Requirements 

Imposed By Licensees On Vendors," dated December 8, 1988, discussed inadequate 

procurement requirements being imposed by licensees on vendors supplying 

components under the ASME Code which may result in the vendor's failure to 

implement critical portions of 10 CFR Part 50, Appendix B, quality assurance 

requirements. Specifically, the IN stated that compliance with ASME Section 

I11 satisfies 10 CFR Part 50, Appendix B, requirements for items covered by 

the Code, however, this is not sufficient to ensure that safety-related items 

exempt from Code requirements comply with 10 CFR Part 50, Appendix B. In 

addition, IN 90-03, "Mal function Of Borg-Warner Bo1 ted Bonnet Check Valves 

Caused By Failure Of The Swing Arm," dated January 23, 1990, discussed an 

event which occurred at the Comanche Peak Steam Electric Station (CPSES) where 

a 4-inch 150 pound BW/IP (formerly Borg-Warner Nuclear Valve Division) bolted 

bonnet swing check valve, installed in the service water system at the CPSES, 

Unit 1, exhibited excessive backleakage. The NRC concluded that BW/IP's 

classification of the swing arm as a non-pressure boundary valve internal, and 

exempt under the requirements of ASME 111, resulted in the failure of BW/IP to 

impose nuclear quality assurance requirements on the swing arm manufacturer. 

Inadequate heat treatment by a supplier who was not on BW/IP's approved 

suppliers list was identified as the cause of the radial fracture of the swing 

arm. 

assembly to the clevis. 

The disc separated from the swing arm which connected the disc and stud 

3.9 Oual if i cat i on of Materi a1 Suwl i ers 

The inspectors reviewed Section 7.2 of the CDC Quality Assurance Manual which 

stated that purchases shall be made from an approved suppliers list jointly 

maintained by the Purchasing Department and the Quality Assurance Department. 

The February 6, 1995, CDC approved suppliers list, defined five categories of 

suppliers, Codes I through V. The inspectors reviewed documentation for 

selected Code I (suppliers of metal and plastic raw materials used for rupture 

discs and rupture disc holders) and Code IV (suppliers of quality assurance 

services, calibration, and testing) suppliers. 

Quality Control Instruction No. 1014, Revision C, dated March 20, 1995, stated 

in paragraph 8.2 of Section 8.0, "Supplier Approval Criteria," that Code I 

through Code IV suppliers may achieve quality approval status by one or more 

of the following means: 1) completion of a Quality System Questionnaire, 2) 

9 

14 
performance of an on-si te qual i ty survey, 3) qual i ty cert i f i cat i on provided by 

the supplier documenting use of an IS0 9000 standard, an ASME Quality System 

Certificate or authorization stamp, 4) the review and approval by the CDC 

Quality Department of the suppl ier’s qual i ty manual. Qual i ty Control 

Instruction No. 1014 also stated that an implementation audit or survey of the 

supplier was not required for those suppliers which hold an ASME Quality 

System Certificate. The inspectors concluded that a7 though “performance of an 

on-site quality survey,” used alone could be an acceptable method of 

qualification, any of the other three methods, ”completion of a Quality System 

Questionnaire,” “quality certification provided by the supplier documenting 

use of an IS0 9000 standard, an ASME Quality System Certificate or 

authorization stamp,” or “the review and approval by the CDC Quality 

Department of the supplier’s quality manual ,“ used alone, would not be 

adequate to ensure the effective implementation of a supplier’s approved 

qual ity assurance program in order to supply safety-related material to be 

used without further verification. 

Typical methods of qual ifying companies for placement of safety-related, 

10 CFR Part 21, purchase orders (material or services to be used without 

further verification) are 1) ASME supplier - inclusion on the approved 

suppliers list based on an ASME Quality System Certificate followed by a 

review of the supplier’s quality assurance program with an associated 

implementation audit or 2) Non-ASME supplier - review of the supplier’s 

quality assurance program with an associated implementation audit. In 

addition, for suppliers that would not accept safety-related, 10 CFR Part 21, 

purchase orders other means of ensuring quality could be used including test 

and measurement , source survei 11 ance, commerci a1 grade surveys, and 

performance history (used in combination with one or more of the other 

methods). 

The inspectors reviewed the documentation for the qualification activities for 

several companies that CDC had purchased material and services from. These 

companies were Teledyne Rodney, Metal Goods, Castle Metals a,-i _’- eph T. 

Ryerson & Son. The suppliers were listed on CDC’s approved suppliers list, 

dated February 6, 1995, as Code I suppliers, and had been used by CDC since 

1977. All of these suppl iers were qual ified by CDC only on the basis of 

providing a satisfactory response to a CDC Quality System Questionnaire issued 

in January 1994 (with the exception of Teledyne Rodney who also had maintained 

an IS0 9000 quality assurance program). In addition, CDC had passed on to its 

customers material test reports and certifications submitted to them from 

Teledyne Rodney and Metal Goods without performing an independent verification 

of the basis of such documentation. 

these suppliers were never verified through the performance of an 

implementation audit, survey or other appropriate means to objectively assess 

quality, CDC had no documented basis for accepting and supplying such material 

test reports to its customers and furnishing certification that CDC was in 

full compl iance with a1 1 purchase order requirements. 

Since the quality assurance programs of 

The inspectors reviewed section 7.8 of the CDC Quality Assurance Manual which 

stated that pressure boundary nuclear safety-re1 ated materi a1 s shall be 

10 

15 
purchased only from suppliers holding current ASME authorizations or 

certificates. 

by suppl iers of pressure boundary nuclear safety-related materials shall be 

reviewed by qual i ty assurance personnel for code compl iance and adequacy. 

Quality Control Instruction No. 1020, "Nuclear Safety Related Materials," 

Revision F, dated February 6, 1990, required in Paragraph 5.5 of Section 5.0, 

"Requirements," that purchase orders fw all pressure boundary and flange 

bolting materials shall be issued only to approved suppliers holding current 

ASME stamps or Quality System Certificates for the material required. 

inspection team identified that only Coulter Steel 81 Forge appeared on CDC's 

approved suppliers list as a Code I supplier holding a Quality System 

Certificate and was the only supplier of rupture disc holder material. 

However, their qualification basis as a Code I supplier was solely based on a 

satisfactory response to a CDC furnished Quality System Questionnaire. The 

NRC inspection team discussed with CDC that possession of a Quality System 

Questionnaire is acceptable to place a supplier on the approved suppliers list 

for programmatic aspects of the suppliers quality assurance program, but prior 

to accepting material or services from the supplier, CDC would have to perform 

an implementation audit to verify that the supplier is effectively 

implementing its approved 10 CFR Part 50, Appendix B, quality assurance 

program. 

Paragraph 7.8.1 stated that CMTRs or certifications furnished 

CDC 

The NRC 

CDC had used a testing laboratory (Sherry Laboratories) to analyze rupture 

disc holder material; however, CDC had not taken action to verify that the 

laboratory had an adequate quality assurance program which was being 

effectively implemented (see section 3.10). 

CDC's failure to ensure that suppliers furnishing material used in the 

manufacture of nuclear safety-related rupture discs and rupture disc holders 

were qualified and effectively implementing the approved quality assurance 

programs has been identified as Nonconformance 95-01-02. 

NRC IN 86-21, "Recognition Of ASME Accreditation Program For N Stamp H-ilders," 

dated March 31, 1986, discussed that the NRC's recognition applied only to the 

programmatic aspects of the ASME Accreditation Program and that 1 icensees and 

their subcontractors were still responsible for ensuring that the supplier is 

effectively implementing its approved quality assurance program. 

3.10 Oualification of Testins Laboratory 

The NRC review identified that some material furnished by Teledyne Rodney and 

Metal Goods had been sent to Sherry Laboratories (formerly Metlab Testing 

Services, Inc . ) , for spectrographi c examination. However Sherry Laboratories , a Code IV supplier listed on CDC's approved suppliers list since 1990, was 

qualified by CDC based on "performance," which is defined by CDC as historical 

satisfactory performance. However, the "Approval Criteria" section of the 

approved suppl iers 1 i st for Code IV suppl iers of qual ity assurance services, 

including calibration and testing, did not recognize the "performance" 

approach to supplier qualification (the approved suppliers list stated 

"testing laboratories have no pre-qualification requirements"). This is 

16 
contrary to Paragraph 8.2 of Quality Control Instruction No. 1014, "Supplier 

Selection and Quality Assessment," Revision Cy dated March 20, 1995, which did 

not recognize supplier "performance" as a means of supplier qualification. 

In addition, CDC passed on to its customers material test reports and 

certifications submitted to them from Sherry Laboratories without performing 

an independent verification of the basis of the documentation. 

quality assurance program had not been verified through the performance of an 

implementation audit, survey or other appropriate means to objectively assess 

quality, CDC had no documented basis for accepting and supplying such material 

test reports to its customers and furnishing certification that CDC was in 

full compl iance with a1 1 purchase order requirements. CDC' s fai 1 ure to ensure 

that the supplier furnishing metallurgical testing services used in the 

manufacture of nuclear safety-re1 ated rupture discs and rupture disc holders 

was qualified and effectively implementing an approved quality assurance 

program has been identified as a second example of Nonconformance 95-01-02. 

Since the 

3.11 Review of CDC Purchase Orders to Sutwliers 

The inspectors reviewed paragraph 4.5 of Section 4.0, "Purchasing Documents," 

of the CDC Quality Assurance Manual which stated that when regulatory 

requirements, design bases or other requirements necessary to assure adequate 

quality are incorporated in a CDC customer's purchase order, these 

requirements shall be referenced in the purchase order to CDC suppliers and 

apply to supplier and sub-tier supplier performance. 

CDC Quality Control Instruction No. 1003, "Purchasing and Supplier Quality 

Assurance Policy," Revision F, dated March 20, 1995, required in Paragraph 6.7 

of Section 6.0, "Supplier Quality Policy," that suppliers shall flow down all 

purchase order requirements to any authorized sub-tier suppliers. 

Additionally, CDC Quality Control Instructiion No. 1020, "Nuclear Safety 

Related Materials," Revision F, dated February 6, 1990, required in Paragraph 

5.1 of Section 5.0, "Requirements," that compliance with 10 CFR Part 21 shall 

be mandatory on all shop orders and all purchase orders to suppliers when 

imposed by the customer purchase order. 

During the NRC's review of safety-related customer purchase orders to CDC, 

discussed in Section 3.7 of this report, it; was noted that in all cases the 

customer purchase orders referenced that the provisions of 10 CFR Part 21 

applied and that any material or items spec:ified in the purchase order be 

suppl i ed and cert i f ied in accordance wi th CDC' s Qual i ty Assurance Manual . Paragraph 1.4 of Section 1.0, "Introduction," of the CDC Quality Assurance 

Manual stated compl i ance to several recognized qual i ty standards and 

specifications, including that of 10 CFR Part 50, Appendix B, and 10 CFR 

Part 21. 

The NRC inspection team reviewed purchase orders to the suppliers identified 

above in Section 3.7 which had furnished material or services used by CDC in 

the manufacture of safety-related rupture discs or rupture disc holders. 

suppl iers i ncl uded Tel edyne Rodney, Cast1 e Metal s , Joseph T. Ryerson & Son, 

The 

12 
Coulter Steel and Forge and Metal Goods. 

identified that the quality requirements imposed on CDC (10 CFR Part 50, 

Appendix B) in licensee purchase orders had not been passed down by CDC to its 

suppl iers of safety-re1 ated materi a1 s and services . purchase orders to the above suppliers were reviewed: 

In a17 cases the NRC review 

The fol 1 owi ng CDC 

0 CDC purchase order No. 49912 to Teledyne Rodney, dated May 18, 1995, 

ordered one hundred pounds of fully annealed, cold rolled, 316L 

stainless steel coil material. The material specified was 0.010-inches 

thick and 24-inches wide which was required to conform to the American 

Society for Testing and Materials (ASTM) ASTM A240 material 

specification. 

0 CDC purchase order No. 49874 to Castle Metals, dated May 16, 1995, 

ordered 144 inches of 7/8-inch, 304 stainless steel hex bar stock which 

was required to conform to ASTM A479 material specification. 

0 CDC purchase order No. 47600 to Joseph T. Ryerson, dated October 26, 

1994, ordered sixteen pieces of l-inch round bar stock which was 

required to conform to ASTM A479 material specification. 

0 CDC purchase order No. 47107 to Coulter Steel & Forge, dated September 

20, 1994, ordered 12-inches of 3 1/4-inch round bar stock which was 

required to conform to ASME SB 160-200 nickel, minimum yield/tensile 

strength of 35,000/60,000 pounds per square inch, respectively. 

0 CDC purchase order No. 49909 to Metal Goods, dated May 18, 1995, ordered 

two pieces of 1 3/8-inch, 304 stainless steel round bar stock which was 

required to conform to ASTM A479 material specification. 

Failure of CDC to impose customer required 10 CFR Part 50, Appendix B, quality 

requirements and in purchase orders to its suppliers furnishing materials or 

services used in the manufacture of nuclear safety-re1 ated ruptiire discs and 

rupture disc holders has been identified as Nonconformance 95-01-bj. 

In addition, CDC Quality Control Instruction No. 1003 required that suppliers 

pass down all purchase order requirements to any authorized sub-tier suppliers 

and Quality Control Instruction No. 1020 required that compliance with 10 CFR 

Part 21 be mandatory on all shop orders and a1 1 purchase orders to suppl iers 

when imposed by the customer purchase order. All of the customer purchase 

orders from licensees discussed in Section 3.7 specified that the requirements 

of 10 CFR Part 21 applied, however, CDC did not impose the requirements of 

10 CFR Part 21 in its purchase orders to the suppliers which furnished 

material or services (as discussed previously in this section). Since CDC had 

not specified 10 CFR Part 50, Appendix B, on the purchase orders to the 

suppliers the requirements of 10 CFR Part 21 were not required (by regulation) 

to be included. However, CDC had placed a procedural restriction on itself 

which did require the inclusion of 10 CFR Part 21. As written, CDC’s 

procedures did not allow them to purchase material or services as commercial 

grade (without passing down 10 CFR Part 21) and to take additional actions 

13 

18 
which would enable them to supply or use the material or services as safetyrelated. 

aware that if the requirements of 10 CFR Part 50, Appendix 8, had been 

specified on CDC’s purchase orders to suppliers that the requirements of 

10 CFR Part 21 would have also been required to be specified by regulation. 

Although CDC’s actions were ultimately consistent with 10 CFR Part 50, 

Appendix B, (not specifying 10 CFR Part 50, Appendix B, and therefore not 

specifying 10 CFR Part 21) the actions were not in accordance with the 10 CFR 

Part 21 requirements specified in Quality Control Instructions No. 1003 and 

No. 1020. CDC’s failure to follow the procedural requirements related to 

10 CFR Part 21, as specified in Quality Control Instructions No. 1003 and 

No. 1020, when purchasing material and services used in the manufacture of 

nuclear safety-related rupture discs and rupture disc holders, has been 

identified as a violation of 10 CFR 21.31, which requires that purchase orders 

to suppliers specify that the provision of 10 CFR Part 21 apply when 

applicable (in this case, when required by CDC Quality Control Instructions 

No. 1003 and No. 1020). This failure constitutes a violation of minor 

significance and is being treated as a Non-Cited violation, consistent with 

Section IV of the NRC Enforcement policy (NUREG-1600). 

The inspectors concluded, based on discussion with CDC, that CDC was 

4 PERSONS CONTACTED 

The NRC staff participating in the inspection and CDC personnel contacted 

during the inspection are listed below. 

Continental Disc Corporation 

Kenneth R. Shaw, President 

*#Dean Dachenhausen, Director Quality Assurance 

U.S. Nuclear Regulatory Commission 

*#Bi 11 Rogers *#Robert Ppt +. i s 

#Gregory Cwal i na 

Team Leader, VIS/PSIB 

Senior Reactor Engineer, VIS/PSIB 

Section Chief, VIS 

*Attended the Entrance Meeting #Attended the Exit Meeting 

14 

19 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 205E4l001 

July 7, 1995 

Ms. S. Kay Fisher 

Manager, Quality Assurance 

Divesco, Inc. 

5000 Highway 80 East 

Jackson, Mississippi 39208 

SUBJECT: NRC lNSPECTION NO. 99901117/95-01 

Dear Ms. Fisher: 

I 

This letter transmits the report of the U.S. Nuclear ,,zgulatory Commission 

(NRC) inspection of Divesco, Incorporated, conducted on February 8 and 9, 

1995, by Mr. Stephen Alexander of this office. The inspection was conducted 

to provide a basis for assessing the validity and completeness of the list 

that you provided to General Electric Nuclear Energy (GE NE) of items from the 

American Heavy Trading Black Fox inventory that you supplied to GE NE. 

inspection was also conducted to determine from your records the disposition 

of the remainder of the Black Fox inventory items, including those supplied to 

D-Tech (formerly OMTECH, Inc., and TEMCO, Inc.). 

The 

Areas examined during the inspection and our findings are discussed in detail 

in the enclosed report. 

selective examination of procedures and representative records, review of 

technical documentation, interviews with personnel , and observations by the 

inspectors. The major areas reviewed included (1) implementation of your 

quality assurance (QA) program based on Appendix 8 to Part 50 of Title 10 of 

the Code of Federal Requlations (10 CFR Part 50, Appendix B) with respect to 

quality assurance records and (2) implementation of your program for reporting 

of defects and noncompliance pursuant to 10 CFR Part 21. 

Within these areas, the inspection consisted of 

Based on the results of this inspection, one part of your 10 CFR Part 21 

implementation program appeared to be in violation of NRC requirements, as 

specified in the enclosed Notice of Violation (Notice). 

10 CFR Part 21 is related to your procedures adopted pursuant to the 

regulation. However, the inspector found no instances in which your other 

practices or records were not in accordance with 10 CFR Part 21; nor did the 

inspector identify any instances in which potential Part 21 issues were not 

properly dispositioned. 

requirements are identified in the enclosed Notice and inspection report. 

The violation of 

The specific findings and references to the pertinent 

You are required to respond to this letter and should follow the instructions 

specified in the enclosed Notice when preparing your response. 

response, you s:,culd document the specific actions taken and any additional 

actions you plan to prevent recurrence. After reviewing your response to the 

Notice, including your proposed corrective actions and the results of future 

inspections, the NRC will determine whether further NRC enforcement action is 

necessary to ensure compliance with NRC regulatory requirements. 

In your 

20 
S. Kay Fisher -2- 

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

this letter and its enclosures will be placed in the NRC Public Document Room. 

The responses requested by this letter and the enclosed notice are not subject 

to the clearance procedures of the Office off Management and Budget as required 

by the Paperwork Reduction Act of 1980, Public Law No. 96-511. 

Should you have any questions concerning this report, 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 Technical Support 

Office of Nucl ear Reactor Regul at i on 

Docket No. 99901117 

Enclosures : 1. Notice of Vi ol at i on 

2. Report No. 99901117/95-01 

cc w/encl: Mr. Forest Hatch, Manager S&P Quality 

GE Nuclear Energy 

175 Curtner Avenue 

San Jose, CA 95125 

Mr. Dick Tettman, President 

D-Tech, Inc. 

15040 10s Gatos Boulevard 

Los Gatos, CA 95032 

21 
NOTICE OF VIOLATION 

DIVESCO, Incorporated 

Jackson, Mississippi 

Docket No. 99901117 

Report No. 95-01 

During an NRC inspection conducted February 8 and 9, 1995, a violation of NRC 

requirements was identified. In accordance with the "General Statement of 

Policy and Procedure for NRC Enforcement Actions," 10 CFR Part 2, Appendix C 

(1992), the violation is listed below: 

Section 21.21(a) of Part 21, "Reporting of Defects and Noncompliance," of 

Title 10 of the Code of Federal Requlations (10 CF8 Part 21) requires, in 

part, that each individual, corporation, or entity subject to the regulations 

in this part adopt appropriate procedures to ensure the proper evaluation of 

deviations and failures to comply and the reporting of defects and failures to 

comply related to a substantial safety hazard to a director or responsible 

officer. Part 21 requires that (1) the evaluation of deviations and failures 

to comply in delivered basic components must be completed within 60 days of 

discovery, (2) reports to a director or responsible officer of defects and 

failures to comply related to a substantial safety hazard must be made within 

5 working days of completion of the evaluation, and (3) an interim report must 

be made to the NRC within 60 days of discovery of the deviation or failure to 

comply if the evaluation cannot be completed within the required time. 

Section 21.21(b) requires that when a supplier determines that it is not 

capable of evaluating the deviation or failure to comply, then it must notify 

affected licensees or purchasers of the deviation or failure to comply within 

5 working days of making this determination. 

Contrary to the above, as of February 9, 1995, the effective revision of 

Divesco's Nuclear Quality Assurance Manual, Procedure No. 10, Revision 00, 

dated June 25, 1987, "Part 21 Evaluation and Notification," which constituted 

the Divesco procedures adopted pursuant to 10 CFR 21.21, would not, as 

written, ensure evaluation and reporting in accordance with the regulation as 

follows: 

if the procedure were established to follow 421.21(b), yet contrary to 

§21.21(b), the procedure required evaluation of those deviations to determine 

safety impact prior to, or as a prerequisite for, customer notification as 

would otherwise be performed under §21.21(a). However, having called for an 

evaluation of the type required by §21.21(a), the procedure did not provide 

for notification of a director or responsible officer within 5 working days 

should the evaluation identify a defect or failure to comply associated with a 

substantial safety hazard as required by 921.21(a). The procedure also did 

not contain the interim reporting requirement and time limit provisions added 

by the version of the regulaticn that became effective on October 29, 1991. 

The procedure called for notification to purchasers of dwiations as 

(95-01-01) 

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

Enclosure 1 
Pursuant to the provisions of 10 CFR 2.201, Divesco is hereby required to 

submit a written statement or explanation to the U.S. Nuclear Regulatory 

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

to the Chief, Special Inspection Branch, Division of Technical Support, Office 

of Nuclear Reactor Regulation (Mail Stop: O-gAI), within 30 days of the date 

of the letter transmitting this Notice of Violation (Notice). 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. Where good cause is shown, consideration will be given to extending 

the response time. Under the authority of Section 182 of the Act, 42 U.S.C. 

2232, this response shall be submitted under oath or affirmation. 

Dated at Rockvi 11 e, Mhryl and 

this 7th day of July, 1995 

-2- 

23 
REPORT NO.: 

ORGANIZATION: 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY: 

INSPECTION DATES: 

INSPECTOR: 

REVIEWED BY: 

APPROVED BY: 

U. S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF TECHNICAL SUPPORT 

99901117/95-01 

Divesco, Incorporated 

5000 Highway 80 East 

Jackson, Mississippi 39208 

S. Kay Fisher 

Manager, Quality Assurance 

Divesco (Formerly NSSS-Divesco) supplies (directly or 

as a consignment agent) surplus equipment, components. 

and piece parts (obtained primarily from cancelled 

nuclear plant projects) to the nuclear industry. 

February 8 and 9, 1995 

.- ~ - 

Vendor Inspection Section (VIS) 

Special Inspection Branch (TSIB) 

C;regory,,C./Cwalina, Chief, VIS/TSIB 

n 

& w .c ?IEnclosure 2 

24 
1.0 SUMMARY OF INSPECTION FINDINGS 

The inspection was conducted to provide a basis for assessing the validity and 

completeness of the Divesco list provided at NRC request to General Electric 

Nuclear Energy (GE NE) of items sold to GE NE from the American Heavy Trading, 

Incorporated (AHT), consignment inventory at Divesco. This inventory, on 

consignment with Divesco as agent for AHT from 1985 to 1989, consisted of 

material purchased by AHT from Pub1 ic Service of Oklahoma's (PSO's) cancelled 

Black Fox Nuclear Station (Black Fox). 

determine from Divesco records the disposition of the remainder of the AHT 

consignment Black Fox inventory items, including those supplied to another 

surplus material dealer called D-Tech (formerly TEMCO, or OMTECH, Inc.) in Los 

Gatos, California, or to others, if any. During this inspection, the NRC 

inspector reviewed Divesco records and evaluated the Divesco system of record 

keeping to accompl sh the above objectives. 

The inspection was also conducted to 

The inspection bas s consisted of the following: 

Appendix 3, "Quality Assurance Criteria for Nuclear Power Plants and 

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

Requlations (10 CFR Part 50, Appendix B) 

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

Divesco Nuclear Quality Assurance Manual (NQAM), Procedure No. 10, 

Revision 00, dated June 25, 1987, "Part 21 Evaluation and Notification" 

1.1 Violation (95-01-01) Contrary to the requirements of 10 CFR 21.21, 

Divesco procedures adopted pursuant to the regulation (1) called for 

evaluating deviations to determine safety impact before notifying affected 

purchasers or licensees, but did not provide for notifying a director or 

responsible officer should the evaluation identify defects or failures to 

comply associated with a substantial safety hazard, and (2) did not contain 

certain reporting provisions and time limits required by the version of the 

regulation that became effective on October 29, 1991 (see SZ..~,,, 3.4 of this 

report). 

1.2 Nonconformances 

None 

2.0 STATUS OF PREVIOUS INSPECTION FINDINGS 

No previous findings were reviewed during this inspection. 

3.0 INSPECTION FINDINGS AND OTHER COMMENTS 

3.1 Backqround 

During an inspec 

NRC requested GE 

994, the 

for NRC 

review 'on the procurement, hand1 ing, and disposition of equipment, components, 

ion of GE NE in San Jose. California, in April of 

NE to research its records and Drovide informatian 

2 

25 
and parts belor,ging to the consignment to Divesco by AHT. 

to the NRC a list of seven procurements of material said to be traceable to 

the AHT Black Fox consignment through Divesco. Table 1 of Appendix A to this 

report lists these seven procurements. During a subsequent inspection at GE 

NE in January 1995, GE NE explained that it did not have its surplus material 

procurement records computerized or organized in a manner conducive to 

efficient searches of the kind requested by the NRC. 

requested Divesco to provide the information. During the January 1995 

inspection, the inspector reviewed the information on the handling and 

disposition of material on the list of the seven procurements of AHT/Black FOX 

material from Divesco and found that it appeared to have been procured, 

inspected, and supplied in accordance with GE NE procedures which the NRC has 

extensively examined in the past (refer to NRC Inspection Report Nos. 

99900403/ 89-01, 90-01, and 94-02 or to 1989, 1990 and 1994 volumes of 

NUREG-0040). No discrepancies were noted with the handling or disposition of 

the material in the Divesco list of seven procurements during this February 

1995 NRC inspection at Divesco. 

GE NE transmitted 

Instead, GE NE had 

During the January 1995 inspection at GE NE, the inspector learned that GE NE 

had not conducted an independent search of its own records as expected. 

Therefore, the inspector searched the records at GE NE and identified 

procurements of material from Divesco that were not listed on Divesco’s list. 

These procurements are listed in Table 2 of Appendix A to this report. The 

inspector also identified several procurements from another surplus equipment 

dealer called OMTECH that had had some dealings with Divesco and AHT. This 

company was later called TEMCO and now is called D-Tech. The source of the 

material was not evident from these records, but it was possibly the AHT Black 

Fox consignment in question. Accordingly, this information was pursued at 

Divesco during this February 1995 inspection and also at D-Tech in March 1995. 

3.2 Entrance and Exit Meetinqs 

During the entrance meeting, the inspector met with the Divesco Quality 

Assurance (QA) Manager and discussed the scope of the inspection. During the 

exit meeting on February 9, 1995, with the Divesco Quality Assurcl;ice Vanager, 

the inspector summarized the inspection findings. 

3.3 Procurement/Receivins/Sales Records 

The Divesco database (for sales in 1986 and later) contained all the sales to 

GE NE that Divesco had identified to GE NE following the April 1994 NRC 

inspection of GE NE (Table 1, Appendix A). Among the GE NE purchase orders 

(POs) to Divesco from 1986 and later (selected from GE NE files in January 

1995) that had not been identified to GE NE by Divesco as having been filled 

from the AH1 Black Fox consignment inventory (Table 2, Appendix A), one item, 

an Agastat relay, purchased on GE NE PO 205-86R686, possibly came from 

material originally supplied to PSO for Black Fox by GE NE, but not from the 

material included in the AHT Black Fox consignment. 

procurement, the type of item, and the way it was identified in material 

receiving reports suggested that it was from another of Divesco’s sources, but 

it was examinea as if it had been part of the AHT consignment. 

The time frame of this 

3 

26 
Also among the POs listed in Table 2 of Appendix A were two other items 

purchased by GE NE in 1985 that were from the AHT Black Fox consignment. 

These were two pressure transducers for Perry (PO 205-85E949) and a hydraulic 

hand pump for Clinton (PO 205-855769). The Divesco QA Manager explained that 

Divesco had not included any of the 1985 sales of Black Fox material in its 

original list to GE NE (Table 1, Appendix A) because Divesco had overlooked 

the fact that some of the pertinent information might be in its separate 

database containing records of 1985 sales only. One additional procurement 

identified among GE NE records, some terminal boards for Perry (PO 205- 

85E813), was not listed in the Divesco database. It was also not found during 

a review of Oivesco paper sales records. Divesco has not yet traced these 

items to the AHT Black Fox consignment. The rest of the GE NE procurements in 

Table 2, Appendix A, selected at GE NE for review at Divesco were traceable to 

non-Black Fox sources, i .e., TVA, Chism Company, or GE All ens Creek. 

Finally, review of the Divesco 1985 sales database revealed two procurements 

by GE NE of two control cards (P/N 204B72156001) for Perry (PO 205-85N643) and 

three GE SBM-type switch handles for Clinton, PO 205-85N91 

number). According to Divesco Invoice 811225-1, 1 isted in-the Divesco 

database for 1985 sales for this line item, the referenced GE PO number was 

205-85N911, but this PO, obtained from GE NE, is for ASCO solenoid-operated 

valves and the PO was issued to "OMTECH," now D-Tech. 

the control cards was GE Allens Creek and the source of the SBM handles was 

listed as GE NE itself, these two procurements were not found among the 

surplus material procurement records at GE NE. 

to resolve this apparent discrepancy. 

(truncated PO 

Although the source of 

Divesco is working with GE NE 

With regard to the remainder of the AHT Black Fox consignment, the inspector 

also reviewed Divesco's records of all sales of AHT Black Fox consignment 

material to identify any other parties to whom the material may have been 

sold. According to Divesco records, there were several procurements by 

various utilities directly from Divesco of material from the AHT Black Fox 

consignment: 8 in 1985 and 31 more from 1986 on. Divesco certified to meeting 

the PO, handling and storage in accordance with applicable QA requirements, 

and traceability to the Black Fox consignment. 

identified in the handling or disposition of this material. NRC licensees are 

responsible for review of the material for suitability of application and 

verification that it meets applicable requirements. 

No deficiencies were 

The only other party with whom Divesco's records indicate it had dealings in 

B1 ack Fox-traceabl e materi a1 was TEMCO (now D-Tech) . The inspector reviewed 

Divesco's records of transactions with TEMCO to determine which, if any, of 

the items involved came from the AHT Black Fox consignment inventory. Divesco 

explained that the material transactions between Divesco and TEMCO (listed as 

"sales" in the Divesco database) were not outright sales, but transfer of 

consignment goods, owned by TEMCO and warehoused by Divesco, for which Divesco 

would receive commission. This information was later confirmed by the 

inspector during the March 1995 inspection at D-Tech. Divesco's database does 

not list any such transfers to TEMCO in 1985. 

TEMCO from December 1987 through the last of them in 1992 lislcJ in the 

Divesco database, all the material came frcrrn non-AHT/Black Fox ,uppliers 

(i.e., GE Allens Creek, GE NE, TVA (IRP), TVA/Chism, or TEMCO itself). 

Of the numerous transfers to 

Sales 

4 

27 
to TEMCO in 1986 were not listed in the Divesco database, although Divesco had 

paper records of them and was in the process of entering them into the 

database. This data was also reviewed by the inspector to confirm the source 

and disposition of the material during the inspection at D-Tech in March 1995. 

In addition, the database showed several procurements by GE NE from Divesco 

(one in 1986 and three in 1988) for which Divesco records showed TEMCO as the 

supplier. During the inspection at D-Tech in March 1995, all the material 

identified as being possibly traceable to Black Fox was pursued to determine 

or confirm its origin and its disposition by D-Tech. 

The inspector also reviewed Divesco records of sales to TEMCO to determine if 

Divesco (and ultimately Black Fox) was the source for any of the material 

supplied by TEMCO to GE NE which had been identified by the inspector at GE NE 

in January 1995 during the record search that also identified the procurement 

from Divesco listed in Table 2 of Appendix A. This search in January 1995 had 

identified nine procurements by GE NE from TEMCO (all 1986 to 1987). None of 

these were listed in Divesco records as of February 1995, suggesting TEMCO’s 

source was other than Divesco for this material. During the March 1995 

inspection at D-Tech, this was confirmed except for one procurement. 

205-86R630 to TEMCO for a Rosemount temperature element (S/N 40249) was 

procured from Divesco in 1987. According to D-Tech records reviewed by the 

inspector in March 1995, this item came from Divesco in August 1987, but it 

did not appear in the Divesco computer database record of sales to TEMCO, 

which went back only to December 1987. The transfer of this item from Divesco 

to TEMCO was later reported to the NRC as confirmed by Divesco after review of 

its paper records. 

GE NE PO 

In 1987, Divesco and AHT dissolved their consignment/joinl venture agreement. 

However, Divesco retained the inventory for the time being and sold several 

items to GE NE and some items directly to nuclear utilities from time to time 

under special agreements in each case with AHT. In January 1988, Divesco 

purchased several items from the Black Fox consignment inventory outright 

(listed in a bill of sale, dated January 24, 1988, and signed by the president 

of AHT). The inspector reviewed the records of items from that purchase that 

have been sold. Finally, according to a receipt and release document on file 

with Divesco, dated July 25, 1989, signed by AHT’s president, AHT acknowledged 

receipt of the remainder of its consignment inventory, which was to be removed 

from Divesco’s site within 30 days. Divesco stated that the material known to 

belong to AHT was removed by AHT. 

To summarize, the material in question is all surplus material from PSO’s 

cancelled Black Fox project. Material from Black Fox was sold to both GE NE 

and AHT. Some of the items sold by PSO directly to GE NE were sold by GE NE 

directly to NRC licensees. Others were sold by GE NE to TEMCO, which placed 

those items, along with other surplus material from GE NE, in storage at 

Divesco and one other location. TEMCO sold some of this material directly to 

NRC licensees, and some to GE NE for sale to NRC licensees. In either case, 

TEMCO would get the material back from Divesco (or one other warehouse of its 

own) and ship !t (or have it shipped direct) to its (TEMCO’s) customer (either 

GE NE or a utif:ty). 

material was commingled with material from the AHT Black Fox consignment. 

The inspector found no evidence that any of this 

5 

28 
The Black fox material purchased by AHT from PSO was placed on consignment 

with Divesco in accordance with a joint venture agreement between AHT and 

Divesco in 1985 and shipped directly from the Black Fox site to Divesco’s 

warehouse. 

consignment to GE NE and some directly to utilities as agent for AHT. As 

discussed above, in 1989, AHT removed what both parties agreed to constitute 

the remainder of AHT’s Black Fox consignment inventory from the Pivesco site 

(except for certain items bought outright by Divesco from AHT). In telephone 

conversations with the inspector, AHT has stated that AHT has offered various 

items of the material it removed from the Divesco site to several utilities on 

the basis that it be used only for training aids or other non-safety-related 

purposes. The NRC has not confirmed the disposition of the material returned 

to AHT except that the inspector has seen photographs said to be of this 

material lying in an open field reportedly located somewhere in the Jackson, 

Mississippi area. 

Over the next few years, Divesco sold some of the AHT Black fox 

The NRC has received no substantive evidence that any of this material is, or 

has been, commingled with used, fraudulent, or refurbished material. The NRC 

also has no evidence that any of it is substandard. However, the material 

returned to AHT has not been maintained in 10 CFR Part 50, Appendix B, QA 

storage, and has apparently suffered some degradation from exposure to the 

elements. 

As discussed above, the NRC has inspected GE NE’S and Divesco’s QA programs 

(including procurement), particularly scruLinizing their handling and 

disposition of the material in question. Neither Divesco nor D-Tech has 

certified as to the quality or suitability of the material for any plant 

applications other than that its condition has been maintained and that it is 

the material specified in customer POs, with records of traceability to the 

Black fox consignment. GE NE has certified to the quality of the material in 

those instances in which the material has been determined to be traceable to 

material and QA records originally supplied to Black Fox and on the basis of 

receiving inspection or, as was the case with the MSIVs for NMP2, based on 

recertification by the original manufacturer. 

Licensee procurement programs are required under 10 CFR Part 50, Appendix B, 

to determine whether any items procured from any supplier, including GE NE, 

TEMCO, Divesco, or another utility, are suitable and of adequate quality, 

condition, and reliability for their plant applications. In addition, 10 CFR 

Part 50, Appendix B, requires licensees to establish measures to control 

nonconforming material to prevent its inadvertent i n s t a l l a t i o n o r use. 

Further, 10 CFR Part 50, Appendix B, requires licensees to establish measures 

to detect and correct conditions adverse to quality. 

requires any entity that supplies a basic component to an NRC-licensed 

facility to evaluate any deviation or failure to comply in that basic 

component of which they become aware and to report it either to the NRC or to 

affected licensees or purchasers. 

industry groups to improve the quality of industry procurement practices, the 

NRC has inspected the procurement programs of numerous licensees, including 

some of those Litat have procured some of this material. 

Finally 10 CFR Part 21 

In addition to working with licensees and 

6 

29 
Should the NRC become aware of any substantive evidence that any of the 

material in question may be in some way substandard, defective, unreliable, or 

otherwise unsuitable for service in a NRC-licensed facility, the NRC would 

follow up on the specific information in a manner appropriate to the 

circumstances. Absent such information or evidence, this matter is considered 

closed. 

In order to investigate the concerns raised by AHT that material from its 

Black Fox inventory was being sold to NRC licensees without traceability and 

hence was of indeterminate quality, the inspector needed first to establish 

the disposition of this material. 

inspector concluded that the disposition of the material in the Black Fox 

inventory has been determined as far as Divesco is concerned. 

Based on review of Divesco records, the 

3.4 10 CFR Part 21 ImDlementation 

The inspector reviewed Divesco Nuclear Quality Assurance Manual (NQAM), 

Procedure No. 10, Revision 00, dated June 25, 1987, "Part 21 Evaluation and 

Notification," and made the following observations: 

The stated purpose (Paragraph 1.0) of the procedure was to "identify the 

requirements for evaluating deviations for potential safety impact and 

informing the purchaser of the deviation to satisfy the requirements of 

10CFR21." The inspector noted that although evaluating deviations was 

addressed, notification of the NRC in accordance with 10 CFR 21.21 was 

not, nor was evaluation of failures to comply as defined in Part 21. 

Evaluation of deviations and reporting to the NRC were addressed in a 

posted company policy statement. The inspector noted that both the 

procedure language and the posted policy statement were inconsistent 

with Divesco's practice, as described by the Vice President and the QA 

Manager, of not performing the 10 CFR 21.21(a) evaluations, but of 

simply informing affected licensees or purchasers of all deviations (and 

failures to comply) involving basic components supplied by Divesco of 

which Divesco becomes aware as provided by §21.21(b). 

Paragraph 2.0, "Applicability," stated that the Divesco program was 

limited to the evaluation of deviations (identified by Divesco or 

reported to Divesco) to determine if a potential safety problem exists 

and to the notification of the purchaser so that a 10 CFR Part 21 

evaluation can be performed. Here again, the procedure described the 

Divesco practice as if it were carried out pursuant to 10 CFR 21.21(b), 

yet provided for an evaluation for safety impact as a prerequisite for 

informing customers. Section 21.21(b) is applicable when the basic 

component supplier has determined that an evaluation will not be 

performed. The language of this paragraph in the Divesco procedure 

mandated an evaluation pursuant to 10 CFR 21.21(a), yet there were no 

subsequent provisions for notifying a director or responsible officer 

who would then effect NRC notification. 

Paragraph 4.3, under "Requirements" (Paragraph 4.0), called for. review 

of deviations (but did not include failures to comply) for "potential 

impact on safety." Paragraph 4.4 stated the requirement for "followup 

7 

30 
notification to purchaser of a defect." In the case that Divesco elects 

to perform an evaluation and determines that a defect does exist or that 

a failure to comply could create a substantial safety hazard, Divesco 

procedures would then be required by §21.21(a) to provide for 

notification of a Divesco director or responsible officer (who, as 

stated above would then effect notification of the NRC), yet the 

procedure contained no such provisions, only purchaser notification. 

Divesco confirmed that it would not normally be in a position to know 

either the intended plant application of its supplied basic component or 

the impact of any deviation or failure to comply on the plant, system or 

parent component safety function. However, this practice was not 

consistent with the language of the procedure. 

that nothing in the regulation should be construed to discourage or 

prohibit reporting under Part 21. However, unless Divesco is fully 

qualified to determine that a given dleviation does not constitute a 

defect as defined in 921.3 or that a given failure to comply could not 

create a substantial safety hazard, then, contrary to the language of 

the Divesco procedure, the performance of the evaluation and the 

determination that a so-called potential safety impact exists should not 

be a prerequisite to informing all affected licensees or purchasers as 

required by 10 CFR Part 21.21(b). 

The inspector explained 

(4) The procedure did not contain the additional provisions and new time 

limits promulgated in the version of 10 CFR Part 21 that became 

effective on October 29, 1991. According to the minutes of a Divesco QA 

meeting held on November 21, 1991, Divesco had reviewed what it believed 

to be the current revision of Part 21 because this version, dated 

October 31, 1989, was included as Attachment 1 to NRC Information 

Notice 91-31, issued June 17, 1991. This was also the version of the 

regulation posted pursuant to §21.6(a). However, the revision of the 

regulation containing the new provisions and time limits was first 

published in the Federal Register (56FR 36081) on July 31, 1991. It was 

then announced in NRC Information Notice 91-76, "10 CFR Parts 21 and 

50.55(e) Final Rules," dated November 26, 1991. The inspector provided 

Divesco with a copy of the current revision of 10 CFR Part 21. 

(5) Paragraph 5.2 stated, in part: "...procurement documents shall specify, 

when applicable, that 10CFR21 [sic] requirements are imposed on the 

supplier." It then stated: "When the source of the material is an 

ex-licensee, 10CFR21 [sic] provisions are not imposed." However, this 

statement in the procedure is inconsistent with 921.31. If the material 

to be procured is a basic component, then 921.31 requires invoking 

Part 21 in procurement documents, regardless of the status of the 

supplier. Although the language of Paragraph 5.2, as written, would 

allow violation of 921.31, the inspector did not identify any instances 

(within the restricted scope of this inspection) in which Divesco had 

failed to comply with 921.31. 

Upon completion of the review of Divesco NQAM Procedure No. 10, the inspector 

concluded that it woltld not, as written, ensure proper evaluation and 

reporting, if required, of deviations or failures to comply in accordance with 

8 

31 
10 CFR 21.21 in that it (1) called for evaluation of deviations to determine 

safety impact prior to notification to purchasers of deviations or failure to 

comply, yet did not provide for NRC notification, and (2) did not contain 

reporting provisions and time limits required by the current version of the 

regulation. These deficiencies were cited as Violation 99901117/95-01-01. 

4 PERSONS CONTACTED 

Divesco, Incorporated: 

Westbrook, T. 

Fisher, S. Kay 

Vice President 

Manager, Quality Assurance 

9 

32 
APPENDIX A 

TABLE 1: 

GE NE PO# DATE QTY ITEM DRAW I NG GE NE CUSTOMER SOURCE 

205-86K317 04/28/86 1 FC valve 112D1459P001 Perry (CEI) AHT/BF 

205-863142 05/14/86 2 MSIV blowers 213A3762P001 C1 inton (IP) AHT/BF 

205-86R638 10/02/86 1 Temp Element 159C4520P005 River Bend( GSU) AHT/BF 

205-86R687 10/06/86 8 MSIVs B21-IF021 - 28 NMP2 (NMPC) AHT/BF 

205-86R689 10/06/86 1 MSIV Blower 47B518664 NMP2 (NMPC) AHT/BF 

205-86R874 11/17/86 2 Relief Valves 21A9508P001 Hatch (GP&L) AHT/BF 

205-87C630 03/13/87 1 Temp Element 159C4520P005 C1 inton (IP) AHT/BF 

DIVESCO LIST OF SALES OF AHT/BLACK FOX ITEM TO GE NE (1986 ON) 

TABLE 2: 6E NE PURCHASES FROH DIVESCO FROH NRC SEARCH OF GE NE RECORDS 

DRAW 1 NG GE NE CUSTOMER SOURCE 

205-85E813" unkn unkn Terminal Bd 14707614G004,5 Perry (CEI) *AHT/BF 

*Not listed on Divesco's 1985 sales records database 

GE NE PO# DATE QTY ITEM -- 

205-85E949 07/05/85 2 Press XDCR MPL : C85N001 Perry (CEI) AHT/BF 

205-853769 11/25/85 1 Hyd Hand Pump 131C8966G001 Clinton (IP) AHT/BF 

205-86R686 10/03/86 1" Agastat 145C3217P041 River Bend *AHT/BF 

(3 of 4 rtnd, 4th: *S/N 77231248 retained, baring GE DWG #, \/!:, "gastat part 

number, E7024PB002, as were the two, S/Ns 85170022,3 from Control Components. 

Therefore, the one finally kept by GE could have been from AHTJBF) 

205-85N648 09/12/85 58 Anlg Is01 204B6220AAG002 Perry (CEI) Allens Ck 

205-87C632 02/03/87 4 CKT Cards 272A8614P101,02,12,20 RB(GSU) Chi sm 

I 

33 
UNITED STATES 

NUCLEAR REGULATORY COMMlSSlOrd 

WASHINGTON, D.C. 20555-OOOl 

August 29, 1995 

Mr. K.J. Cumnings, Plant Manager 

Eaton Corporation 

9 South Street 

Danbury, CT 06810 

SUBJECT: NRC INSPECTION REPORT 99901290/95-01 

Dear Mr. Cummings: 

This letter transmits the report of the U.S. Nuclear Regulatory Commission 

(NRC) inspection of Eaton Corporation (Eaton) , Danbury, Connecticut, conducted 

by Messrs. K.R. Naidu and I. Ahmed on August 8-10, 1995. The inspection was 

conducted to provide a basis for NRC staff confidence that the components 

manufactured by Eaton to upgrade the existing engineered safeguards actuation 

system (ESAS) for Northeast Nuclear Energy Company's Mi 11 stone Nuclear Power 

Station, Unit 2 (MP-2), would perform their intended safety functions. On 

September 10, 1995, at the conclusion of the inspection, the inspectors 

discussed their findings with you and other members of your staff. 

During this inspection, the team evaluated the Eaton quality assurance program 

that was established to implement the provisions of Part 50 of Title 10 of the 

Code of Federal Rewlations (10 CFR Part 50), Appendix B, and the provisions 

of 10 CFR Part 21, in selected areas during the design, manufacture, and 

installation of the ESAS upgrade for MP-2. Within these areas, the NRC team 

(a) examined technical documentation, procedures and representative records, 

(b) held discussions, (c) listened to presentations and (d) observed Eaton 

technicians' activities. 

During the c:'idltic:cion of your activities at Danbury, the inspectors noted the 

proactive approach of your staff to acknowledge weaknesses in the existing 

quality program and willingness to correct them. 

the employees who were interviewed during the inspection exhibited good 

technical expertise and positive attitudes. 

The inspectors noted that 

The procedure adopted by you to implement 10 CFR Part 21, which was developed 

in 1978 by Consolidated Controls Corporation, was last revised in 1981 and 

failed to meet the current requirements. 

of minor significance and is being treated as a Non-cited Violation, 

consistent with section IV of the NRC Enforcement Policy. 

you are in the process of revising the current 10 CFR Part 21 Procedure to 

reflect the latest regulations, and that you are reorganizing the quality 

assurance manual to reflect the practices of your current organization. 

This failure constitutes a violation 

We understand that 

34 
K. Cummings -2- 

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

this letter and its enclosure will be placed in the NRC Public Document Room. 

No response to this letter or its enclosure is required. Should you have any 

questions concerning this report, we will be pleased to discuss them with you. 

Thank you for your cooperation during this process. 

Si ncerel y , 

Robert M. Gallo, Chief 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nucl ear Reactor Regul at i on 

Docket No.: 99901290 

Enclosure: Inspection Report 99901290/95-01 

35 
REPORT NO. : 

ORGANIZATION: 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY : 

INSPECTION DATES: 

LEAD INSPECTOR: 

OTHER INSPECTORS: 

REVIEWED BY: 

APPROVED BY: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE Of NUCLEAR REACTOR REGULATION 

DIVISION OF INSPECTION AND SUPPORT PROGRAMS 

99901290/95-01 

Eaton Corporation 

9 South Street 

Danbury, Connecticut 06810 

G A. DeRome 

(203) 798-3216 

Instrumentation and control 

systems for safety and nonsafety-re1 ated appl ications. 

August 8-10, 1995 

*.- -/-,---i-ycc 

Kamal akar 'RRaidu. Team Leader 

Vendor Inspection Section (VIS) 

Iqbal. Ahmed, Senior Engineer, 

Instrumentation and Control Branch 

gory C. Cwalina, Chief, VIS 

cia1 Inspection Branch (PSIB) 

Robert M. Gallo. Ch'ldf. PSIB 

%@- 

Division of Inspection-and Support Programs 

Enclosure 

36 
1 SUMMARY OF FINDINGS 

During this inspection, the inspectors evaluated the implementation of the 

qual i ty assurance program adopted by Eaton Corporation (Eaton) in selected 

areas relating to the supply of material and services for upgrading the 

existing engineered safeguards actuation system (ESAS) panel s for Northeast 

Nuclear Energy Company's (NNECO's) Millstone Nuclear Power Station, Unit 2 

(MP-2). 

a 10 CFR Part 21 item. 

The inspectors also reviewed the actions taken by Eaton regarding to 

The inspection basis consisted of the following: 

0 Appendix By "Qual ity Assurance Criteria for Nuclear Power Plants 

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

of Federal Requlations (10 CFR 50, Appendix B) 

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

The inspection identified a violation of minor significance that is being 

treated as a Non-Cited Violation, consistent with Section IV of the NRC 

Enforcement Policy. (Paragraph 3.7.1) 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

This was the first NRC inspection of this vendor. 

3 INSPECTION FINDINGS AND OTHER COMMENTS 

3.1 Entrance and Exit Meetinas 

During the entrance meeting on August 8, 1995, the NRC inspectors discussed 

with Eaton staff the scope of the inspection, areas to be reviewed, and 

established the persons to contact within Eaton management and staff. 

the exit meeting on August 10, 1995, the NRC inspectors summarized their 

findings and concerns to the management and staff of Eaton. 

during this inspection are identified in Section 4. 

During 

Persons contacted 

3.2 Bac kqround Informat i on 

Eaton, formerly Consol idated Controls Corporation, designs, manufactures, and 

provides field services to install new systems and upgrades for instrumentation and control systems for safety and nonsafety-related applications in 

commerci a1 and mi 1 it ary nucl ear power pl ants . Consolidated Controls Corporation, designed the original ESAS panels to 

Specification 76044-480 for NNECO's MP-2. The purpose of the ESAS is to 

continuously monitor the operation of the plant to detect accident conditions 

and to actuate the safeguards systems. 

Consolidated Controls Corporation, and continued to service the equipment 

supplied. 

In the early 1980'~~ Eaton acquired 

Between June 1991 and May 1992, MNECO issued several purchase 

2 
orders (POs) to Eaton to replace all ESAS modules including the sequencer and 

the actuation modules, and the power supplies with an upgrade design. 

instance, the original ESAS design for MP-2, which iqtilized +15-Volt Series 

300 High Noise Immunity Logic, is obsolete and Eaton could not supply the 

spare parts required to maintain it. Eaton submitted a proposal to MP-2 with 

a system upgrade using +&Volt Series 74 HC Complementary Metal Oxide Semiconductor Logic to replace the existing +15-Volt logic. The benefits of the 

upgrade were small er si ze , 1 ower power di ss i pat i on , shorter propagat i on del ay 

periods, and extended service life. During the installation of the upgrade, 

MP-2 experienced a number of unrelated problems including a partial loss of 

normal power. To resolve these problems, NNECO re-evaluated the design 

capabilities, and vulnerability of various ESAS components to electromagnetic 

interference (EMI) and radio frequency interference (RFI). These studies and 

various tests of the ESAS resulted in several modifications to the design, and 

change orders to PQs for the procurement of components and services. 

For 

3.3 Review of NNECO Purchase Orders to Eaton 

NNECO issued six POs to Eaton, including several change orders, for the supply 

of equipment for ESAS. Because Eaton was not an approved vendor of NNECO, 

NNECO took compensatory measures by imposing selected provisions of its 

quality assurance program, including establishing hold points, and conducted 

quality control surveillances to witness hold points and acceptance tests. 

The following table summarizes the NNECO POs: 

- Date 

1973 

June 1991 

May 1992 

May 1992 

July 1992 

July 1993 

July 1993 

NNECO 

- PO 

N/A 

881661 

885480 

886009 

886476 

277176 

278294 

Eaton 

N/A 

Sales Order 

35- 1936 

35-2809 

35-2822 

35-2827 

35-3827 

35-3829 

Brief DeSCriDtiOn 

Or i gi nal contract 

Module upgrade 

Automatic Test 

Insertion (ATI) 

added 

Power supply upgrade 

Field Services to 

install modules 

Field Services and 

Mi scel 1 aneous i tems 

Test rack and additional 

tests to observe impact 

of EM1 and RFI. 

3 

38 
The inspectors reviewed the procurement documents and determined that Eaton 

had prepared detailed proposals to meet the specific technical requirements 

for each item and that NNECO had issued POs based on the proposals. 

3.4 Review of Eaton's Design Review Process 

The Eaton engineering department developed Standard Procedure Instructions 

(SPI) for "Class 35 Power Industry Controls," to implement the provisions of 

the Eaton Quality Assurance Manual , Revision 11, dated February 6, 1995, 

relative to Section 111, "Design Control .I' Eaton engineers followed these 

instructions during the design review of the modules intended for MP-2. 

Eaton's "Design Review Committees" met on several occasions to review the 

adequacy of the preliminary and final designs of the MP-2 modules. Eaton's 

"Design Review Committee" minutes, dated August 28, 1991, indicated that the 

layout and electrical design of the replacement modules (6N636,-37, -38, -40, 

and 5N636, -37, -38, and -40) for MP-2 were reviewed and found acceptable to 

replace the existing modules. A design review meeting was held on December 4, 

1991, to review the final design on bistable module 6N636-1, isolation module 

6N637-1 and block mcldule 6N640-1. The committee also conducted the 

intermediate design review of AT1 module 6M639-1 and actuation module 6N638-1, 

and initial design review of sequencer module 6N641-1 and U/V input module 

6N642- 1. 

The inspectors determined that even though engineers followed the SPIs during 

the design of the MP-2 equipment, the engineers could not readily retrieve the 

necessary documents to demonstrate adherence to SPIs at various stages of 

review and approval. This is an indication of weakness in the implementation 

of the quality assurance program, While acknowledging this weakness, Eaton's 

management assured the inspectors that it will make appropriate enhancements 

to the quality program to ensure that documents generated during the review 

and approval cycles are readily retrievable. 

3.5 Review of NNECO PO No, 277176 

The inspectors reviewed in detail the NNECO PO No. 277176 to Eaton to examine 

the effectiveness of the procurement and installation process. This NNECO PO 

required Eaton to procure and install current/current (I/I) converters, noise 

suppression equipment, and auctioneered power supplies in ESAS Sensor 

Cabinets, and to reconfigure the sump recirculation actuation signal (SRAS) 

logic and AT1 alarm. The inspectors selected two items (I/I converters and 

noise suppression equipment) to verify that Eaton developed design output 

documents, such as the field change procedures and respective drawings as 

required by the PO. 

discrepancies between the design drawings (and the material supplied) and the 

as-built configuration. 

During this review, the inspectors noted two 

The first discrepancy concerned the voltage rating of the 1/1 converter. The 

NNECO PO, which was based on Eaton's Proposal No. "Mar 381," dated May 11, 

1993, specified a 125-Vdc 1/1 converter. 

voltage rating, but Eaton's schematic Drawing No. SGN548-13, Revision By 

identified the input voltage rating of these 1/1 converters to be 18 to 60 

Vdc. The inspectors could not find any documentation either from Eaton or 

Eaton's proposal did not mention the 

4 

39 
from NNECO reconciling this discrepancy in the PO requirement. 

comment in its letter to Eaton dated March 24, 1994, was a request to add the 

NNECO drawing number on the Eaton drawing. 

NNECO’s only 

The second discrepancy concerned the addition of noise suppression devices to 

the coils of all output relays in the ESAS and to the sequencer inputs. 

illustrated the specified noise suppression devices consisting of a series 

combination of voltage-regulating Zener diode and a resistor (part number 

KLK2900-1) in Drawing No. KLK2900-1, and provided the instructions to install 

them in the Field Change Procedure, FCP KRH 136. 

suppressor assembly to be soldered across pins 13 and 14 of each output relay 

socket. Both Eaton and NNECO engineers informed the inspectors that instead 

of this noise suppressor assembly, the as-built configuration consisted of a 

general purpose diode (without the resistor) soldered onto the relays across 

the relay coil instead of being soldered to the socket pins as required by the 

FCP. The inspectors could not find any documentation to indicate that NNECO 

evaluated this deviation from the manufacturer’s design in the installation, 

or that Eaton either acknowledged this change with comments or concurred with 

the change. 

performed the installation in accordance with verbal instructions from NNECO 

personnel. 

Eaton 

FCP KRH 136 required a noise 

Eaton engineers informed the inspectors that their technicians 

The inspectors informed Eaton that lack of formal documentation on the changes 

tc the voltage rating of the 1/1 converters, and the noise suppression devices 

was a weakness in the quality assurance program. 

3.6 Process to Manufacture Printed Wirina Boards (PWBs) 

Typically, Eaton design engineers prepare the schematics for the modules’ 

design and submit them to NNECO for review and approval. 

approval, computer assisted designers generate artwork (silk screen, component 

side and circuit side of the PWB), and send it to a subcontractor for the 

fabrication of PWBs. Eaton populates (inserts components) the PWBs according 

to design drawings, sends them through the wave soldering machine to solder 

the components, and builds a prototype module. 

successfully passes the tests at both Eaton and MP-2, does Eaton commence the 

manufacturing of production modules. 

After NNECO’s 

Only after the prototype 

The inspectors reviewed the circumstances that led to the failures of 6N638-1 

actuation modules and determined that the prototype modules successfully 

passed the tests at Eaton and MP-2. 

modules were tested at Eaton in the presence of NNECO quality control 

inspectors, they failed the insulation resistance tests during Hypot testing 

because the spacing between copper conductors on the PWB (clad runs) was 

inadequate. 

However, when the 6N638-1 production 

Eaton’s Work Order Instruction 1936-990 dated June 9, 1993, indicated that the 

PWBs that failed the tests were not shipped to MP-2 and that they were 

5 

40 
scrapped. Eaton redesigned the 6N638-1 actuation modules, increasing the 

distances between clad runs. PWBs manufactured from the revised drawings 

successfully passed the insulation resistance tests. 

The records of the identification and disposition of nonconforming items were 

not readily retrievable. For instance, the documentation relating to the 

failure of the actuation modules during the Hypot tests, the corrective action 

taken to redesign the modules, and the disposal of the nonconforming PWBs was 

not readily retrievable. The inspectors identified this matter as a weakness 

in the implementation of the established quality assurance program. 

quality control personnel did not document the insulation resistance failure 

of the actuation modules 6N638-1 during the Hypot tests in a discrepancy 

report (DR) . such as the investigation of the failure, the root cause (inadequate spacing 

between the clad runs), the action taken to correct the unacceptable spacing 

(redesign the module by increasing the spacing), and the final disposal of the 

failed PWBs (scrapped). 

assurance document with adequate description on the problem. When the 

inspectors pointed out this weakness to the Eaton Quality Manager, he 

responded that he will correct this weakness during the next revision of the 

quality assurance manual. 

Eaton 

Eaton could have used the DR to document subsequent actions, 

The DR could have been a readily retrievable quality 

The inspectors found that the actions taken by Eaton were acceptable even 

though there was a weakness in the documentation on the dispositioning of the 

nonconforming PWBs. 

3.7 Review of 10 CFR Part 21 (Part 21) Prosram 

The inspectors reviewed the program established to implement the reporting 

requirements of Part 21 as discussed in the following sections. 

3.7.1 

(SPI) No. 1563-031, "Reporting of Defects and Noncompliance," dated September 

14, 1981. The procedure had not been revised to reflect that Eaton was the 

current entity responsible for implementing the reporting requirements of Part 

21. Furthermore, the procedure did not reflect the current ;L,~* *ements of 

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

its evaluation," of Part 21 which requires Eaton to adopt appropriate 

procedures for evaluating and reporting defects and failures to comply. 

The inspectors reviewed Revision A to Standard Procedure Instruction 

Contrary to the above, SPI No. 1563-031 did not have provisions that would 

implement the above requirements. The inspectors informed Eaton engineers 

that failure to have a procedure to implement these provisions of Part 21 

constituted a violation of minor significance and would be treated as a Noncited Violation, consistent with Section IV of the NRC Enforcement Policy 

(NUREG 1600). 

Section 21.21(b) of Part 21 states that if a supplier of basic components 

determines that it does not have the capability to perform evaluations to 

determine if a defect exists, then it must inform the purchasers or licensees 

within five working days of making this determination. Section 2i.21(~) of 

Part 21 requires a director or responsible officer to notify the commission 

6 

41 
when he or she obtains information reasonably indicating a failure to comply 

associated with substantial safety hazards, or a defect. However, Part 21 

does not explicitly require these provisions to be included in procedures 

adopted pursuant to the regulations. Nevertheless, the inspectors expressed 

their concern that either insufficient or incorrect guidance may fail to 

prevent or even lead to violations of Part 21. 

3.7.2 In a letter dated May 10, 1994, Eaton notified the NRC pursuant to 

Part 21, of a problem with its design of the 6N642 module. 

a higher-than-normal failure rate of an integrated circuit (IC) on the 6N642 

electronic module assembly. The 6N642 electronic module is part of the ESAS 

at MP-2. Eaton experienced failures of the IC (Part U7) on this module. 

Eaton corrected the problem with the following actions: 

The design caused 

1. Upgraded the drawings for the schematic and artwork 

2. Upgraded the spare unit which was being manufactured. 

3. Upgraded the two existing spare units in the possession of 

Millstone-2. 

4. Prepared and released "Field Change Procedure for Correction of U7 

of 6N642-1," drawing KRT 136 so that the other units that had been 

suppl ied could be upgraded on site. 

The inspectors concluded that despite procedural weaknesses, this Part 21 

issue had been satisfactorily dispositioned. 

handling of Part 21 issues was identified. 

No instance of unsatisfactory 

3.8 Quality Control Training 

Eaton has established an acceptable training program for all employees. 

qualify personnel performing quality control (QC) inspection and testing 

activities, Eaton provides the "Study Guide for Inspector/Tester Qualification 

Program (Ref. QCR-82)," to its supervisors so that new employees are trained 

on appropri2L ' iformation which is required to pass the Inspector/Tester 

To 

qualification program test. 

records that inspection personnel were qualified to perform assignments. In 

addition to training documented in these qualification records, QC personnel 

receive additional training in a variety of subjects. 

Eaton Quality Assurance demonstrated through 

The inspectors reviewed the records maintained by Eaton's "Training 

Facilitator" and observed examples that Eaton had trained quality control 

department employees (Classified 0481) in the following technical areas: 

0 Quick response training (QRT) terminal board soldering 

0 SPI 571-2, Revision AN (list of manufacturing procedures) 

0 Use of MIL - standards at the workbench. 

7 
QRT Securing Capacitors 

Qual i ty Auditor Workshop 

Fundamentals of purchasing 

Principles of Materials Management 

Nuclear Coatings Seminar 

The training program for quality control and purchasing personnel did not 

provide guidance to detect the various fraudulent, or otherwise unacceptable 

products that have entered the nuclear industry and did not mention the 

numerous generic communications issued by the NRC on this subject. Eaton 

personnel concurred with the inspectors and committed to upgrade the training 

program accordingly . 

4 PERSONS CONTACTED 

Eaton CorDorat i on 

0 K.J. Cummings 

* G.A. DeRome 

* A. Emanuele 

W. Herrity * R. Magner * A. Mancini 0 N.J. Tarasovic 

D. luck 

P1 ant Manager 

Manager, Power Industry Controls 

Customer Service 

Design Engineer 

Qual i ty Control Engineer 

Senior Marketing Engineer 

Qual i ty Advancement Manager 

Training Facilitator 

 ADD^ ied Enersv Services, Overland Park. Kansas 

* S.A. Yousif Senior Project Manager 

* Attended the entrance meeting on 8/8/95 

Attended the exit meeting on 8/10/95 

8 

43 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON. D C. 205554001 

+***+ 

July 7, 1995 

Dr. Stephen R. Specker 

Vice President and 

General Manager 

General Electric Nucl ear Energy 

175 Curtner Avenue 

San Jose, CA 95125 

SUBJECT: NRC INSPECTIOSJ REPORT 99900403/95-01 

Dear Dr. Specker: 

This letter transmits the report of the U.S. Nuclear Regulatory Commission 

(NRC) inspection conducted January 17-19, 1995, at the GE Nuclear Energy 

(GE NE) facility in San Jose, California. The inspection was conducted by 

Mr. S.D. Alexander of this office, and the findings were discussed with the 

cognizant members of your staff identified in the report at the conclusion of 

the inspection. 

Areas examined during the inspection and our findings are discussed in the 

enclosed report. The inspection was conducted to provide a basis for 

assessing the validity and completeness of your list of items from the 

American Heavy Trading Black Fox inventory supplied to GE NE by Divesco, 

Incorporated; to determine from your records the disposition of Black Fox 

inventory items, including those supplied to GE NE by D-Tech (formerly Temco, 

Inc., and OMTECH, Inc.). 

The inspectors also reviewed the actions taken by your staff to correct 

inspection findings identified in Inspection Report 99900403/94-02. 

these areas, the inspection consisted of an examination of procedures and 

representative records, interviews with personnel, and observations by the 

inspectors . 

Within 

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 Public Document Room. 

Sincerely n 

Speci a1 Inspection Branch 

Division of Technical Support 

Office of Nuclear Reactor Regulation 

Docket No.: 99900403 

Enclosure: Inspection Report 99900403/95-01 

cc: See next page 

44 
Dr. Stephen R. Specker -2- July 7, 1995 

cc w/encl: Mr. Dick Tettman, President 

D-Tech, Inc. 

15040 Los Gatos Boulevard 

Los Gatos, CA 95032 

Ms. S. Kay Fisher 

Quality Assurance Manager 

Divesco, Inc. 

5000 Highway 80 East 

Jackson, MS 39208 

45 
REPORT NO.: 

ORGANIZATION: 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY: 

INSPECTION DATES: 

LEAD INSPECTOR: 

OTHER INSPECTORS: 

REVIEWED BY: 

APPROVED BY: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF TECHNICAL SUPPORT 

99900403/95-01 

GE Nuclear Energy 

175 Curtner Avenue 

San Jose, California 95125 

Kenneth W. Brayman 

Manager, Quality Assurance Systems 

GE Nuclear Energy’s activities within the scope of 

this inspection include supplying replacement parts 

and equipment to the nuclear industry. 

January 17-19, 1995 

vendor Inspection Section (VIS) 

Special Inspection Branch (TSIB) 

None 

Robert M. Gallo, Chief, TSIB/DOTS 717/?f 

Date 

Enclosure 

46 
1 SUMMARY OF INSPECT I ON FIND I NGS 

The inspection was conducted (1) to provide a basis for assessing the validity 

and completeness of the list of items from the American Heavy Trading Black 

Fox inventory supplied to General Electric Nuclear Energy, (2) to determine 

the disposition by GE Nuclear Energy (GE NE) of the Black Fox inventory items, 

including those supplied to GE NE by Divesco, Inc., D-Tech (formerly Temco, 

Inc., and OMTECH, Inc.), or others, if any, and (3) to close out previous 

Nonconformance 94-02-01. 

The inspection basis consisted of the following: 

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

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

Federal Requlations (10 CFR Part 50) 

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

1.1 Violations 

None 

1.2 Nonconformances 

None 

1.3 ODen Item 

(99900403/95-01-01) Review of GE NE pol icy, procedure, and practice regarding 

QA, QC, and supervisory review of test data records or other doeuments 

associated with activities affecting quality (See Paragraph 3.2 of this 

report). 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

2.1 Nonconformance 99900403/94-02-01: (Closed) 

Contrary to the requirements of Criterion V!, "Instructions, Procedures, and 

Drawings" of Appendix B, "Qual i ty Assurance Criteria for Nuclear Power P1 ants 

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

Resulations (10 CFR Part 50, Appendix B), GE NE Test Instruction (TI) 4389, 

used for dedicating molded-case circuit breakers for safety-related 

applications did not have appropriate acceptance criteria for determining, 

during the instantaneous magnetic trip test, that the breaker would not trip 

below the lower tolerance limit of the design magnetic trip band. 

criteria were also not found in the test equipment operating instructions. 

Consequently, for example, in GE NE dedication Work Order 93554, the hold 

current value (the test current pulse for which the breaker does not trip) was 

not recorded. 

Such 

As a result of the GE NE response to Nonconformance 94-02-01 (GE NE Letter 

dated October 7, 1994), the inspector reviewed the dedication documents and 

test instructions (TIS) associated with Work Order (WO) 093554 again. 

2 

47 
Specifically reviewed were: (1) Selected Item Drawing (SID) DD213A9893, 

Revision 6, dated December 20, 1993 (the item tested under WO D93554 was a 

TEC36100SSTlZRS molded-case circuit breaker (MCCB) which is Part 4 on SID 

DD213A9383), (2) Test Instruction TI 4353, Revision 7, dated October 22, 1993, 

(3) TI 4389, Revision 2, dated September, 22, 1993, and (4) Work Order 093554 

itself, completed on February 1, 1994. The WO also referenced the above 

mentioned SID and TI 4353. 

magnetic trip test using the PS-600 test set in accordance with TI 4389. 

Step 4.1.10 of TI 4353 requires performing the 

The GE NE response stated that GE NE had determined that "TI 4389 did require 

all pertinent data to be recorded in Paragraph 2.11." However, Paragraph 2.11 

in the version reviewed during the April 1994 inspection (Revision 2, dated 

September 22, 1993) merely stated: "Record the pertinent [sic] data required 

for dedication for the following tests:" without specifying what data (and in 

what form) was pertinent and expected to be recorded. 

of TI 4389 in the inspection report, i.e., that it was not specific enough to 

ensure that all required data would always be recorded, remained valid. 

Therefore the criticism 

Also stated in the GE NE response was that TI 4389 had been revised for 

clarification, speci f ical ly addressing recording of "hold point" test 

currents. Presuming that this meant that hold current (i.e., a test current 

value at which the MCCB does not trip instantaneously) would be recorded, the 

response stated further that the test data sheet, Form QC 348, "now requires" 

hold data to be recorded. However, in reviewing the QC 348 form that is 

Attachment 1 to TI 4389, as well as the QC 348 that is also Attachment 3 of TI 

4353, the inspector found that Form QC 348 already provided for recording hold 

current. 

questions: 

Therefore, the language of the GE NE response raised the following 

(1) Was the text of Paragraph 2.11 of TI 4389 revised to require recording 

hold current or was it revised to require recording the specific data 

required by the test (for which blocks are already provided on the QC 

348 that is an attachment to TI 4389)? 

(2) The language of the response implied that the QC 348 form had been 

revised, presumibly by a revision to TI 4389, to which the QC 348 Form 

in question is an attachment. However, as is GE NE practice, the QC 348 

attached to WO 093554 was Attachment 3 to TI 4353, the dedication 

procedure for a type of MCCB, not TI 4389, the detailed test 

instruction. Therefore, were the QC 348s that are Attachment 3 to 

TI 4353, TI 4337, TI 4271, and any other TIS in which Form QC 348 is an 

attachment, also revised, either as a revision to the standard form or 

by revisions to all the procedures to which the form is an attachment? I 

Finally, the GE NE response, stated that TI 4389 sets up the PS-600 test set 

to perform the tests by ramping up the test current until the MCCB trips. 

However, the Multi-Amp "Instruction Manual for Circuit Breaker Test Set Model 

PS-600," Revision 2, dated August 15, 1991, and the PS-600 settings given in 

TI 4389, indicate that an incremental pulse method is used. Specifically, for 

the breaker tested under WO D93554, the PS-600 would have been set up to put 

out a series of pulses of 12 cycles duration (JOG ON CYCLE setting of 12) with 

a one second pulse interval (JOG OFF SECONDS setting of 01) with each 

successive pulse incremented in magnitude by the test set as a preset function 

of the JOG OFF pulse interval setting. 

by either a breaker trip or by the limit settings in the PS-600. 

The pulse series would be terminated 

3 

48 
The inspector agreed with GE NE'S conclusion that in this case, the 

performance of the magnetic trip of the MCCB tested under WO 93554 was 

satisfactorily verified. 

the test set applies test pulses. 

the test set operation and of its display and from discussion with the test 

technician that the assertion made in the GE NE response that "any current 

value below the trip value is the hold current value" is valid. If this were 

true, the test technicians who performed the other tests, documented in other 

WOs reviewed by the inspector, could have arbitrarily selected and recorded 

any value below the trip value captured by the test set display as the hold 

current. Whereas, the test technician explained that the values recorded for 

hold current are determined by noting the captured displayed trip value, 

determining the pulse amplitude increment for the JOG-OFF setting in use, and 

subtracting the applicable increment from the trip value, or if possible 

noting the pulse amplitude current value displayed for the pulse preceding the 

one for which the MCCB tripped. 

This was based on the inspector's knowledge of how 

However, it was not clear from knowledge of 

During this inspection, the inspector reviewed the revised version of TI 4389 

(revised since the April 1994 inspection) and found that contrary to what was 

stated in the GE NE response to Nonconformance 94-02-01, TI 4389 had in fact 

been revised in a manner responsive to our original concerns. 

address the first question stated above, the inspector determined that the TI 

was revised to include instructions for recording specific data and 

appropriate acceptance criteria. 

explained that it had not intended to imply in its response to the April 1994 

inspection report that the QC 348 form itself had been revised (which it had 

not) and that admittedly, the phrase "now requires" was an inappropriate 

choice of words. Neither the QC 348 form itself nor the attachments of Form 

QC 348 to other TIS were, or needed to be, revised. 

Therefore to 

With regard to the second question, GE NE 

3 INSPECTION FINDINGS AND OTHER C9MMENTS 

3.1 Black Fox Parts 

During an April 1994 inspection (See NRC Inspection Report 99900403/94-02, 

dated September 13, 1994), the NRC requested GE NE to research its records and 

provide information on the procurement, handling, and disposition of 

equipment, components, and parts belonging to the consignment to NSSS Divesco, 

Inc. (Divesco), from the cancelled Public Serjlice of Oklahoma (PSO) Black Fox 

Nuclear Plant (Black Fox) Project. Subsequently, GE NE transmitted to the NRC 

a list of seven procurements by GE NE of material traceable to Black Fox 

through Divesco. These procurements are listed in Table 1 of Appendix A to 

this report. During this inspection in January 1995, GE NE explained that it 

did not have its surplus material procurement records computerized or 

organized in a manner conducive to efficient searches of the kind requested by 

the NRC. Instead, GE NE had requested Divesco to provide the information. 

During this inspection, the inspector reviewed the information on the handling 

and disposition of material on the list of the seven procurements of AHT/Black 

Fox material from Divesco and found evidence that it was procured, inspected, 

and supplied in accordance with GE NE procedures which the NRC has extensively 

examined in the past (Refer to NRC Inspection Report Numbers 9990403/89-01, 

90-01, and 94-02 or to 1989, 1990 and 1994 volumes of NUREG-0040). No 

discrepancies were noted with the handling or disposition of the material in 

the Divesco list of seven procurements during this inspection. 

4 

49 
Also, because GE NE had not conducted an independent search of its own records 

as expected, the inspector searched the records at GE NE during this (January 

1995) inspection and identified procurements of material from Divesco and also 

another surplus equipment dealer formerly called OMTECH, Incorporated, then 

later, TEMCO, and now D-Tech, that were not listed on the Divesco list of 

seven procurements discussed above. 

of Appendix A to this report. It was not evident from the surplus material 

procurement records at GE NE what the source of the material was, but it was 

possibly from the AHT Black Fox consignment in question. Accordingly, this 

information was pursued at Divesco during a February 1995 inspection (See NRC 

Inspection Report 99901117/95-01) and also at D-Tech in March 1995. 

These procurements are 1 isted in Table 2 

During the visit to D-Tech in March 1995, conducted as part of the January 

1995 inspection of GE NE, the inspector determined that the material supplied 

to GE NE by D-Tech had been originally sold by GE NE to D-Tech and was being 

bought back by GE NE for resale to other GE NE customers. 

at one time belonged to the Black Fox project because some of the Black Fox 

surplus inventory was bought back from PSO by GE NE when Black Fox was 

cancelled. However, the records indicated that this material was not part of 

the American Heavy Trading consignment of Black Fox material to Divesco. 

Some of it may have 

During the inspections of Divesco, as discussed in Inspection Report No. 

99901117/95-01, and at GE NE and D-Tech, as discussed above, the inspector 

found no evidence that surplus material from the cancelled Black Fox project 

traceable to the AHT Black Fox consignment to Divesco was commingled with 

other material or procured, handled, Dr resold in a manner inconsistent with 

NRC regulations or detrimental to safety. 

3.2 QC Review 

In April 1994 and during this January 1995 inspection, the inspector noted 

that the GE NE test technicians who performed tests documented in dedication 

WOs would routinely sign those WOs in the QC review block. 

noted that this was the case on a WO reviewed in connection with a 10 CFR 

Part 21 notification by an NRC licensee regarding some MCCBs that failed 

during onsite tsbling. The GE NE WOs for the dedication of these MCCBs 

clearly showed that one of the test results was out of tolerance, yet the test 

technician erroneously signed the block at the bottom of the data sheet 

indicating that the QC review was complete and presumably that no 

discrepancies had been identified. The inspector learned that there is no 

other required, routine or random review of these WOs by independent QC, QA 

personnel or supervisors which would (or should) provide an opportunity to 

detect errors of this sort, i.e., missing data, as in the case cited in the 

previous nonconformance, incorrect data, or missed out-of-tolerance data as in 

the Part 21-reported case. The issue of GE NE policy, procedure, and practice 

regarding independent QA, QC, and supervisory review of test data records or 

other documents associated with activities affecting quality (i.e., by someone 

other than the technician who performs the tests) will be addressed in a 

future NRC inspection. This issue is designated Open Item 95-01-01. 

The inspector also 

4 PERSONS CONlACTED 

4.1 GE NE 

5 

50 
Forest Hatch, Manager, Services & Projects Qual ity 

Kenneth W. Brayman, Manager, Quality Assurance Systems 

Noel Shirley, Principal Licensing Engineer, Safety Evaluations Project 

Elanor Schock, Program Manager, Safety Evaluations Project 

Newel1 Metras, Dedication Testing Supervisor 

Robert Thomas, Procurement Engineer (Retired) 

4.2 D-Tech 

Dick Tettman, President 
APPENDIX A 

TABLE 1: DIVESCO LIST OF SALES OF AHT/BLACK FOX ITEMS TO GE NE (1986 ON) 

GE NE PO# DATE 

205-86K317 

AHT/BF 

205-865142 

AHT/BF 

205-86R638 

AHT/BF 

205-86R687 

AHT/BF 

205-86R689 

AHT/BF 

205-86R874 

AHT/BF 

205-87C630 

AHT/BF 

gTY ITEM 

04/28/86 1 FC valve 

DRAWING GE NE CUSTOMER SOURCE 

112D1459POOl Perry (CEI) 

05/14/86 2 MSIV blowers 213A3762P001 Clinton (IP) 

10/02/86 1 Temp Element 159C4520P005 River Bend(GSU) 

10/06/86 8 MSIYs 821-FO21 - 28 NMP2 (NMPC) 

10/06/86 1 MSIV Blower 47B518664 NMP2 (NMPC) 

11/17/86 2 Relief Valves 21A9508P001 Hatch (GP&L) 

03/13/87 1 Temp Element 159C4520P005 Clinton (IP) 

TABLE 2: GE NE PURCHASES FROM DIVESCO FROM NRC SEARCH OF GE NE RECORDS 

GE NE PO# DATE DRAW I NG GE NE CUSTOMER SOURCE 

205-85E813* unkn unkn Terminal Bd 147D76146004,5 Perry (CEI) 

*AHT/ BF 

*Not listed on Divesco’s 1985 sales records database 

205-85E949 07/05/85 2 Press XDCR MPL:C85N001 Perry (CEI) AHT/BF 

205-855769 11/25/85 1 Hyd Hand Pump 131C8966G001 Clinton (IP) AHT/BF 

205-86R686 10/03/86 1* Agastat 145C3217P041 River Bend *AHT/BF 

(3 of 4 rtnd, 4th: *S/N 77231248 retained, baring GE DWG #, vice Agastat part 

number, E7024PB002, as were the two, S/Ns 85170022,3 from Control Components. 

Therefore, the one finally kept by GE could have been from AHT/EF) 

205-85N648 09/12/85 58 Anlg Is01 204B6220AAG002 Perry (CEI) Allens Ck 

205-87C632 02/03/87 4 CKT Cards 272A8614P101,02,12,20 RB(GSU) Chism 

52 
ilNITED 57 ATES 

NUCLEAR REGULATORY COMMISSIOR; 

;?/ASHIRGTOhr 0 2 20555 0001 

July 10, 1995 

Mr. Charles L. Perry, General Manager 

ITT Barton 

ITT Fluid Technology Corporation 

900 South Turnbull Canyon Road 

City of Industry, CA 91749-1882 

SUBJECT: NRC INSPECTION NO. 99900113/95-01 

Dear Mr. Perry: 

This letter transmits the report of the U.S. Nuclear Regulatory Commission 

(NRC) inspection of ITT Barton at City of Industry, California, conducted by 

Mr. R.C. Wilson of this office on June 12-15, 1995. The purpose of the 

inspection was tcu review activities conducted under your 10 CFR Part 50, 

Appendix B, quality assurance program and 10 CFR Part 21 reporting program 

The inspection consisted of an examination of procedures and records, 

interviews with personnel, and observations by the inspectors. 

The NRC inspectors found no instances where the implementation of your qua 

assurance program failed to meet certain NRC requirements. No response to 

this letter is required. 

i ty 

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

of this letter and its enclosures will be placed in the NRC's Public Document 

Room. 

Sincerely, 

Robert M. Gallo, Chief 

Special Inspection Branch 

Division cf Technical Support 

Off i ce of Nucl ear Reactor Regul at i on 

Docket No. 99900113 

Enclosure: Inspection Report 999001 13/95-01 

53 
REPORT NO.: 

ORGANIZATION: 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY : 

INSPECTION DATES: 

INSPECTOR: 

REVIEWED BY: 

APPROVED BY: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF TECHNICAL SUPPORT 

99900113/95-01 

ITT Barton 

ITT Fluid Technology Corporation 

900 South Turnbull Canyon Road 

City of Industry, California 91749-1882 

Jerald E. Anderson 

Director of Quality Assurance 

818/961-2547 

Instrumentation such as Pressure, Level, and Flow 

Transmitters and Indicating Switches, and Valve 

Actuators 

June 12-15, 1995 

Richard C. Wilson, Senior Engineer 

Vendor Inspection Section (VIS) 

Special Inspection Branch (TSIB) 

- 

Gregory, walina, 677 Ch’ef, VIS/TSIB 

c 

Robert M. Gallo, Chief, TSIB 

Date 

Date 

Enclosure 

54 
1 SCOPE OF INSPECTION: 

ITT Barton supplies a variety of pressure, differential pressure, level, and 

flow transmitters and indicating switches, as well as valve actuators. The 

commercial nuclear portion of sales varies with the specific type of 

instrument, generally amounting to about 10% or less. Barton Industrial Sales 

in Glenwood, 111 inois, also manufactures and supplies nuclear safety-related 

differential pressure units; that facility was not covered in this inspection. 

The NRC inspector reviewed the implementation of selected portions of Barton's 

quality assurance (QA) program for supplying safety-grade components, and 

reviewed Barton's 10 CFR Part 21 program including reports that have been 

submitted to the NRC. 

and 10 CFR Part 21. 

The inspection bases were 10 CFR Part 50, Appendix B, 

No violations or nonconformances resulted from this inspection. Within the 

inspection scope, the inspector found that adequate programs were in place, 

and that some improvements were being incorporated. 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

No open findings remained from previous NRC inspections of Barton. 

3 INSPECTION FINDINGS AND OTHER COMMENTS 

3.1 Entrance and Exit Meetinqs 

In the entrance meeting on June 12, 1995, the NRC inspector discussed the 

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

interfaces with Barton management and staff. 

1995, the inspector discussed his find' 

In the exit meeting on June 15, 

d concerns with Barton management 

manual provided the top level requirements, with im 

lower tier QA instructions. Although the QU manua 

format, it was readily auditable and appeared to s 

10 CFR Part 50. Barton performed all activiities u 

whether or not safety-related, and manufactured a1 

particular model in the same manner. A limited am 

documentation (e.g. , testing and certification) was' 

safety-re1 ated POs. 

QA manager, the inspector determined that the QA organization had sufficient 

authority and organizational freedom to identify quality problems, initiate 

solutions, and verify implementation of the solutions. 

ased on the review of the QA manual and instructions and discussions with the 

2 

55 
3.3 Translation of Purchase Order Reauirements 

The NRC inspector selectively reviewed files for eight purchase orders (POs) 

to determine if the PO requirements were correctly translated into documented 

specifications, procedures, drawings, testing requirements, and products. The 

selected POs covered a variety of transmitter and indicating switch models, a 

valve actuator, and three types of replacement piece parts. 

Barton assigned a Register Number (comparable to a shop order or work order 

number) to each PO, and identified each specific instrument by a Product 

Identification Code (PIC) number. The PIC numbers contained two-digit fields 

identifying the applicable version for each of the parts of the instrument, 

such as housing, bellows, fill, and range spring. In one PO, a Model 288A 

differential pressure indicating switch with a model 352 remote sensor was 

defined by an 18-field PIC number. 

When Barton received an inquiry for a replacement instrument, the PIC number 

for the original equipment was retrieved as the definitive description of the 

replacement assembly. 

inquiry covered a replacement for an indicating switch supplied in 1983. 

While reviewing the 1983 PIC number, Barton found a drawing revision that 

changed the original range. That information was provided to the customer for 

review, before PO placement. As a result, the original issue of the 1995 PO 

specified the proper PIC number and range for the instrument to be supplied in 

1995. The NRC inspector considered Barton’s practice of ensuring the accuracy 

of the original PO to be a positive feature of their program. 

In an example reviewed by the NRC inspector, a 1995 

Barton engineering was preparing a detailed compilation of the PIC numbers 

covered by specific environmental qualification test reports. When completed, 

the list will replace the present practice of specifically reviewing the 

detailed characteristics of a specific configuration during the engineering 

review of customer inquiries. This approach will facilitate determination of 

environmental qualification pedigrees. 

The NRC inspector concluded that Barton’s controls effectively ensured that 

customer PO requirements were correctly incorporated into finished products. 

3.4 Manufacturinq and Testing 

The NRC inspector witnessed various material handling and manufacturing 

operations, but no activity specific to safety-related POs was in progress 

during the inspection. 

portion of final calibration testing of a model 753 pressure transmitter for 

foreign use. 

of Processes and Procedures for the PO. 

parameters such as range agreed with the register sheet. 

operations were in accordance with the calibration sheet and procedure. 

Barton personnel pointed out that the calibration procedure required previous 

elevated temperature testing; thus, each harsh environment instrument is 

actually exposed to its design basis temperature during final acceptance 

testing. 

control over final acceptance testing activities. 

The inspector witnessed the accuracy and repeatability 

The test procedure was the same as specified on the Certificate 

The inspector verified that 

All observed 

The inspector concluded that Barton exercised appropriate procedural 

3 
3.4 Accuracv and Calibration of Test Eauiment 

The NRC inspector examined the calibration traceability of a model 200 

differential pressure indicator, serial number 88063, shipped on April 13, 

1995, under Duquesne Light Company PO No. D193001. This is the same 

instrument described in Section 3.3 above, where the range had changed since 

the original unit was shipped in 1983. The accuracy was specified on Barton’s 

calibration certification sheet as ++ % of full scale. This sheet showed the 

final acceptance test data for an ascending and a descending calibration run, 

and specified the applicable test instrument as Barton # 93-3-53. 

calibration report for # 93-3-53 identified it as a Heise model CMM pressure 

gage, with NIST [National Institute of Standards and Technology] traceability 

through # 95-41-16, and provides data from a current calibration. 

calibration report for # 95-41-16 covered a Meise model 179E/QBT transfer 

standard, which in turn was calibrated against a Ruska model 2465-751 

deadweight tester. The deadweight tester was calibrated by the Ruska 

Instrument Corporation under Barton PO 37873. 

certificate identified the NIST test numbers for the Ruska standards used to 

calibrate the piston and masses of Barton’s deadweight tester. 

inspector reviewed the report of an audit of Ruska by ITT Barton on 

June 23, 1992, and considered it adequate to dedicate Ruska’s commercial grade 

calibration services by the standards of that time. 

concluded that Barton’s documentation adequately documented the calibration of 

the delivered pressure indicator. 

The 

The 

The Ruska calibration 

The NRC 

The NRC inspector 

The NRC inspector also reviewed Barton’s report of a June 11, 1992, audit of 

SIMCO Electronics, which performs most of Barton’s external calibrations, and 

had no concerns. Since Barton’s procedures require triennial audits of 

calibration service suppliers, the inspector inquired about plans for future 

audits of these vendors. Barton stated that future vendor audits will be 

contracted out to EGS Corporation of Huntsville, Alabama. Barton mentioned 

that EGS has been audited by a licensee group. 

3.5 10 CFR Part 21 Proqram 

The NRC inspector reviewed Barton’s procedure for reporting in accordance with 

10 CFR Part 21: QA Manual Procedure QU-121, “NRC Regulations to lOCFR, 

Part 21,“ Revision 0, dated June 1, 1994. The procedure satisfied the 

requirements of 10 CFR Part 21, but it focused on evaluating deviations and 

failures to comply with the technical requirements of procurement documents, 

and only briefly addressed the identification and reporting of such concerns. 

The inspector suggested revising the procedure to emphasize such reporting. 

The inspector also pointed out that Barton would rarely have the plantspecific information necessary to perform the required evaluation of 

deviations, and suggested instead that the procedure concentrate on the fiveday notification of customers addressed in 10 CFR 21.21(b), so that customers 

can perform the evaluation. 

procedure that would better address the Part 21-related activities that Barton 

normally performs. 

The suggested changes would result in a shorter 

4 

57 
The NRC inspector noted that an October 1994 licensee group audit of Barton 

reported a finding related to a Barton internal deviation report: that Barton 

failed to complete evaluation of a deviation within the 60-day evaluation 

period required by 10 CFR Part 21, and failed to submit a timely interim 

report. In fact, Barton was still investigating the possible occurrence of a 

deviation, and had not yet determined that a deviation had occurred. 

in the previous paragraph, the inspector discussed this confusion with Barton 

personnel. 

As noted 

Barton had recently initiated a charge to customers for safety-related 

equipment labelled a "10 CFR 21 configuration control engineering charge." 

The inspector pointed out to engineering and QA personnel that the 

configuration control activities actually apply to meeting the QA requirements 

of Appendix B to 10 CFR Part 50 as imposed by licensee POs for safety-related 

equipment or services, and not to the reporting requirements of 

10 CFR Part 21. 

dedication activities are subject to Appendix B. 

Even though Part 21 provides dedication guidance, the 

Review of selected specific issues, as detailed below, indicated that Barton's 

Part 21 reporting program was functioning properly. 

copies of QU-121, 10 CFR Part 21, and section 206 of the Energy Reorganization 

Act of 1974 were properly posted. 

the clarifications discussed above, Barton's activities with respect to 

10 CFR Part 21 appeared to be acceptable. 

The inspector noted that 

The inspector concluded that, subject to 

3.6 Review of SDecific Part 21 Repom 

Switch chatter in Model 288A and 289A differential pressure switches - 

Interim Report dated December 19, 1994, and Final Report dated 

February 15, 1995 (NRC Log Nos. 94-324 and 95-052) 

Barton began an engineering evaluation of mild environment equipment 

qualification in late 1994 as a result of a licensee group audit. 

review of 1980 and 1986 seismic test reports, Barton determined that switch 

chatter may have occurred that was not detected. 

used to mor,;t<:! :ontact chatter was not identified in the test reports, but 

was suspected of being an incandescent lamp. 

seismic tests early in January 1995, using instrumentation capable of 

measuring contact chatter as rapid as two milliseconds. 

showed that higher G levels, and setpoints very close to actual parameter 

values, produced the most chatter; there was rio chatter at 4 G. Barton 

provided all affected customers with a table showing the duration of contact 

chatter as a function of G level and the proximity of the trip setpoint to the 

actual differential pressure value. 

During 

The specific instrumentation 

Barton conducted additional 

The new testing 

The NRC discussed preliminary test results with Mr. Anderson in a February 7, 

1995, telephone call. After reviewing the seismic test report, the NRC again 

discussed the concern with Messrs. Anderson and Larson in a May 17, 1995, 

telephone call. 

inspection. The review of data from earlier seismic tests, and conducting 

additional tests, revealed a possible concern that had gone unnoticed for 

The inspector briefly reviewed the concern during the 

5 

58 
several years. 

notifications to be acceptable, and no further action is required. 

Possibly unqualified relays in Model 288A differential pressure indicating 

switches - Interim Report dated October 5, 1994, and Final Report dated 

October 17, 1994 (NRC Log Nos. 94-266 and 94-271) 

The inspector considered Barton's activities including 

In January 1994 Barton supplied a model 288 indicating switch containing 

relays as a replacement for a unit sold in 1972. 

act on notification from engineering that the new relays were different than 

the original relays. 

repeat order later in 1994 in accordance with procedure. 

additional seismic testing, which showed that the new relays constitute 

qualified replacements for the 1972 models. 

The sales department did not 

The discrepancy was discovered during processing of a 

Barton conducted 

The NRC addressed this concern during the May 17, 1995, telephone call and 

during this inspection. Barton personnel stated that several 1972 tests were 

customer- and lot-specific. Barton did not initi3t.e a configuration control 

program for indicating switches containing relays until 1978. Subsequent 

1980s seismic testing of model 288 switches did not include relays. The 1994 

testing demonstrated equivalence of the new relays to those supplied in 1972, 

but Barton personnel stated that they still do not sell a qualified model 288A 

indicating switch containing relays. (The relays were sometimes added to 

increase the power handling capability of the output microswitches, or to 

provide more contacts.) An isolated failure of sales to act on engineering's 

review of a purchase inquiry caused the concern. 

Barton's subsequent actions including notifications to be satisfactory. 

The inspector considered 

Qualification limitations on all Series 200 differential pressure 

indicators - Barton Industry Advisory dated March 13, 1995, and transmitted 

to all affected customers (No formal Part 21 report to NRC and no NRC Log 

NO.) 

During Barton's engineering evaluation of mild environment equipment, 

engineering identified the possibility that confusion might exist concerning 

the various bellows fill fluids used in Model 200, 227A, 288A, and 289A 

differential pressure indicators and switches. The Industry Advisory stated 

that for applications below 40" F and less than 1 Mrad gamma, Barton had 

recommended an aqueous sol uti on of ethyl ene glycol (B-f i 1 1 ) . Qual i f i cation 

reports for these instruments, which did not include B-fill samples, showed a 

3 Mrad radiation limit. The purpose of the industry advisory was to notify 

all affected customers that the 3 Mrad limit did not apply to B-fill, which 

was known to disassociate above 1 Mrad into gases which prevent proper 

operation. 

After notification by a licensee on April 10, 1995, the NRC addressed this 

concern in an April 11 telephone call with Barton, and also during this 

inspection. Barton personnel stated that the qualification report clearly 

identified that the 3 Mrad limit covered testing of D-fill and M-fill units, 

and did not mention B-fill. Certifications specifically cited the test 

report. Records showed no instance of supplying B-fill where the PO specified 

6 

59 
a radiation requirement. The inspector considered Barton's actions including 

notifications to be satisfactory. 

Other Recent Part 21 reports - 

The inspector considered other Part 21 reports submitted by Barton in the past 

five years and previously reviewed by the NRC, together with those discussed 

above, and concluded that no underlying root cause remained unaddressed. 

3.7 Commercial Grade Item Dedication 

Three of the eight licensee PO files reviewed by the NRC inspector covered 

dedication of commercial grade microswitches, O-rings, and bezel gaskets for 

nuclear safety-related use. 

to establish the desired parts characteristics, which essentially became the 

critical characteristics 1 ists. The Barton microswitch source control drawing 

defined the desired characteristics of the purchased switch, which had a 

unique manufacturer's part number. Barton then selected switches which have a 

limited range of actuation force, and assigned a unique Barton part number to 

that group. 

In each case Barton had worked with the supplier 

All of the specified characteristics were verified by testing. 

The elastomer supplier is audited triennially for the specific part numbers 

used. 

distributor accepts excess quantity from production runs beyond Barton's 

needs. An oil-exposure test is used to verify lots of EPT (ethylene propylene 

terpolymer) elastomer, and a receipt inspection for dimensions is performed. 

Barton is able to order custom production runs of O-rings because a 

Barton was also improving the process for dedication of commercial grade piece 

parts. 

(FMEA) reports prepared for each instrument, covering all of the variations of 

all parts. 

model 764 differential pressure transmitters. The critical characteristics 

identified in this process, that are not verified by either the assembly 

process or an existing functional test of the part or a higher level assembly, 

are defined as "Barton Critical Characteristics." These characteristics will 

then be listed on the part drawing, and QA is procedurally responsible to 

define and perform verifications of them. 

Barton considered this process to be 85% complete, including all harsh 

environment equipment. 

evaluation to be a strength of Barton's program. 

The basis of the new program is failure modes and effects analysis 

The NRC inspector briefly reviewed the 36-page report covering 

At the time of the inspection 

The inspector considered the FMEA method of dedication 

A licensee group audit of Barton in October 1994 identified the incomplete 

FMEA and critical characteristics lists for the series 200 differential 

pressure indicators used in mild environments, and also found errors in the 

FMEA report for the NH-90 Series actuators. This licensee activity appears to 

be adequately addressing the implementation of Barton's dedication program. 

The inspector also reviewed two individual licensee surveillances in 1995 that 

had no findings. 

The NRC inspector considered Barton's dedication activities to be adequate. 

7 

60 
4 PERSONNEL CONTACTED 

+ * 

t * 

* 

* 

* 

* 

* 

* 

+ 

t 

t 

t 

t 

+ 

+ 

i- 

+ 

+ 

t 

t 

* 

* 

* 

* 

* 

C.L. Perry, General Manager 

J.E. Anderson, Director of Quality Assurance 

T.W. Holdredge, Quality Manager 

D.L. Norman, Quality Assurance Administrator 

R.L. Krechmery, Director of Engineering 

J.K. Meyer, Engineering Manager, Nuclear Products 

M.K. Larson, Senior Staff Engineer, Nuclear Products 

M.P. Loo, Contracts Manager 

J.E. Incotri, Marketing Manager 

T.E. Roide, Fabrication Manager 

G.M. Busch, Materials Manager 

L.F. Dropulic, Product Manager for Differential Pressure Units 

R. Einem, Product Manager for Actuators 

S. R. Goldberg, Product Manager for Electronics 

R.W. Pownell, Metrology Engineer 

T. Tran, Electronics Technician 

t Attended the entrance meeting on June 12, 1995 

* Attended the exit meeting on June 15, 1995 

8 

61 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555oOol 

. +*t** ' 

August 23, 1995 

Mr. R. Nim Evatt, President 

and Chief Executive Officer 

Liberty Technologies, Inc. 

555 North Lane 

Conshohocken, PA 19428-2208 

SUBJECT: NRC INSPECTION REPORT NO. 99901225/95-01 

Dear Mr. Evatt: 

This letter addresses the U.S. Nuclear Regulatory Commission (NRC) irxpection 

of your facility at Conshohocken, PA, conducted by Messrs. J.B. Jacobson and 

T. Scarbrough of this office on August 8 and 9, 1995, and the discussion of 

their findings with you at the conclusion of the inspection. The inspection 

was conducted to evaluate your actions with regard to open items identified 

during a previous NRC inspection (99901225/91-01) , to review current technical 

issues pertaining to the use of the "VOTES" valve operation test and 

evaluation system, and to review Liberty Technology's implementation of 

requirements delineated in Part 21, "Reporting Defects and Noncompliance," of 

Tit1 e 10 of the Code of Federal Reaul ations (10CFR). 

Areas examined during the NRC inspection and our findings are discussed in the 

enclosed inspection report. 

procedures and representative records, discussion, and observations by the 

inspectors. 

This inspection consisted of an examination of 

The inspectors determined that you have taken appropriate actions with regard 

to previous NRC open items and have implemented an effective program for 

meeting the requirements of 10 CFR Part 21. 

identified in your corrective action program for meeting the requirements of 

Criterion XVI "Corrective Action" of Appendix B to 10 CFR Part 50. 

Specifically, the inspectors determined that potential safety issues and 

nonconformances are not being uniformly documented within your quality 

program. Although evaluations are apparently being performed as safety issues 

arise, complete documentation of the evaluations was not avaible for review in 

some instances. 

Weaknesses, were however 

The inspectors also performed a limited review of your actions taken to 

Val idate the accuracy with which your new "Motor Power Monitor" equipment can 

predict motor actuator thrust at torque switch trip. 

equipment, the inspectors identified a weakness in not comparing Motor Power 

Monitor readings against a known accurate source other than the VOTES 

equipment. 

With regard to this 

Please provide us within 30 days from the date of this letter a written 

statement in accordance with the instructions specified in the enclosed Notice 

62 
R. Nim Evatt -2- 

of Nonconformance. We will consider extending the response time if you can 

show good cause for us to do so. 

The response requested by this letter and the enclosed Notice are not subject 

to the clearance procedures of the Office of Management and Budget as required 

by the Paperwork Reduction Act of 1980, Public Law. 96-511. In accordance 

with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter and 

the enclosed inspection report will be placed in the NRC Public Document Room. 

If there are any questions concerning this inspection please contact 

Mr. Jeffrey B. Jacobson al; (301) 415-2977. 

Sincerely, 

Robert M. Gallo, Chief 

Speci a1 Inspection Branch 

Division of Inspection and Technical Support 

Office of Nucl ear Reactor Regul at i on 

Docket Ne. : 99901225 

Enclosures: 1. Notice of Nonconformance 

2. Inspection Report No. 99901225/95-01 

63 
NOTICE OF NONCONFORMANCE 

Liberty Technologies, Inc. 

Conshohocken, PA 

Docket No.: 99901225 

Based on the results of an NRC inspection conducted on August 8 and 9, 1995, 

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

with NRC requirements. 

A. Criterion XVI, ''Corrective Action," of Appendix B to Title 10 of the 

Code of Federal Reaulations (10 CFR) Part 50, states in part, that 

measures shall be established to assure that conditions adverse to 

quality such as failures, malfunctions, deficiencies, deviations, 

defective material and equipment, and nonconformances are properly 

ident i f ied and corrected. 

Liberty Technologies Quality Assurance Procedure No. QA-NCR-004, 

"Processing Safety Concerns," Revision 0, states in part, that any 

individual that discovers a condition that is, or is suspected of being, 

a safety concern shall complete a Safety Concern Evaluation. 

Liberty Technologies Quality Management System Process No. QMS-QA-06, 

"Nonconformance/Corrective Action Control," dated May 4, 1995, states in 

part, that all nonconformances to established procedures and errors in 

software, services, and management systems, be documented, processed and 

resolved correctly. The individual identifying the nonconformance shall 

issue a Nonconformance/Corrective Action Report. 

Contrary to the above, Liberty Technologies, Inc. did not initiate a 

Safety Concern Evaluation or a Nonconformance/Corrective Action Report 

for an issue involving a software virus nor for an issue involving 

potential inaccuracies of Votes equipment at low thrust values. 

(Nonconformance 99901225/95-01) 

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: (1) a description of steps that have been or will be taken to correct 

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

prevent recurrence; and (3) the dates your corrective actions and preventive 

measures were or will be completed. 

Dated at Rockville, Maryland 

this 23rd day of August, 1995. 

Enclosure 1 

64 
U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF TECHNICAL SUPPORT 

REPORT NO.: 9990 1225/95-01 

ORGANIZATION: Liberty Technologies, INC. 

555 North Lane 

Conshohocken, PA 19428 

ORGANIZATIONAL 

NUCLEAR INDUSTRY Liberty Technologies, Inc. supplies systems 

ACTIVITY: 

Susan Yirnkanich, Quality Program Manager 

CONTACT : (215) 834-0330 

for testing and diagnosing the condition of motor 

operated valves. 

INSPECTION DATES: August 8 and 9, 1995 

/3 

8 i 2, ./ qjDate 

LEAD INSPECTOR: 

OTHER INSPECTORS : Thomas Scarbrough, NRR 

>>_q) 

Date 

REVIEWED BY: 

Donald P. Norkin, Section Chief 

Special Inspection Branch 

APPROVED BY: ~/2&- 

Robert M. Gallo, Chief Date 

Special Inspection Branch 

Enclosure 2 

65 
1 SUMMARY Qf INSPECTION 

1.1 SCODB 

This inspection was conducted to review Liberty Technology's response to open 

items identified during NRC inspection #99901225/91-01, to review current 

technical issues pertaining to the use of the "VOTES" valve operation test and 

evaluation system, and to review Liberty Technology's implementation of 

requirements delineated in Part 21, "Reporting Defects and Noncompliance," of 

Title 10 of the Code of Federal Requlations (IOCFR). 

1.2 Violations 

No violations were identified during this inspection. 

1.3 Nonconformances 

1.3.1 Nonconformance 95-01-01 

This nonconformance, described in sections 3.1.2 and 3.1.3 of the report 

identifies weaknesses in Liberty Technology's programs for implementing the 

requirements of Criterion XVI, "Corrective Action," of Appendix B to lOCFR 50. 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

2.1 Open Item 91-01-01 (CLOSED) 

During a previous NRC inspection of Liberty Technologies (99901225/91-01) the 

inspectors identified that Liberty had not (1) validated particular 

uncertainty terms by testing, (2) provided clear overall system error 

information in the VOTES User's Manual, (3) included uncertainties in the use 

of MOV diagnostic equipment (such as rate-of-loading and torque switch 

repeatability effects) in the VOTES User's Manual, and (4) validated the 

overall system error by testing. These items were tracked as Open Item 

91-01-01. 

During this inspection the staff reviewed Liberty's actions in response to 

this open item. 

actions taken by Liberty since the 1991 inspection are summarized below: 

I 

~ 

The previously identified weaknesses and the corresponding 

I 1. Validation of Uncertainty Terms 

During the 1991 inspection, the inspectors noted that several 

uncertainty terms used in the error analysis for the VOTES diagnostic 

equipment had not been verified. 

With respect to the assumed error for machining tolerances of valve 

stems in the determination of effective stem diameter, Liberty 

conducted a testing program of valve stems with threads machined to 

the limit of the allowable tolerance. 

Liberty performed a study to determine the sensitivity of the 

Using a finite element model, 

66 
effective stem diameter to thread tolerance. The VOTES software 

version 2.3 provides revised effective stem diameters. Liberty also 

has included a discussion of effective stem diameter in Addendum 5, 

"Stem Material Constants and Torque Correction," of the VOTES User's 

Manual. 

With respect to the assumed error in the values selected for the 

modulus of elasticity and Poisson's ratio for various stem materials, 

Liberty performed a stem material study that indicated that the ratio 

of these values used in the VOTES software needed to be revised. On 

October 2, 1992, Liberty notified the NRC in accordance with 10 CFR 

Part 21 of this problem and the actiion taken to alert VOTES users. 

Liberty has incladed a discussion of the proper assumptions for 

modulus of elasticity and Poisson's ratio in Addendum 5 of the VOTES 

User's Manual. 

With respect to the assumed effects on the modulus of elasticity and 

Poisson's ratio as a result of temperature changes, Liberty reviewed 

existing literature and increased the assumed error resulting from 

temperature changes. 

Error Analysis in Addendum 4 of the VOTES User's Manual to address the 

increased error. During this inspection, it was identified that 

Liberty inappropriately combined the! error to the assumed ratio of the 

modulus of elasticity and Poisson's ratio resulting from temperature 

changes with other errors through a square-root-of-sum-of-squares 

methodology. The error resulting from temperature changes is a biased 

error and therefore, should not be applied as a random error with 

other errors. 

be negligible. 

Liberty has revised the discussion of the VOTES 

The overall effect on the error analysis was found to 

With respect to the assumed error for changes in the modulus of 

elasticity of the yoke material caused by temperature changes, Liberty 

reviewed existing literature to provide additional support for the 

assumed error resulting from yoke temperature changes. This error 

also appears to have been combined as a random error with other 

errors. Again the overall effect on the error analysis was 

negl igi ble. 

Liberty has taken several actions to address the uncertainty intended 

to account for the non-linearity of the yoke, torsional effects of the 

yoke, stem directional erfects, and other effects. Liberty has 

revised the VOTES software to allow for calibration of the yoke sensor 

using a curve-fit analysis to minimize yoke non-linearity effects. 

Liberty has prepared guidance for calibration of the yoke sensor when 

the stem is in tension (valve opening direction). In the October 2, 

1992, Part 21 notice, Liberty discusses the increased error that could 

result from torque effects of the valve stem when the yoke sensor is 

calibrated based on strain of the threaded region of the stem. 

Liberty has revised the VOTES software to include a torque correction 

factor and discusses this issue in Addendum 5 of the User's Manual. 

3 
2. Overall System Error Informat ion 

During the 1991 inspection, the inspectors identified that the VOTES 

User's Manual did not discuss the basis for statistical uncertainty of 

the overall VOTES system error of 9.2 percent as determined by 

Liberty. Since then, Liberty has included a discussion of statistical 

uncertainty of the VOTES system error in Section 30-6, "Overall Thrust 

Measurement Accuracy," of the VOTES User's Manual. 

3. Torque Switch Repeatability and Rate-of-Loading Uncertainties 

During the 1991 inspection, the inspectors identified that the VOTES 

User's Manual did not discuss uncertainties resulting from torque 

switch repeatability and rate-of-loading effects when using the VOTES 

equipment. Since then, Liberty has included a discussion of these 

uncertainties in Section 30-7, "Factors Affecting Thrust at Torque 

Switch Trip," and in Addendum 4 of the VOTES User's Manual. 

4. Validation of Overall VOTES System Error by Testing 

During the 1991 inspection, the inspectors identified a lack of 

overall testing to verify the calculated VOTES system error. 

Following the 1991 inspection, Liberty participated in a testing 

program conducted by the MOV Users Group (MUG) of nuclear power plant 

licensees. 

determination of the overall VOTES system error. 

The results of the MUG testing program supported Liberty's 

The inspectors concluded that Liberty had adequately addressed the i ssues 

identified in Open Item 91-01-01. 

3 INSPECTION FINDINGS AND OBSERVATIONS 

3.1 Review of VOTES Technical Issues 

The inspectors reviewed Liberty's evaluation of potentially significant MOV 

diagnostic equipment issues. 

implementation of the Liberty 10 CFR Part 21 program and Liberty's 

notification to VOTES users of significant information concerning the use of 

their MOV diagnostic equipment. 

3.1.1 

As part of this review, the staff evaluated the 

The specific issues reviewed were as follows: 

Extrapolation of Open Thrust Data 

In May 1993, Liberty became aware of potentially large errors associated with 

thrust readings obtained from the VOTES equipment beyond the calibration range 

in the valve opening direction. 

evaluation to determine the safety impact on nuclear plant operation, the 

cause of the problem, and possible corrective action. Liberty considered the 

issue to not be safety significant, and not requiring a 10 CFR Part 21 notice 

because the torque switch is bypassed in the valve opening direction. 

However, Liberty believed that, in the long term, structural or motor output 

capability might be affected. Liberty issued Customer Service Bulletin (CSB) 

Liberty initiated a "safety concern" 

4 
31 (November 19, 19931, "Stem Tension," and CSB 31 Addendum (February 25, 

1994), "Extrapolation Errors at 09 Quantified," to address the issue and to 

provide guidance to VOTES users. 

notification was appropriate for this issue. 

The inspectors concluded that this method of 

3.1.2 Accuracy of VOTES equipment at Low Thrust Levels 

At recent industry meetings, the reduced accuracy of MOV diagnostic equipment 

at low thrust levels compared to its accuracy at high thrust levels has been 

discussed. On June 16, 1994, Liberty issued Customer Service Bulletin 34, 

"Running Load Differences Between VOTES and Packing 'nForcer," that discussed 

observed differences in thrust at low thrust levels between VOTES equipment 

and their Packing 'nForcer equipment that is used for packing load 

measurements. At the end of the bulletin, Liberty stated that "even if a 

significant running load error were to exist in a VOTES trace, the affect on 

the static thrust margin is almost always negligible, and thus no corrective 

actions are deemed to be required." 

During the inspection, Liberty agreed that the inaccuracy of the VOTES 

equipment at low thrust levels could be greater than the published value of 

9.2 percent but was unable to provide documentation regarding the safety 

evaluation of this issue. The inspectors were concerned that additional 

inaccuracies at low thrust levels could be significant for MOVs with minimal 

thrust margin. 

Safety Concern Eva1 uat i on or Nonconformance/Correct i ve Act ion Report as 

Nonconformance 95-01-01. 

The inspectors identified Liberty's failure to initiate a 

3.1.3 VOTES Virus 

In 1994, Liberty discovered a virus in the VOTES software that prevented 

infected computers from operating. Liberty alerted VOTES users to the problem 

through the nuclear computer network and a problem report letter. Liberty has 

improved its software and procedures to reduce suscept i bi 1 i ty to virus attack. 

Although Liberty responded to the virus problem, Liberty did not implement 

their procedures for evaluating potential safety concerns or nonconformances. 

This was cited as another example of Nonconformance 95-01-01. 

3.2 Review of Motor Power Monitor 

Liberty presented a summary of the features of its new Motor Power Monitor 

(MPM) diagnostic equipment. This equipment is designed to non-intrusively 

calculate actuator thrust output at torque switch trip under static 

conditions, by measuring motor current and voltage from the motor control 

center. 

determining the accuracy of the MPM, Liberty compared thrust data obtained 

with the MPM to data obtained using its VOTES diagnostic equipment, for 230 

valve strokes on 22 motor operated valves. 

a published accuracy of 9.2 percent. Review of the data seemed to support 

Liberty's claims of the MPM being accurate to within 15 percent; however, the 

The published accuracy of this equipment is 15 percent. In 

The VOTES diagnostic equipment has 

5 

69 
 

inspectors saw a weakness in not comparing MPM readings against a known 

accurate source other than the VOTES equipment. The inspectors noted that MPM 

users will be expected to justify the accuracy of the MPM when used to make 

decisions regarding the operabil i ty of safety-re1 ated MOVs. 

4 PERSONS CONTACTED 

The NRC staff participating in the inspection and Liberty personnel contacted 

during the inspection are listed below. An (*) indicates individuals whom 

attended the exit meeting. 

NAME TITLd 

* Jeffrey B. Jacobson Inspection Team Leader, NRC 

* Thomas Scarbrough Senior Mechanical Engineer, NRC * R. Nim Evatt President and Chief Executive Officer, Liberty 

* Susan Yankanich Manager, Qual i ty Programs , Liberty * Paul J. Schott Mgr. , Nuclear Marketing & Int’l. Sales, Liberty * Michael J. Delzingaro Manager, Service Engineering, Liberty 

Robert L. Leon V.P. and Chief Technical Officer, Liberty 

6 

70 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-OOOl 

+**** - 

September 27, 1995 

Mr. E.A. George, Jr., Vice President 

Mid-South Nuclear, Inc. 

40-B Sayerton Drive 

Birmingham, AC 35202 

SUBJECT: NRC INSPECTION NO. 99901270/95-01 

Dear Mr. George: 

This letter addresses the U.S. Nuclear Regulatory Commission (NRC) inspection 

of your facility at Birmingham, Alabama, conducted by Mr. U. Potapovs of this 

office on August 22 through 25, 1995, and the discussions of his findings with 

you and members of your staff at the conclusion of the inspection. The 

inspection was conducted to evaluate your quality assurance program and its 

implementation in selected areas such as (1) control of purchased material and 

services, (2) upgrading of material purchased from non-qualified sources 

(commercial grade item dedication), and (3) the implementation of your 

corrective action commitments resulting from the NRC inspection which was 

conducted on January 25 through 28, 1994. 

The inspection was accomplished through ob,jective evaluation of selected 

procedures and records, discussions, and observations by the inspector. The 

specific areas examined during the NRC inspection and the findings are 

di scussed in the encl osed i nspect i on report. 

Our review of your activities in these areas indicated that, although 

significant improvements have been achieved in defining the methods of 

identification and verification of critical characteristics of commercial 

grade material , tliese methods do not always assure that material suppl ied and 

certified to ASME Code or 10 CFR Part 50, Appendix B qual i ty programs compl ies 

with the procurement document requirements. 

references to the pertinent requirements are identified in the enclosures to 

this letter. 

The specific findings and 

Please provide us within 30 days from the date of this letter a written 

statement in accordance with the instructions specified in the enclosed Notice 

of Nonconformance. 

show good cause for us to do so. 

We will consider extending the response time if you can 

The responses requested by this letter and the enclosed Notice are not subject 

to the clearance procedures of the Office of Management and Budget as required 

by the Paperwork Reduction Act of 1980, Public Law No. 96-511. In accordance 

with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter and 

the enclosed inspection report will be placed in the NRC Public Document Room. 

71 
€.A. George, Jr. -2- 

If there are any questions concerning this inspection we will be pleased to 

discuss them with you. 

Sincerely ,- 

Speci a1 Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Docket No.: 99901270 

Enclosures: 1. Notice of Nonconformance 

2. Inspection Report 99901270/95-01 

72 
Mid-South Nucle r, In 

Birmingham, A1 abama 

NOTICE OF NONCONFORMANCE 

D k t NO.: 99901270/95-01 

Based on the results of an NRC inspection conducted on August 22 through 25, 

1995, it appears that certain of your activities were not conducted in 

accordance with NRC requirements. 

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

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

be establ ished to assure that purchased material conforms to procurement 

documents. 

Section 3 of Mid-South Nuclear (MSN) Quality System Program, Revision 2, 

dated April 12, 1995, requires, in part, that applicable provisions 

necessary to meet customer purchase order (PO) requirements shall be 

included in appropriate documents or instructions and that the material 

to be supplied shall be processed in accordance with these documents. 

Contrary to the above, the inspection identified the following examples 

where the established measures did not assure that material was supplied 

in accordance with the customer purchase order requirements. 

(Nonconformance 99901270/95-01-01) 

1. TVA (Sequoyah) PO 95N5F-133214, dated April 21, 1995, for 24 feet of 

1 1/2 inch diameter ASME SA-479, Type 316 bar to be supplied as 

American Society of Mechanical Engineers (ASME) Boiler and Pressure 

Vessel Code (Code) Class 2 material required the vendor to provide 

documentation that his quality program meets ASME Section 111, 

Division 1, NCA 3800 current edition and addenda, and that the 

material has been supplied in accordance with the quality 

requirements of that program. 

MSN did not supply this material in accordance with their NCA 3800 

quality assurance program. Instead, MSN certified the material as 

being supplied in accordance with ASME Section 111, Article NC 2610 

which exempts the material from most of the NCA 3800 quality 

assurance provisions. Additionally, paragraph NC 2610 limits the 

nominal crossection of bar stock that can be supplied under that 

paragraph to less than one square inch. 

this dimension. 

The bar supplied exceeded 

2. TVA (Sequoyah) PO P95N5F-129724, dated February 22, 1995, for 32 

internally threaded, one inch, Class 3000, ASME SA-105 pipe caps, 

required these caps to be supplied as ASME Section 111, Class 2 

material with the same qual ity program appl icabil i ty statement as 

discussed in example 1, above. 

Enclosure 1 

73 
-2- 

MSN did not supply this material in accordance with their NCA 3800 

quality assurance program. Instead, MSN certified this material as 

supplied in accordance with ASME Section 111, paragraph NC 2610. 

Additionally, material hardness test records in the sales order file 

indicated that the hardness level of the caps supplied under this PO 

exceeded the maximum values permitted by the applicable material 

specification. 

3. TVA (Sequoyah) PO P95N5-135199, for 80 feet (20 foot lengths) of 

ASTM A-36 angle iron (6 x 6 x .375 inch), included the provision 

that commercial material, procured from unqualified source and 

dedicated by the supplier must have all critical attributes (e.g. 

chemical , tensile, hardness) required by the applicable material 

specification independently verified. 

MSN certified the material provided under this PO as meeting the 

stated requirements without independently verifying that the tensile 

properties of the material complied with ASTM A-36 requirements. 

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

Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555, 

with a copy to the Chief, Special Inspection Branch, Division of Inspection 

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

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

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

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

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

been or will be taken to prevent recurrence; and (3) the dates your corrective 

actions and preventive measures were or will be completed. 

Dated at Rockvi 11 e, Mary1 and 

this 27th day of September, 1995 

74 
REPORT NO.: 

ORGANIZATION: 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY: 

INSPECTION DATES: 

INSPECTOR: 

REVIEWED BY: 

APPROVED BY: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF INSPECTION AND SUPPORT PROGRAMS 

99901270/95-01 

Mid-South Nuclear, Iiic. 

40-8 Sayerton Drive 

P.O. Box 10063 

Birmingham, Alabama 35202 

E. A. George, Jr., Vice President 

Mid-South Nuclear. Inc. is a supplier of metal 

products to the nuclear industry. 

August 22 through 25, 1995 

07-2 1-9s 

Date 

Vendor Inspection Section (VIS) 

Special Inspection Branch (PSIB) 

Greclorv C. Cwalina. Chief 

Vendor" Inspection Section 

Special Inspection Branch 

n 

Enclosure 2 

75 
1 SUMMARY OF INSPECTION FINDINGS 

During this inspection, the NRC inspector evaluated Mid-South Nuclear Inc. 's 

(MSN) Commercial Grade Item (CGI) dedication process and assessed the 

effectiveness of MSN's corrective actions for nonconformances identified 

during the previous (January 25-28, 1994) NRC inspection. The nonconformances 

related to procedural and implementation deficiencies in MSN's CGI dedication 

program and improper application of the American Society of Mechanical 

Engineers (ASME) Boi 1 er and Pressure Vessel Code (Code) Section I I I, "Rules 

for Construction of Nuclear Power Plant Components," (Section 111) paragraph 

NX 2610(b) which provides for the exclusion of small parts from certain 

qual Sty assurance program requirements. 

selected sales orders and related documentation for safety related material 

processed after the 1994 inspection. 

The evaluation included the review of 

The inspection basis consisted of the following: 

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

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

Federal Rectulations (10 CFR Part 50) 

0 Section I11 of the ASME Code 

0 MSN' s Qual i ty Systems Program Manual and imp1 ement i ng procedures. 

1.1 Vi ol at i ons 

No violations were identified during this inspection. 

1.2 

1.2.1 

Nonconformances 

Nonconformance 95-01-01 

This nonconformance, described in Sections 3.3.1 and 3.3.2 of the report 

identifies three examples where the implementation of MSN quality systl!m 

program did not assure that material conformed to customer purchase order (PO) 

requirements due to: (1) Improper application of ASME Code, Section I11 

paragraph NC 2610, (2) supply of material with hardness level in excess of 

specification limit, and (3) failure to perform tensile testing as required by 

the PO. 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

2.1 Violations 

2.1.1 Vi ol at i on 94-01-01 (C1 osed) 

Contrary to Section 21.21, "Notification" of Title 10 of the Code of Federal 

Reaulations (10 LFR), MSN fai:ed to adopt a procedure to implement the 

provisions of 10 CFR Part 21 that were effective October 29, 1991. 

2 

76 
By letter dated March 31, 1994, MSN advised the NRC that it had obtained a 

current copy of 10 CFR Part 21 and revised procedure SOP-601 to include the 

applicable requirements. No additional discrepancies were identified in this 

area. 

2.2 Nonconformances 

2.2.1 Nonconformance 94-01-02 (Closed) 

Contrary to the requirements of 10 CFR 50, Appendix B and Section 3 of the 

MSN's quality system program, neither the MSN critical characteristics forms 

nor the sales orders for certain materials identified adequate critical 

characteristics and verification methods to ensure that the items being 

supplied met the customer procurement document requirements. 

As discussed in letters dated March 31, 1994 and May 22, 1995, MSN has revised 

their procedure SOP-701 "Dedication of Commercial Grade Items" and most of the 

material critical characteristics forms to provide additional guidance for the 

identification and verification of critical characteristics. Review of this 

guidance indicated significant improvement, however weaknesses were identified 

in the revised program requirements (see discussion in paragraph 3.3.2) 

2.2.2 Nonconformance 94-01-03 (Closed) 

Contrary to the requirements of ASME Code Section 111, Article NC 2610, MSN 

issued a certificate of compliance indicating that ASME SA-213, Type 304 

tubing had been furnished to TVA in accordance with the requirements of NC 

2610 without the required involvement of a Certificate Holder. 

By letters dated March 31, 1994 and May 22, 1995, MSN advised the NRC that it 

had obtained their customer's (Certificate Holder's) consent to use the 

provisions of NC 2610 and that training had been provided to employees to 

review the requirements of "Certificate Holder" consent for customer orders 

under ASME Section I11 when utilizing paragraph NC 2610. The review of recent 

sales order files dursng this inspection, however, identified instances of 

misapplication of or improper certification to Section 111, paragraph NC 2610. 

(See examples 1 and 2 of Nonconformance 99901270/95-01-01 and Section 3.3.1 of 

this report) 

3 INSPECTION FINDINGS AND OBSERVATIONS 

3.1 Entrance and Exit Meetinqs 

During the entrance meeting on August 22, 1995, the NRC inspector discussed 

the inspection scope and developed general information about MSN's products 

and activities. 

discussed his findings and observations with MSN's management. 

During the exit meeting on August 25, 1995, the NRC inspector 

5 

77 
3.2 DescriDtion of Facilities 

HSN has been accredited by the ASME as 3 Material Organization, authorized to 

manufacture and/or supply ferrous and nonferrous bars, threaded fasteners , castings, forgings, plate, seamless fittings, flanges, NPT stamped tubular 

products, structural shapes, welding material, and similar items. The scope 

of their Quality Systems Certificate (QSC) also includes the qualification of 

material manufacturers and suppliers of subcontracted services and upgrading 

of stock material. 

According to MSN management, approximately 85% of their products are supplied 

for safety related nuclear applications with carbon steel strtictural shapes 

providing the highest volume of material processed. 

under their ASME QSC as well as non-Code material under 10 CFR Part 50, 

Appendix B quality assurance requirements. Most of the Appendix B material is 

purchased as commercial grade and dedicated. 

MSN provides material 

MSN does not perform any material manufacturing operations at their facility 

and does not warehouse material. MSN has on-site capability to perform 

hardness testing, hydrostatic testing, limited flattening tests, and visual 

and dimensional inspections. Chemical and mechanical testing is subcontracted 

to qual if i ed 1 aboratori es . 

3.3 Quality Assurance Proqram ImDlementation 

3.3.1 Material Supplied to ASME Code Requirements 

MSN’s program for supplying ASME Code material is described in their Quality 

System Program Manual (QSPM), Revision 2, dated April 12, 1995 which is 

committed to meeting the requirements of ASME Section 111, NCA 3800 as well as 

the applicable portions of NQA-1 and 10 CFR Part 50, Appendix B. 

controls are described in the manual and in referenced implementing 

procedures. 

The program 

The NRC inspector reviewed several recent customer purchase orders and 

accompanying data packages and determined that, in most instaL-., the 

material appeared to have been supplied in accordance with the applicable QSPM 

provisions. 

compliance with customer purchase order and ASME Code requirements. 

the review identified inconsistencies in the processing and certifying 

materi a1 suppl i ed under ASME Sect ion I I I, paragraph NX 2610, i ncl udi ng 

instances of improper application of the provisions of this paragraph. 

review also identified inconsistencies and apparent contradictions in customer 

POs related to the acceptance criteria for material supplied under NX 2610 as 

illustrated in the following examples: 

3.3.1.1 TVA (Sequoyah) PO 95N5F-133214, dated April 21, 1995, for 24 feet of 

1 1/2 inch diameter ASME SA-479, Type 316 bar to be supplied as ASME 

Code Class 2 material required the vendor to provide documentation 

that his quality program meets ASME Section 111, Division 1, NCA 

Sufficient documentation was generally available to demonstrate 

However, 

The 

4 

78 
3800 current edition and addenda, and that the material has been 

supplied in accordance with the quality requirements of that 

program. 

MSN procured this material from an unqual ified suppl ier in two 

pieces, removed test coupons from each piece and sent the coupons to 

a qualified laboratory for chemical and mechanical testing. The 

chemical analysis was performed on each coupon while tensile testing 

was done on coupon representing on:y one of the two bars. According 

to MSN, this practice is consistent with their CGI dedication 

program which permits one test on unverified heat lot based on the 

supplier’s performance history. 

provided in accordance with ASME Section 111, paragraph NC 2610 

(small parts exclusion) which exempts the material from most of the 

NCA 3800 qual i ty assurance requirements. 

material in accordance with their CGI dedication program. 

The material was certified as 

MSN processes such 

The inspector noted that processing such material under MSN’s CGI 

program, while consistent with the requirements of NC 2610, did not 

appear to satisfy the PO requirement that the material must be 

supplied in accordance with the requirements of their NCA 3800 

quality program. It was also noted that the material processed 

under this PO exceeded the maximum size limit (one square inch cross 

section) of material that can be supplied under paragraph NC 2610 

and, therefore, should have been supplied and certified under MSN’s 

QSC (NCA 3800 program) in order to comply with the applicable ASME 

Code requirements. Improper application of the ASME Code, Section 

111, paragraph NC 2610 was identified as example 1 of Nonconformance 

99901270/95-01-01, 

Before the completion of the inspection, MSN requested that TVA 

clarify whether the statement in their POs which requires 

certification that material is supplied in accordance with MSN’s 

NCA 3800 quality program precludes the use of paragraph NX 2610. 

TVA responded that the statement in question does not preclude the 

uco of ASME Section I11 paragraph NX 2610 where applicable. 

According to TVA, the referenced PO paragraph is intended to 

indicate that the supplier is required to maintain a quality 

assurance program that meets ASME Section 111, NCA 3800 requirements 

and that the material is supplied with all the required 

document at i on. 

The inspector noted that acceptance of the NX 2610 small parts 

exclusion while requiring certification to NCA 3800 program 

requirements appeared contradictory and that the maintenance of an 

NCA 3800 quality program had no effect on the material supplied if 

the material is not required to be supplied in accordance with that 

program. It was also noted that, according to accepted ASME 

practice, material supplied in accordance with NX 2610 can not be 

certified as produced under the Material Organization’s QSC (NCA 

3800 program). 

5 

79 
3.3.1.2 TVA (Sequoyah) PO 95N5F-129724, dated February 22, 1995, item 2, for 

32 Class 3000 ASME SA-105 internally threaded (NPT l-inch) pipe caps 

specified that this material was to be supplied in accordance with 

ASME Code, Section 111, Class 2 requirements. 

MSN certified this material as supplied in accordance with ASME 

Code, Section 111, paragraph NC 2610. MSN obtained the material 

through a distributor (Dodson Steel Products) from an unqualified 

manufacturer, Bonney Forge (BF) with Certified Material Test Reports 

(CMTR) from two heat lots. The material was upgraded using MSN’s 

CGI dedication program. 

chemical analysis of one $ample from each heat lot and a hardness 

test (Rockwell B) on each piece. ASME SA 105 allows hardness 

testing (Brinell method) as an acceptable a1 ternate for verifying 

tensile properties on forgings too small to permit obtaining a 

subsize tensile specimen when such forgings are produced on 

equipment unsuitable for the production of separately forged test 

bars. 

Bri ne1 1 (HB) . 

The upgrading consisted of performing 

SA-105 specifies an acceptable hardness range of 137 to 187 

The inspector noted that the MSN test report indicated measured 

hardness levels of 93-96 and 92-96 Rcckwell By respectively, for the 

two heat lots of material supplied under this order. 

convert to 200-216 and 195-216 HB, which is significantly higher 

than the 187 HB maximum hardness permitted by ASME SA-105. 

to assure compliance with applicable procurement document 

requi rements was identified as example 2 of Nonconformance 

These ranges 

Failure 

99901270/95-01-01. 

It was also noted that, as discussed in paragraph 3.3.1.1, above, 

although the customer’s PO required the material to be supplied in 

accordance with the quality requirements of NCA 3800, MSN certified 

the material as supplied under NC 2610 which exempts the material 

from most of the NCA 3800 requirements. 

3.3.1.3 TVA PO 95N2T-1485851 for 66 feet of 2 inch, schedule 160, ASMt SA106, grade B pipe, specified this material to be supplied in 

accordance with ASME Code, Section 111, Class 2 requirements. The 

PO also required the vendor to supply documentation that his QA 

program meets ASME Section 111, NCA 3800, current edition and 

addenda, and that the material was supplied in accordance with this 

program. 

MSN obtained this material by commercial grade purchase from M&R 

pipe supply, who purchased it from Texas Pipe & Supply Co. Inc., 

who, in turn, obtained the pipe from Koppel Steel Co. The material 

was supplied in three pieces with Koppel Steel CMTR which stated 

that Koppel had performed bend and hydrostatic tests on this 

materi a1 . 

MSN upgraded this miiterial in accordance with their CGI dedication 

program by performing chemical analysis and hardness test on each of 

6 

80 
the three pieces. 

requirements for SA-106 material, no tensile testing was done. The 

results of Koppel Steel (unqual ified vendor) hydrostatic and bend 

testing were accepted without Val idation. 

chemical analyses were performed on each piece of the material 

because MSN did not have a documented performance history of this 

vendor. 

Consistent with their CGI dedication program 

According to MSN, 

MSN provided a Certificate of Compliance (COC) for this material 

which indicated by "X" marks that the material was manufactured and 

processed in accordance with requirements which included: 

ASME Section 111 NC 2610, 1989 edition 

MSN QA Program, Revision 2, dated April 12, 1995 

QSC 560, Expiration date May 5, 1998 

The inspector noted that the COC was contradictory and misleading, 

since it certified that the material was processed in accordance 

with their ASME QSC (NCA 3800) while the supporting documentation 

shows that the material was processed under MSN's CGI dedication 

program. The inspector also noted that this was another example 

where the customer's PO required the material to be provided in 

accordance with MSN's NCA 3800 program but was supplied under 

paragraph NC 2610. 

3.3.2 Commercial Grade Item Dedication Program 

MSN's program for purchasing and dedicating commercial grade material is 

described in Procedure SOP-701, Dedication of Commercial Grade Items. The 

current revision is Rev. 6, dated September 8, 1994. This procedure is used 

for supplying ASME Code material under the small parts exemption of Section 

I11 paragraph NX 2610 and for supplying all safety related material to the 

requirements of 10 CFR Part 50, Appendix B. The procedure has been revised 

since the last NRC inspection to address some of the concerns identified 

during that inspection. Additional guidance is provided for the 

identification and verification of critical characteristics. The procedure 

references Form 701, "Material Critical Characteristics Form" for the 

identification of critical characteristics and verification methods applicable 

to different materials and product forms. SOP-701 also requires justification 

for the selected critical characteristics to be identified on Form 701. This 

is accomplished by referencing a justification code. Engineering evaluations 

for all justification codes are compiled on Form 701B, "Critical 

Characteristics Selection Engineering Justification Code." 

Although the revised procedure required additional testing to verify material 

conformance to the applicable specification, certain materials and product 

forms were permitted to be dedicated based on an "indirect verification" 

method. 

properties conform to the specification requirements. The method was 1 imited 

to mild steel products for which approximate hardness versus tensile strength 

relationships are shown in ASME SA-370. 

extensive test data to support this relationship. 

This method utilizes hardness testing to verify that material tensile 

Additionally, MSN has compiled 

7 

81 
Mr. William A. McC 

Power Di stri but i on 

9870 Crescent Park 

West Chester, OH 

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-0001 

August 22, 1995 

oy, President 

Services 

Dr i ve 

5069 

SUBJECT: NRC INSPECTION REPORT 99901286/95-01 

Dear Mr. McCloy: 

This letter transmits the report of the U.S. Nuclear Regulatory Comidssion 

(NRC) inspection of Power Distribution Services, Inc. (PDS), West Chester, 

Ohio, conducted by Messrs K.R. Naidu, J.L. Knox, and R. Mendez on June 19-22 

1995. The inspection was conducted to provide a basis for NRC staff 

confidence that the switchgear manufactured by PDS for 4.16-kV Yaskawa circu 

breakers would perform their intended safety function. 

discusses an observation made during a January 25-28, 1995, inspection at Wy 

Laboratories, Huntsville, Alabama, when the qualification testing activities 

related to the 4.16 kV switchgear manufactured by PDS were in progress. 

June 22, 1995, at the conclusion of the inspection, the inspectors discussed 

the findings with you, other members of your staff, and representatives of 

National Technical Systems (NTS). 

This report also 

On 

t 

e 

During this inspection, the team evaluated the NTS/PDS quality program that 

was established to implement the provisions of 10 CFR Part 50, Appendix By and 

the provisions of 10 CFR Part 21 in selected areas during the manufacture of 

the switchgear cubicles. Within these areas, the NRC team (a) examined 

technical documentat irrn , procedures and representative records , (b) he1 d 

discussions, (c) listened to presentations and (d) observed PDS technicians 

worki ng act i v i ti es . 

During the evaluation of your activities at West Chester, the team noted the 

proactive approach being taken by your staff to correct adverse customer 

findings. The team noted positive PDS employee attitudes and technical 

expertise that were shown by the personnel who were interviewed during the 

inspection. However, the team observed that PDS personnel are experiencing 

difficulties adapting to a written quality program. 

Based on the results of the inspection, the inspectors found that the 

implementation of the NTS/PDS program failed to meet NRC requirements as 

specified in the enclosed Notice of Nonconformance. 

identified an inadequacy in the control of purchased materials. 

Specifically, the team 

82 
w. McCloy -2- 

Please provide us within 30 days from the date of this letter a written 

statement in accordance with instructions specified in the enclosed Notice of 

Nonconformance. We will consider extending the resoonse time if you can show 

good cause for us to do so. 

The response requested by this letter and the enclosed Notice are not subject 

to the clearance procedures of the Office of Management and Budget as required 

by the Paper Reduction Act of 1980, Public Law No. 96-511. 

In accordance with 10 CFR 2.790 of the Commission’s regulations, a copy of 

this letter and the enclosed inspection report will be placed in the NRC’s 

pub1 ic document room. 

Should you have any questions regarding this matter, please do not hesitate to 

call. 

Sincerely, 7- 

Robert @% M. Ga 0, Chief 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nucl ear Reactor Regul at i on 

Docket No.: 99901286 

Enclosures: 1. Notice of Nonconformance 

2. Inspection Report 

cc: Mr. Gregory M. Ruegger 

Nuclear Power Generation 

B14A 

Pacific Gas & Electric Company 

77 Beale Street, Room 145 

P.O. Box 77000 

San Francisco, CA 94106 

Mr. M. Basu 

Electrical Project Eng. 

77 Beale Street, Room 145 

P.O. Box 77000 

San Francisco, CA 94106 

Mr. D.R. Michaud ’ 

Product Manager 

NTS 

533 Main Street 

Acton, MA 01720 

83 
NOTICE OF NONCONFORMANCE 

Power Distribution Services, Incorporated 

West Chester, Ohio 

Docket No.: 9901286 

Report No.: 95-01 

Based on the results of a. U.S. Nuclear Regulatory Commission (NRC) inspection 

conducted at the Power Distribution Services, Incorporated (PDS) , West 

Chester, Ohio, facility on June 19-22, 1995, it appeared that one of your 

activities were not conducted in accordance with NRC requirements. 

Criterion VII, "Control of Purchased Material , Equipment, and Services" 

of Appendix B to Fart 50 of Title 10 of Code of Federal Requlations, (10 

CFR 50) states, in part, "Measures shall be established to assure that 

purchased material, equipment, and services, whether purchased directly 

or through contractors and subcontractors, conform to the procurement 

documents. These measures shall include provisions, as appropriate, for 

source evaluation and selection, objective evidence of quality furnished 

by the contractor or subcontractor, inspection at the contractor or 

subcontractor source and examination of products upon delivery." 

Contrary to the above, PDS failed to establish appropriate measures to 

control purchased materi a1 in Standard Operating Procedure (SOP) No. 02, 

Revision 1, "Purchasing Materials & Services for Nuclear Orders." 

Specifically, there were no provisions to utilize the same technical 

description of material equipment or services in the purchase order that 

had been approved by the National Technical Services (NTS)/PDS staff in 

the "Nuclear Purchase Requisition." Furthermore, the measures did not 

require that purchase orders for safety-related items be issued only to 

vendors listed in the NTS approved vendors list. 

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 (1) a description of steps that have been or will be taken 

to correct this item, (2) a description of steps that have been or will be 

taken to prevent recurrence, and (3) the dates your corrective actions and 

preventive measures were or will be completed. 

Dated at Rockville, Maryland 

this 22nd day of August, 1995. 

Enclosure 1 

84 
REPORT NO.: 

ORGANIZATION: 

ORGAN IZAT IONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY: 

INSPECTION DATES: 

LEAD INSPECTOR: 

OTHER INSPECTORS: 

REVIEWED BY: 

APPROVED BY: 

U. ICLE REGULATOR1 COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF INSPECTION AND SUPPORT PROGRAMS 

99901 286/95-01 

Power Distribution Services, Inc. 

9870 Crescent Park Drive 

West Chester, OH 45069 

Mr. J.L. Bachman 

(513) 777-4445 

Fabricating 4.16 kV switchgear cubicles and 

reconditioning low voltage metal-clad circuit 

breakers 

June 19-22, 1995 

.L-/; 7 l'? s" 

Kamalakar R. Naidu. Team Leader Date 

Vendor Inspection Section (VIS) 

Rogel io Mendez, Region I I I 

John L. Knox, NRR/EELB 

\ 

Robert KQL-Ah M. Gallz, Chief, PS1B:DISP Date 

Division of Inspection and Support Programs (DISP) 

Enclosure 2 

85 
1.0 SUMMARY OF FINDINGS ' 

During this inspection, the inspection team evaluated the National Technical 

Services,/Power Distribution Services, Inc. (NTS/PDS) quality program and its 

implementation during the fabrication of 4.16 kV retrofit switchgear which 

includes 4.16-kV, 350-MVA (million volt-amperes), SF, Gas Fluopac Series, 

Rotary-arc, circuit breakers manufactured by Yaskawa Electric Corporation 

(Yaskawa), Japan, intended for Pacific Gas and Electric Company's (PG&E's) 

Diablo Canyon Power Plant (DCPP). The Yaskawa circuit breakers have a higher 

short circuit interrupting capacity than the existing GE breakers (350 versus 

250 MVA), require less maintenance, and are compact enough to fit into the 

existing stationary GE cubicles. 

The inspection basis consisted of the foll owing: 

0 Appendix 6, "Qual i ty Assurance Criteria for Nuclear Power P1 ants and 

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

Requlations (10 CFR Part 50, Appendix B) 

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

One nonconformance was identified and is discussed in Paragraph 3.4.2 of this 

report. 

2.0 STATUS OF PREVIOUS INSPECTION FINDINGS 

This was the first NRC inspection of this vendor. 

3.0 INSPECTION FINDINGS AND OTHER COMMENTS 

3.1 Entrance and Exit Meetinqs 

During the v+,v*aclce meeting on June 19, 1995, the NRC inspection team 

discussed with PDS and NTS staff the scope of the inspection, the areas to be 

reviewed, and established the persons to contact within PDS and NTS management 

and staff. During the exit meeting on June 22, 1995, the NRC inspection team 

sumnarized its findings and concerns to the management and staff of PDS and 

NTS. Persons contacted during this inspection are identified in Section 4. 

3.2 Bac kqround 

NTS issued Purchase Order (PO) No. 36986 dated March 20, 1994, to PDS for the 

project management, engineering, quality assurance, production testing, 

manufacturing and technical labor associated with the supply of vertical-lift 

drawout cubicles with 4.16-kV, 350-MVA, SF, Yaskawa circuit breakers. PDS 

fabricates metal enclosures to permit the installation of 4.16-kV Vaskawa 

circuit breakers and other components into the existing stationary GE cubicles 

at DCPP. NTS provides the quality assurance (QA) coverage. In January 1995, 

NTS contracted Wyle Laboratories, Huntsville, Alabama, to subject a represent2 

86 
ative switchgear bay to DCPP site-specific seismic qualification tests. 

test specimen consisted of three GE stationary cubicles with retrofit drawout 

cubicles manufactured by PDS with 4.26-kV llaskawa circuit breakers. 

The 

In addition to the work being performed for DCPP, PDS has received three 480 

Volt GE metal-clad circuit breakers from the Waterford Nuclear Station 

(Waterford) and one similar circuit breaker from the River Bend Nuclear 

Station (River Bend) for complete reconditiioning. PDS informed the inspectors 

that it has submitted the procedures that lit had developed to perform the 

required re-conditioning to Waterford and River Bend and is awaiting necessary 

approval s. 

PDS fabricates electrical device enclosures using material , components, and 

sub-assemblies from other equipment manufacturers. 

supplies remanufactured low, medium and high voltage switchgear for various 

commercial power generation and distribution companies. 

variety of services for non-nuclear electric utility companies including 

testing and maintaining protective and power apparatus (circuit breakers, 

starters, transformers , network protectors, re1 ays, and electrical conductors 

from 600 V through 765 kV). 

It also assembles and 

PDS also provides a 

3.3 10 CFR Part 21 Proaram 

PDS personnel informed the inspectors that PDS and NTS entered into a Teaming 

Agreement, which was expressly developed to enable PDS to manufacture the 

vertical-lift drawout retrofit cubicles for. DCPP. According to the agreement, 

PDS implements the NTS/PDS quality program during the manufacture of the 

retrofit switchgear, and NTS maintains the 10 CFR Part 50, Appendix By QA 

program and 10 CFR Part 21 reporting responsibilities as defined in the NTS 

Qual i ty System for safety-re1 ated activities performed by PDS personnel on NTS 

nuclear orders. The NTS/PDS quality program requires NTS to review all 

Nonconformance/Corrective Action Reports (MCARs) initiated by PDS during the 

manufacture of the DCPP switchgear for Part. 21 reportability. 

The team reviewed the location and the adequacy of documents posted at the PDS 

facility pursuant to 10 CFR 21.6. PDS posting consisted of Section 206 of the 

Energy Reorganization Act of 1974 and a notice in accordance with 10 CFR 

21.6(b). The inspectors determined that the posting was acceptable. 

3.4 PA Proaram ImDlementation 

The NTS/PDS Quality Manual (QM) provided programmatic guidelines to supplement 

the NTS Qual i ty Assurance Manual as appl icable to PDS activities. 

developed standard operating procedures (SOPs) to implement the written 

program. NTS trained selected PDS individuals in quality control inspection 

techniques to implement the NTS/PDS Quality Program and the SOPs and certified 

them as PDS Quality Control Inspectors. NTS personnel perform quality 

assurance functions. The inspectors selected the following SOPs for review. 

3.4.1 SOP No. 1, Revision 1. dated April 17, 1995, "Nuclear Control 

Reviews," describes the methodology by which PDS reviews, and approves NTS 

nuclear orders. The inspectors observed that this procedure does not 

NTS/PDS 

3 

a7 
explicitly state that change orders to purchase orders (POs) should receive 

the same degree of review and control as the original PO. PDS initiated NCAR 

95-29 on June 22, 1995, to revise the SOP clarifying the requirement. 

3.4.2 SOP No. 2, Revision 1, dated April 17, 1995, "Purchasing Materials & 

Services for Nuclear Orders," describes the preparation, review, approval, 

issuance and verification process of procure:,rent documents by PDS for NTS 

nuclear orders. 

No. 2 requiring the purchasing agent to transcribe the same technical 

description that NTS approved in the "Nuclear Purchase Requisition" (NPR) prepared by PDS into the PO. 

According to the SOP, when a NPR is prepared, the technical description of the 

material is reviewed and approved by the NTS/PDS staff. However, there is no 

requirement for the same technical description to be transcribed into the 

purchase order. Paragraph 1II.B of SOP No. 2 states, in part, "Completed and 

PDS approved requisitions are forwarded to NTS for review and approval.. . NTS 

will review the requisitions per the relevant supplier file at PDS," assuring 

that the description of the purchased material is controlled. Paragraph II1.C 

of the procedure which discusses the preparation of the PO states, in part, 

approval of the resulting PO and all associated paperwork by NTS is indicated 

by signature and/or quality stamp and date on the hard copy of the resulting 

P.0 .... a copy of the NTS approved requisition and PO must be filed in "P.O. 

Requirements Review Sheet" and indicate approval of each requisition by 

signature, initials, or quality stamp and date for all nuclear purchases. A 

copy of the "P.O. Requirements Review Sheet, "...will be sent to Purchasing for 

the preparation of the actual hard copy P.O. However, the SOP does not 

require the transcription of the technical description that had been 

previously approved in the NPR into the PO. The inspectors informed the PDS 

staff that they were concerned that the intent of the review and approval of 

the technical description of a component or material in the NPR is defeated if 

the same technical description is not restated in the PO. 

The inspectors observed that there were no provisions in SOP 

Additionally, Paragraph H of SOP No. 2 did not require the purchase of safetyrelated material from the NTS approved vendors list (AVL). Instead, 

Paragraph H only discusses the control of the NTS AVL. 

As noted above, the inspectors were concerned that the purchase of materials 

and services cannot be adequately control led if safety-re1 ated materi a1 is 

purchased from a vendor not listed on the NTS AVL, and if the PO does not use 

the same description of material and services that was specified and approved 

in the NPR. 

a purchase order for cable to a vendor not listed on the NTS/PDS AVL, and the 

technical description in the PO was different from the NPR. 

procurement of the cable are discussed in Paragraphs 3.7 and 3.8.6 of this 

report. The inspectors identified the failure to establish adequate measures 

to control purchased materials and services as a nonconformance. 

(Nonconformance 95-01 -01 ) 

3.4.3 SOP No. 3, Revision 1, dated April 17, 1995, "Standard Receiving, 

Hand1 ing, Storage and Shipping," describes the methodology to assure PDS 

purchased materials and services are properly received, handled, inspected, 

The inspectors identified to NTS/PDS an instance where PDS issued 

Details of the 

4 

88 
and stored. The inspectors were concerned that the procedure did not provide 

guidance on accepting certificates of conformance (CoCs) (from manufacturers 

instead of distributors,) to detect the various fraudulent, or otherwise 

unacceptable products that has entered the nuclear industry and did not 

mention the numerous information notices issued by the NRC on this subject. 

For instance, during receipt inspection of cable, the PDS Receipt Inspector 

did not identify that the CoC was unacceptable because it was from a cable 

distributor instead of the cable manufacturer. PDS concurred with the 

inspectors and initiated NCAR 95-29 in which the corrective action recommends 

indoctrination on NRC information notices on fraudulent or otherwise 

unacceptable products. 

3.5 Control o f Measuri na and Test Eau i Dmersl 

During a January 25-27, 1995, inspection at Wyle Laboratories, (Wyle) 

Huntsville, Alabama, an NRC team observed some of the qualification tests 

being performed on the 4.16 kV switchgear manufactured and supplied by 

PDS/NTS. During the testing, the inspectors observed that NTS/PDS used a 

relay test set (RTS) which was not within its current calibration schedule. 

The calibration due date on the RTS had expired in December 1994. NTS/PDS 

used this RTS to check the calibration of the protective relays mounted on the 

breaker cubicles. The PG&E representatives stated that the purpose of the 

seismic testing was to specifically qualify the PDS retrofit drawout cubicles 

with Yaskawa breakers and not the relays mounted on the stationary cubicles. 

However, subsequent discussions indicated that PG&E had intended to 

seismically qualify the entire breaker cubicle including the instruments and 

relays mounted on the cubicle. However, due to the poor performance of the 

induction relays during seismic testing, the DCPP licensee decided to use 

sol id state protective relays. The PG&E representative informed the 

inspectors that the PDS switchgear successfully withstood the qualification 

tests at Wyle. 

During the current inspection, the inspector reviewed control of measuring and 

test equipment (M&TE) and concluded that P0S had an acceptable M&TE program. 

The inspectors noted that the M&TE was of the proper range, type, accuracy and 

to1 erance . In addition, MATE was cal i brated, uti 1 i zi ng standards traceabl e to 

the National Institute of Standards and Technology. 

Even though the M&TE program was generally acceptable, the inspectors observed 

a weakness related to the root cause analysis of a calibration problem. 

During a calibration check the calibration laboratory (GE Electronic Services) 

found that a AC/DC power supply (manufactured by Phenix Technologies) had 

exceeded its f 1.0% tolerance. The calibration of the DC portion required an 

accuracy of f 1.0% of full scale voltage. 

document that the accuracy of the power supply exceeded the tolerance. 

disposition, NTS/PDS stated that exceeding the f1.0% tolerance for the DC 

voltage range was acceptable because it was within the f3.0% accuracy listed 

in the H&TE master equipment list and took no further action. 

independently reviewing the root cause, the! inspectors observed that the power 

supply had four different power supply func:tions, each with its own 

calibration accuracy. 

was applicable to the AC power supply range even though it does not use this 

NTS/PDS initiated NCAR 95-14 to 

In the 

In 

PDS, had erroneously selected the f3.0% tolerance which 

5 

89 
function to take measurements. 

re-examined this matter and concurred with the inspectors that the original 

disposition of NCAR 95-14 was incorrect because it extracted the erroneous 

f3.0% tolerance from the M&TE master equipment list. After the discovery, 

NTS/PDS issued a revision to NCAR 95-14 requiring a review of previous jobs 

where the power supply with the incorrect accuracy was used to determine if 

there were any adverse affects. Additionally, before the conclusion of the 

inspection, PDS corrected the MATE master equipment list specifying the 

accuracy of this power supply as f1.046 of the full scale voltage and reopened 

NCAR 95-14, Revision 1, which initially identified the incorrect calibration 

of the power supply. 

At the inspectors request, NTS/PDS personnel 

The inspectors identified no problems other than a weakness in the 

investigation of the root cause of a problem. 

3.6 Pacific Gas and Electric ComDanv fPG &E) Audit of PDS 

The inspectors reviewed the results of an audit performed at PDS by PG&E on 

February 21-24, 1995. Replying to a question from the inspectors regarding 

the timeliness of the PG&E audit, the PG&E project engineer stated that PG&E 

wanted to audit PDS after the completion of the seismic qualification tests of 

the prototype retrofit breakers and before PDS commenced the manufacture of 

the safety-re1 ated swi tchgear. 

acceptable. 

implemented the NTS/PDS quality program during the fabrication of the specimen 

retrofit 4.16-kV switchgear that was tested at Wyle. PDS was contracted to 

manufacture a total of 105 identical Class 1E drawout cubicles with 350-MVAY 

4.16-kV circuit breakers rated for 1200 and 2000 Amperes which will meet or 

exceed the quality of the cubicles that successfully withstood the seismic 

qualification tests. 

The inspectors considered PG&E’s reply 

PG&E conducted the audit to verify that PDS had effectively 

The audit focused on the following areas: 

dedication and fabrication of stock material 

dedication of parts and components 

receiving inspection and test of Yaskawa circuit breakers 

welding and assembly of the NTS/PDS circuit breaker cubicles 

production testing of the NTS/PDS circuit breakers. 

The audit was very comprehensive and identified eight findings. 

dated April 12, 1995, NTS acknowledged PG&E’s audit findings and responded to 

them outlining actions planned to correct them. Actions taken to correct 

PG&E’s adverse audit findings included revising the NTS/PDS Quality Manual and 

the standard operating procedures (SOPS) that PDS uses to implement the 

program, and reassigning specific quality functions to NTS personnel. 

In a letter 

No problems were identified in this area. 

3.7 Review of Purchase Orde rs 

The team selected the following purchase orders (POs) issued by PDS to examine 

the implementation of the NTS/PDS Quality Program in areas related to the 

control of purchased materi a1 s. 

items are discussed in Section 3.8. 

The receipt inspections for these purchased 

6 

90 
0 

0 

0 

0 

0 

0 

PO No. 10472-HQ, dated May 18, 1995, to Hillman Fastener, 

Cincinnati, Ohio, for the supply of various hardware. 

PO No. 9819-HQ, dated February 17, 1995 to Monti, Cincinnati, 

Ohio, for various shapes of metallic components, fiberglassreinforced polyester angle type GP03. 

PO No. 10764-HQ, dated June 19, 1995, to Century Springs, Los 

Angeles, California, for 110 Type ASTM A 227 springs. 

PO No. 9826-HQ, to Copper & Brass Sales, Detroit, Michigan, for 

the supply of several pounds each of 3/8" x 3" rectangular ASTM B 

187, Alloy 110, full round copper bus bar, and 1/8" x 1-1/2" 

rectangular ASTM B 187, C 110, full round edge bus bar. 

PO No. 9818-HQ, dated February 17, 1995, to Central Steel 81 Wire 

Company for the supply of various sizes and shapes of bus bar 

material. 

PO No. 9384, dated December 20, 1994, to Anixter Southern, 

Cincinnati, Ohio, which stated, "600 Volt Tefzel (EFTE) Insulated , 

Wire 14 AWG, Single Conductor 14/19 black." 

SOP No. 2 is the applicable procedure for the preparation and issuance of 

these POs. The inspectors reviewed the implementation of this procedure. 

According to SOP No. 2, the first step is for PDS to prepare a "Nuclear 

Purchase Requisition," (NPR) with the technical description of the item. The 

second step is for NTS/PDS to review the NPR for the adequacy of the technical 

description of the item and, if acceptable, approve it. The next step is for 

PDS to transcribe the technical description of the item into the PO. 

The inspectors determined that the technical description of the cable in the 

NPR, which was reviewed and approved on December 15, 1994, was different than 

the description stated in the PO. 

retardant per [attached] description. " The attached description stated 

"Control cables shall have adequately sized stranded conductors, no less than 

X14 AWG, and at least 600 V insulation with highly flame retardant 

characteristics. Tefsel [sic] or special1,y flame retardant type SIS 

insulating and jacketing compounds of neoprene, hypalon, or flame retardant 

XLPE/XLPO are acceptable. 

wiring by the supplier." 

The NPR stated "#14 AWG, 600 V w/flame 

The cables shall be approved by PG&E prior to 

The description in the PO, which was prepared from the above NPR, stated "600 

VOLT TEFREL (EFTE) INSULATED WIRE 14 AWG SINGLE CONDUCTOR 14/19 BLACK." This 

PO was approved by NTS QA. 

effect that the items checked on the "Attached Purchase Order Requirements" 

sheet were an integral part of the PO, and that the material ordered under 

this purchase order was classified as a "critical item" and required a 

receiving inspection beyond the standard receiving criteria. This material is 

not to be "accepted" (or tagged as such) until the receiving inspection of 

critical items has been satisfactorily completed in accordance with NTS/PDS 

SOP No. 3. 

The PO also contained standard instructions to the 

7 

91 
In the "Purchase Order Requirements Sheet" attached to the PO, the annotated 

requirements stated "items must be new, not used, refurbished, altered or 

repaired and must be free from defects, all supplied items must be received in 

standard manufacturer packaging which is unopened and unaltered, all supplied 

items shall have a uniform configuration and appearance that is in accordance 

with any applicable manufacturer specifications or PDS drawings/purchase 

order." The PO required a certificate of conformance (CoC), signed by a 

vendor authorized individual other than someone in sales, marketing, or 

customer service, attesting to the quality of all supplied items. 

Other than inadequate measures to control purchased materials, which has been 

identified as a nonconformance in Paragraph 3.4.2, no further problems were 

identified in this area. 

3.8 Control of Purchased Mater1 a1 

The inspectors reviewed the process through which NTS/PDS controlled purchased 

materials. PDS personnel performed receipt inspections on material received 

using SOP No. 3, Revision 1, and NTS Work Procedure 60431-95N-1466-FAS, 

Revision 1, June 2, 1995, to inspect the material and document the results of 

the receipt inspection in NTS/PDS "Standard Receiving Report." Accepted 

material is identified and kept in a pending status awaiting detailed 

inspections to accept or reject it. 

to its designated permanent location. 

performed detailed inspections and dedicated the commercial-grade items for 

use in safety-related applications. During the detailed inspections, NTS 

inspectors establish a sample size depending on the lot size utilizing 

guidance provided by the Electric Power Research Institute (EPRI Report No. 

NP-7218, "Guideline For The Utilization of Sampling Plans For Commercial-Grade 

Item Acceptance (NCIG-19)." 

Acceptable material is then transferred 

NTS Quality Assurance inspectors 

3.8.1 Regarding the hardware received from Hillman Fasteners, PDS performed 

a receipt inspection and documented the observations in a "NTS/PDS Standard 

Receiving Report" dated June 20, 1995. The attributes verified were: 

a. The material received appeared new, uniform, unused, not altered, 

not tampered with, and not repaired or refurbished. 

b. The packing slip establishes traceability of the received material 

to the point-of-manufacture. 

c. The adequacy of the packaging, clean1 iness, identification/marking, 

workmanshi p and vendor documentation was a1 so veri f i ed and 

documented. 

The inspectors verified that for bolts, the PDS inspectors examined the 

marking on the head of the bolt, and used a Go-No-Go thread gauge to verify 

the correct size of the threads on the samples. 

to Massachusetts Material Research for special tests, such as chemical 

composition and hardness. 

NTS/PDS sent some specimens 

8 

92 
3.8.2 

procured from Monti, the critical characteristic was dielectric strength and 

the failure mode for these components was identified as pinholes. 

acceptance is verified when the installed components successfully withstand a 

dielectric strength voltage of 19+.5 kV for 60 seconds. 

For various shapes of fiberglass-ireinforced polyester angle type GP03, 

The 

3.8.3 For the steel springs received from Century Springs, the acceptance 

criteria is provided in Paragraph 4.1.3.5, "Helical Springs" of NTS Procedure 

60431-95N-1466-Bar, Revision 0, "Receipt Inspection and Sampling Procedure for 

Safety Related Bar Stock and Components For the PG&E Units." The technical 

specifications for the 110 steel springs stated in the PO are: outside 

diameter: 0.625", inside diameter: 0.510", length: 3.00", 15 coils, 9.16 

pounds per inch spring rate, 1.496" deflection, closed ends, zinc plated. 

When the springs are received, PDS quality control inspectors verified the 

overall length of the spring. NTS quality assurance personnel used Procedure 

No. 60431-95N-3, Revision 1 to dedicate the springs. The procedure focusses 

on the compression strength on a random sample of the springs to provide 

assurance that the springs wi 11 have equi wal ent performance with those 

specimens that successfully withstood the seismic tests. 

3.8.4 

PDS personnel verified that the dimensions for each piece of bar stock met the 

PO requirements, measured the resistance, and hardness (Rockwell B [HRB 751). 

NTS identified the critical characteristic of the bar stock to be resistivity, 

and NTS established the acceptability of the round copper bus bar to ASTM B 

187, by correlating the hardness numbers to the resistivity. 

personnel measured the hardness of the copper and compared them to the 

acceptance values establ i shed by NTS. 

For the various shapes of copper received from Copper & Brass Sales, 

NTS quality 

3.8.5 For the various bus bar components received from Central Steel and 

Wire Company, PDS/NTS determined the acceptability by measuring the hardness 

and comparing them with predetermined values. 

3.8.6 PDS received 600 Volt Tefzel 14 AWG, single conductor 14/19 black 

cable from Anixter Southern with a certificate of conformancc :"-") from Basic 

Wire and Cable, Chicago, Illinois. The CoC was addressed to Anixter Southern, 

and stated "It is herewith certified that all articles in the quantities as 

called for in your purchase order No. 850--120498-861 are in conformance with 

requirements, specifications and drawings 1 isted on that order." 

NTS/PDS "Standard Receiving Report" for the cable, the PDS receipt inspector 

identified no unacceptable findings and the NTS quality assurance person noted 

that the vendor was not on the NTS approved vendors list (AVL). 

3.4.2 of this report, the inspectors identified a nonconformance relative to 

the inadequate control of purchased material. 

In the 

In paragraph 

The inspectors were also concerned that SOP No. 2 does not provide guidance to 

distinguish between a distributor and an equipment manufacturer and does not 

provide sufficient guidance on scrutinizing the authenticity of certificates 

of conformance (CoCs). NTS/PDS initiated NCAR 95-29 and included in it a 

corrective action to revise SOP No.2 and to indoctrinate the staff so that 

personnel who issue purchase orders can precisely define the type of CoCs that 

are acceptable to help quality control inspectors to recognize genuine CoCs. 

9 

93 
The inspectors identified that POS procured cable from a vendor not listed on 

the NTS AVL and accepted a CoC from a vendor who was a distributor and not the 

manufacturer of the cable. 

Other than inadequate measures to control purchased materials and services, 

which has been identified as a nonconformance in paragraph 3.4.2, no further 

problems were observed in this area. 

3.9 Review of the Circuit Breaker Dedication 

PDS purchased a total of 132 SF6 Rotary-Arc Yaskawa circuit breakers. 

these, eight 2000-Ampere (A)-rated and ninety seven 1200-A rated circuit 

breakers are intended to perform safety-related functions at DCPP. 

the remaining breakers for seismic testing, manufacturing non-Class 1E 

cubicles for DCPP, and spares. 

Of 

PDS used 

The inspectors reviewed the following NTS Procedures to assess the 

effectiveness of the implementation of the NTS/PDS quality program in the 

areas of electrical design, receipt inspection, and production testing 

requirements and to assess their conformance with ANSI/IEEE recommended 

practices for circuit breakers. 

0 NTS Procedure No. 60431-95N, Revision 1, of May 1, 1995, 

"Dedication/Acceptance Basis for Class 1E Retrofit Circuit 

Breakers, 4 kV, 350 MVA for Diablo Canyon Power Plant Units 1 & 2 

Pacific Gas & Electric Company." 

0 NTS Procedure No. 60431-95N-1466-RIY Revision 1, "Receipt 

Inspection/Test Procedure for Yaskawa Circuit Breakers, Type: 5GYB1-1200-350, 5GYB-2000-350." 

0 NTS Procedure No. 60431-95N-1466-CPTY Revision 2, "Conversion 

Production Test Procedure for PDS SF Retrofit Circuit Breaker, 

Types: 5GYB1-1200-350 and 5GYB1-2000-350. I' 

The inspectors reviewed the adequacy of the receipt and production test 

programs with respect to their demonstrating the functional capability of 

selected component parts of the procured Yaskawa SF, circuit breaker. The 

breaker's expulsion membrane, and pressure switch were selected for review. 

PG&E indicated that the essential purpose of these components is to maintain 

the integrity of the SF, insulating medium for the circuit breakers' main 

contacts. 

dielectric test. NTS/PDS performs dielectric tests on the circuit breaker as 

part of the final production tests. 

would be repeated as part of Diablo Canyon site receipt and periodic test 

programs. 

This purpose is demonstrated by performance of an insulation 

PG&E also indicated that dielectric tests 

The inspectors found the program procedures to be consistent with ANSI/IEEE 

recommended practices and PG&E's requirements. 

concerns in this area. 

The inspectors identified no 

10 

94 
3.10 Observa tion of Assemblv Activities In Prowess 

The inspectors toured the PDS fabrication areas where the DCPP switchgear was 

being assembled. During the tour, the inspectors observed incoming material 

staging, and storage areas. In these areas, the inspectors examined Yaskawa 

SFe circuit breakers, completed receipt inspection/test data sheets for the 

Yaskawa circuit breakers, prototype enclosures that had been used for seismic 

testing which contained the adapted Yaskawa SF, circuit breakers for use at 

Di ab1 o Canyon. 

During the tour, the inspectors noted that electrical control cables on the 

prototype enclosures were routed next to sharp edges and in the vicinity of 

moving parts. The inspectors expressed concern that this routing could, over 

time, cause chafing and failure of the cable's insulation system. 

response, PDS stated that the control cables when installed in the production 

enclosures would be reconfigured, routed, and tie-mounted to the enclosure 

such that the cable's insulation system will not be exposed to chafing from 

sharp edges or moving parts of the converted Yaskawa SF, circuit breaker. 

Even though a completed production enclosure (with the proposed cable routing 

installed) was unavailable for inspection, the inspectors concluded that the 

proposed reconfiguration, routing, and tie-moupting is feasible and is 

standard industry practice, and can be performed such that the cable and the 

cable's insulation system will not be subjected to conditions or stresses for 

which they are not designed. In addition, given the passive nature of cable 

systems, the inspectors concluded that the proposed control cable 

reconfiguration, routing, tie-mounting (although different from that used in 

the prototype enclosures) will not affect the seismic qualification of either 

the prototype or production enclosures. 

In 

The inspectors examined the welds on two drawout cubicles and observed that 

the size, length and location of the welds met the drawing requirements. The 

inspectors reviewed the qualifications of the weld procedure and the welders 

and determined them acceptable. 

No. 11A, "Procedure for Structural Welding and Weld Inspection," and observed 

that the procedure imp1 ied that welders themselves could evaluate reje.:ted 

welds. 

inspector shall disposition welds as 'accept', 'reject' or 'rework', record 

the results, sign and date the Weld data sheet." The NTS/PDS representatives 

concurred with the inspectors that the deletion of "the welder" from the 

sentence of this paragraph would minimize confusion. 

NTS/PDS issued NCAR 95-29 to delete "the welder" from the procedure. 

The inspectors reviewed the NTS/PDS procedure 

Paragraph 1V.D of the procedure states in part, "The welder and 

On June 22, 1995, 

Other than a weakness in the NTS/PDS procedure No. 11A, which was being 

corrected, the inspectors did not identify any unacceptable findings in this 

area. 

3.11 Review of Training Records 

SOP No. 10 describes the control of qualification activities for personnel at 

PDS. 

following weaknesses: 

The inspectors reviewed the training documents and observed the 

11 

95 
0 The agenda of the training was not deta 

0 It was not clear if the insDectors were 

led. 

trained on what 

constituted a valid certifikate of conformance (CoC) or what was 

required for a CoC to be valid. 

a The topics that were used to discuss the numerous ways to detect 

fraudulence were not documented, and the generic communications 

that the NRC had issued on fraudulence, (e.g. Bulletin 88-10) were 

not mentioned. 

NTS/PDS informed the inspedtors that even though they had discussed these 

issues they had not documented them in an auditable form. 

conclusion of the inspection PDS initiated NCAR 95-29 to document the topics 

discussed during training sessions. 

Before the 

The inspectors informed PDS personnel that NRC issued Generic Letter 89-02 

stressing the importance of personnel performing safety-related activities 

being trained in the detection of fraudulent material. There were no records 

at PDS to indicate that such information was collected and used in training 

sessions on fraudulent material known to have been previously supplied to the 

nuclear industry to educate the individuals on the significance of CoCs, to 

stress the importance of verifying the authenticity of CoCs, and to enable the 

inspection personnel to detect fraudulent or otherwise unacceptable material 

during receipt inspections. 

enhance training requirements will be included in NCAR 95-29 and records will 

be developed to reflect the training. 

NTS/PDS informed the inspector that action to 

Other than some weakness in the depth of training, the inspectors did not 

identify any unacceptable findings in the training area. 

4.0 PERSONS CONTACTED 

Name Title 

Power Distribution Services. Inc. (PDS) 

t * J.E. Bachmann Qual i ty Control Inspector 

t * J.L. Bachmann Assistant to the President 

t * J.P. McCloy Vice President Operations 

t * W.A. McCloy Pres i dent 

t * T. Miracle P1 ant Manager 

t * D.R. Robling Manager, Technical Services 

* E.J. Kuehne Senior Vice President 

National Technical Services (NTS) 

t * F.W. Bean Qual i ty Representative 

t * W.E. Copeland Quality Technical Specialist 

t * J.E. Dozier Qual i ty Manager 

t * D.T. Grand Site Engineer 

12 

96 
t * M.E. Lilly 

t D.R. Nichaud 

t * M.P. Saniuk 

Qual i ty Representative 

Di vi si on Program Manager 

Engineering Manager 

Pacific Gas and Electric ComDanv (PG&E) 

t* 

t 

M. Basu 

R.A. Carve1 

Project Engineer 

Suppl ier Assessment Auditor 

* Individuals who attended the entrance meting on June 19, 1995. 

t Individuals who attended the exit meeting n June 21, 1995. 

13 

97 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-0001 

July 5, 1995 

Mr. Mark Van Sloun 

Vice President and General Manager 

Rosemount Nuclear Instruments, Incorporated 

12001 Techno1 ogy Drive 

Eden Prairie, MN 55344 

SUBJECT: NRC INSPECTION NO. 99900271/95-02 

Dear Mr. Van Sloun: 

This letter transmits the report of the inspection conducted by Mr. Stephen 

Alexander of this office and Mr. S.V. Athavale of the Instrument and Control 

Branch from April 5 to 7, 1995, at your facilities at Eden Prairie and 

Chanhassen, Minnesota. At the conclusion of the inspection, the findings were 

discussed with you and the members of the Rosemount staff identified in the 

enclosed report. 

to the inspection, your staff provided additional information relevant to the 

inspection that is documented in the report. 

In telephone conversations and telefax messages subsequent 

Areas examined during the inspection are identified in the report. 

included (1) an assessment of the validity and comprehensiveness of the 

methods by which Rosemount researched its records and determined the serial 

numbers of the nuclear transmitter sensor modules that potentially contained 

Monel isolators (initially from the lot used in the failed transmitters from 

St. Lucie) and the customers to whom these modules or transmitters containing 

these modules were supplied, (2) a review of Rosemount’s supplemental measures 

taken to determine if the isolator lots identified in the initial search were 

the only Monel isolators to be inadvertently used in nuclear transmitters, 

(3) an examination of the isolator assembly manufacturing proceqq and the 

circumstances surrounding the original error in selecting Monel rbil strip 

stock to make 1152/3/4 foil disc assemblies, (4) an examination of the 

circumstances surrounding the identification and documentation of the apparent 

error and the ultimate inappropriate disposition of the discrepancy report, 

(5) a review of the quality control measures subsequently established that 

would minimize the probability of such errors, and (6) a review of testing and 

root cause analysis thus far and of design information relating to the 

exclusion of Monel from applications with a high hydrogen concentration 

environment. Within these areas, the inspection consisted of selective 

examinations of procedures and representative records, interviews with 

personnel, and observation of activities in progress. 

They 

During this inspection, we determined that the implementation of your quality 

assurance (QA) program failed to meet certain NRC requirements. The 

nonconformance cited was for failure to prevent inadvertent use of certain 

nonconforming parts and failure to take adequate corrective action by 

inappropriate disposition of a discrepancy report identifying the use of the 

nonconforming parts. 

requirements are identified in the enclosures to this letter. 

The specific findings and references to the pertinent 

98 
THIS PAGE LEFT BLANK INTENTIONALLY. 
Mr. Mark Van Sloun -2- 

Please provide us within 30 days from the date of this letter a written 

statement in accordance with the instructions specified in the enclosed Notice 

of Nonconformance. 

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

this letter, its enclosures, and your response will be placed in the NRC 

Public Document Room (PDR). 

include any personal privacy, proprietary, or safeguards information so that 

it can be placed in the PDR without redaction. 

necessary to include such information, you should clearly indicate the 

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

the public. 

To the extent possible, your response should not 

However, if you find it 

The responses directed by this letter and the enclosed Notice are not subject 

to the clearance procedures of the Office of Management and Budget as required 

by the Paperwork Reduction Act of 1980, Pub. L. No. 96-511. 

Sincerely, 

n Qw% Robert M. allo. Chief 

Special Inspection Branch 

Division of Technical Support 

Office of Nuclear Reactor Regulation 

Docket No. 99900271 

Enclosures: 1. Notice of Nonconformance 

2. Inspection Report No. 99900271/95-02 

cc w/encl: Paul Blanch 

135 Hyde Road 

West Hartford, CT 06117 

Ernest Hadley, Esquire 

414 Main Street 

Post Office Box 3121 

Wareham, MA 02571 

100 
REPORT NO.: 

ORGANIZATION : 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY: 

INSPECTION DATES: 

LOCAT ION ( S ) : 

LEAD INSPECTOR: 

OTHER INSPECTORS : 

REVIEWED BY: 

APPROVED BY: 

U. S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF TECHNICAL SUPPORT 

99900271 /95-02 

Rosemount Nuclear Instruments , 12001 Technology Drive 

Eden Prairie, Minnesota 55344 

J. Valley 

Quality Assurance Manager 

Incorporated 

Rosemount manufactures and suppl ies nuclear qual ified 

pressure and differential pressure transmitters to 

most of the commercial and government nuclear 

facilities market. 

April 5 through 7, 1995 

Engineering offices at the Eden Prairie Facility; 

sensor cell manufacturing facility at Chanhassen, 

Minnesota 

vendor Inspection Section (VIS) 

Special Inspection Branch (TSIB) 

Instrument and 

Division of 

S.V. Athavale 

Enclosure 2 

101 
NOTICE OF NONCONFORMANCE 

Rosemount Nuclear Instruments, Incorporated 

Eden Prairie, Minnesota 

Docket No. 99900271 

Report No. 95-02 

Based on the results of an NRC inspection conducted on April 5-7, 1995, it 

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

NRC requi rement s . 

A. Criterion XV, "Nonconforming Material, Parts or Components," of 10 CFR 

Part 50, Appendix B, states: "Measures shall be established to control 

materials, parts, or components which do not conform to requirements in 

order to prevent their inadvertent use or installation. These measures 

shall include, as appropriate, procedures for identification, documentation, disposition, and notification to affected organizations." 

Criterion XVI, "Corrective Action," of 10 CFR Part 50, Appendix B, 

states, in part: "Measures shall be established to assure that 

conditions adverse to quality, such as failures, malfunctions, 

deficiencies, deviations, defective material and equipment, and 

nonconformances are promptly identified and corrected." 

Contrary to the above, measures established by Rosemount to control 

certain nonconforming material did not prevent its inadvertent use. 

In addition, action taken to correct a condition adverse to quality was 

inadequate in that Discrepancy Report 491585, written October 30, 1989, 

was dispositioned inappropriately by the Material Review Board. The 

discrepancy report documented the use of strip stock material that was 

not in accordance with the bill of materials to make Lot 20 of Part No. 

01153-0252-0042 disc assemblies. The Material Review Board inappropriately dispositioned the discrepancy by directing that the strip stock 

part number and lot number on the traveller be corrected. As a result, 

Lot 16 of C10181-0014 Monel foil strip stock, documented as having been 

used to m.ke Lot 20 of the disc assemblies, was actually so used, yet 

the traceability data on the traveller was erroneously changed to read 

Lots 23 and 24 of Part No. CO9851-0011 (316L stainless steel foil strip 

stock) (95-02-01). 

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

Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555 

with a copy to the Chief, Special Inspection Branch, Division of Technical 

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

the letter transmitting this Notice of Nonconformance. This reply should be 

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

for each nonconformance (1) the reason for the nonconformance, or if 

contested, the basis for disputing the nonconformance, (2) the corrective 

steps that have been taken and the results achieved, (3) the corrective steps 

that will be taken to avoid further noncompliances, and (4) the date when your 

corrective action will be completed. Where good cause is shown, consideration 

will be given to extending the response time. 

Dated at Rockvi 11 e, Mary1 and 

this 5th day of July, 1995 

Enclosure 1 

102 
1. SUMMAR' F INSPECTION FINDINGS 

The 

0 

0 

0 

nspection basis consisted of the following: 

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

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

Resulations (10 CFR Part 50) 

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

Rosemount Quality Assurance (QA) Program Documents and Procedures 

The inspectors reviewed historical documents relating to the manufacturing 

process used for replacement sensor cells provided to licensees for the fill 

oil loss problem. The inspectors examined Rosemount's factory procedures and 

process for receiving, inspecting, and punching metal foil strip stock to be 

used for sensor cell isolator diaphragms; handling, cleaning, and welding of 

discs to weld rings; and testing of finished isolator assemblies. 

inspectors interviewed engineers, QA and qual ity control (QC) personnel, and 

factory workers (instrument builders) to gather data relating to past 

manufacturing and QA/QC errors. 

documentation; Rosemount's methodology of identifying, scoping, and bounding 

the lots of suspect transmitters; Rosemount's efforts to obtain the services 

of the Southwest Research Institute (SwRI); and Rosemount's efforts to provide 

required support to its affected customers. 

The 

The inspectors a1 so reviewed foll owup 

The inspectors determined that Rosemount's methodology to bound the problem 

with the affected transmitters acceptable for transmitters made with isolators 

from the same lot as those that failed at St. Lucie. The inspectors concluded 

that production control measures subsequently establ ished by Rosemount should 

have precluded such errors since the incident in question and should continue 

to do so in the future. The inspectors noted that Rosemount was supporting 

replacement of the suspect sensor cells on a first priority basis. 

stepped up production and diversion of resources to this project from those 

with less urgent needs, Rosemount estimated that lead time for v~nlacement 

units could be reduced from its normal 12-week period to as low a5 2 weeks. 

With 

1.1 Violations 

None 

1.2 Nonconforman- 

(95-02-01) Contrary to the requirements of Criterion XV of 10 CFR Part 50, 

Appendix B, Rosemount's measures to control nonconforming parts did not 

prevent the inadvertent use of certain transmitter sensor cell isolator 

diaphragms of the wrong material. Contrary to the requirements of Criterion 

XVI, Rosemount took inadequate corrective action in that the disposition of a 

discrepancy report, written October 30, 1989, that identified the use of the 

wrong material cited above, was inappropriate in that the disposition stated 

was to change the part number on the traveller rather than verify what 

material was actually used and to take the steps necessary to capture any 

2 

103 
incorrect material and prevent its use in transmitters designed for nuclear 

safety-related service (hereinafter referred to as nuclear transmitters). 

(See Section 3.5 of this report.) 

2.0 STATUS OF PREVIOUS INSPECTION FINDINGS 

No previous findings were reviewed during this inspection. 

3.0 INSPECTION FINDINGS AND OTHER COMHENTS 

3.1 Backsround 

3.1.1 Identifying Event 

‘- . 

On November 22, 1994, Florida Power and Light Corporation’s (FP&L’s) St. Lucie 

Unit 1 Nuclear plant (St. Lucie), suffered an inadvertent safety injection 

(SI) event when two of the four Rosemount Model 1153 nuclear pressurizer 

pressure transmitters failed with high outputs during repressurization of the 

reactor coolant system following a full depressurization for an outage. The 

two high outputs removed the manual SI block (imposed during shutdown) and the 

two normal outputs, transmitting the actual (low) system pressure, caused SI 

initiation. Gas entrapped in the fill oil cavities of the transmitter sensor 

cells was the apparent cause of the failures. 

Rosemount undertook a root cause analysis, and also performed examination and 

testing of the failed transmitters using the services of SwRI which had the 

To address this concern, 

capabi 7 i ty of hand1 i ng potent i a1 

3.1.2 SwRI Tests and Analysis 

The tests and analysis indicated 

cells was pure, diatomic (molecu 

process inleakage, (3) there was 

y radiologically contaminated material. 

that (1) the gas entrapped in the sensor 

ar) hydrogen, (2) there was no evidence of 

no evidence of fill oil decomDosition. and 

(4) the failed transmitters had Monel Alloy 400 isolating diaphragms ibstead 

of Type-316L stainless steel (316L) isolating diaphragms that are supposed to 

be used for the nuclear grade (Types 1152, 1153, and 1154) transmitters. 

3.1.3 Postulated Failure Mechanism 

The SwRI test results suggested the fol 1 owing postul ated fai 1 ure mechanism: 

The entrapped hydrogen gas came out of solution upon depressurization of the 

plant. Upon repressurizati on, the coalesced hydrogen bubbles di spl aced f i 11 

oil from the process side cavity of the sensor cell into the center chamber, 

deflecting the sensing diaphragm (capacitor plate) and causing the high output 

signal. In addition, the hydrogen dissolved in the fill oil altered the fill 

oil dielectric constant, also causing a high output signal. The postulated 

source of the entrapped molecular hydrogen was recombination of monatomic 

hydrogen that most likely diffused through the Monel isolating diaphragms. 

Monel is known to be highly permeable to hydrogen. 

externally produced monatomic hydrogen was not yet conclusively established at 

the time of preparing this report. 

The source of the 

However, the two prevailing theories are 

3 

104 
(a) that the monatomic hydrogen was in the process fluid (reactor coolant) in 

the pressurizer as a minority constituent of the relatively large amount of 

hydrogen expected to be in the pressurizer of a pressurized water reactor 

plant such as St. Lucie, or (b) that the monatomic hydrogen was generated on 

the isolator surface as a product of general corrosion of the wetted metal 

surfaces of the isolator and weld ring, and/or galvanic corrosion in a cell 

formed of the Monel isolator diaphragm (cathode) and the stainless steel weld 

ring with the coolant acting as electrolyte. 

these postul ated phenomena can occur. 

It is also possible that both of 

3.1.4 Rosemount Internal Investigation 

Prompted by SwRI’s finding that the entrapped gas was hydrogen, and confirmed 

by the finding that the St. Lucie isolator diaphragms were made of Monel, 

Rosemount’s in-house investigation revealed that due to a manufacturing error 

compounded by a QA/QC error in 1989, sensor cells having isolation diaphragms 

manufactured using Monel instead of 316L were used for as many as 451 Type 

1152, 1153, and 1154 nuclear grade transmitter sensor modules of Range Codes 6 

through 10. The modules were being supplied to nuclear utilities or being 

used to repair transmitters returned by utilities for various reasons, 

including correcting the fill oil loss prolplem (as was the case with St. 

Lucie). Rosemount identified affected manufacturing lots of the modules, and 

issued a notification to all affected licensees or purchasers pursuant to 10 

CFR 21.21(b) on March 21, 1995. On March 22, 1995, the NRC issued Information 

Notice 95-20, “Failures in Rosemount Pressure Transmitters Due to Hydrogen 

Permeation Into the Sensor Cell.” Appendix A to this report gives the names 

of the affected organizations in the Rosemount Part 21 notification, and 

Appendix B to this report contains the chronology of events from the incident 

at St. Lucie until the inspection documented in this report. 

3.1.5 Regulatory Response Groups ’ Responses 

On April 4, 1995, in support of the NRC’s examination and tracking of this 

issue, and in preparation for this Rosemount inspection, the inspector 

reviewed the responses to this issue submitted by the Regulatory Response 

Groups (RRGs) of the Jestinghouse Owners Group (WOG), the Boiling Water 

Reactor Owners Group (BWROG), the Babcock 81 Wilcox Owners Group (B&WOG), and 

the Combustion Engineering Owners Group (CEiOG). 

submitted in reply to NRC activation of the RRGs and questions on affected 

Rosemount sensor modules and transmitters. 

on assumptions as yet unconfirmed regarding the primary source of entrapped 

hydrogen and its solubility in the transmitter fill oil. The B&WOG response 

in particular pointed out that one explanation why transmitters had not yet 

failed in a similar application to St. Lucie (i.e., pressurizer pressure 

transmitters at Florida Power Corporation’s (FPC’s) Crystal River Plant) was 

that Crystal River had much longer instrument lines (and hence, longer 

hydrogen diffusion lengths) on its pressurizer pressure instruments than those 

on the St. Lucie pressurizers. 

explanation for a failure of the type in question not yet occurring at Crystal 

River is the assumption that the coolant in the pressurizer is the primary 

source of monatomic hydrogen operative in this failure mode. In addition, 

without documenting its basis, this same utility asserted that 500 psig was 

These responses were 

Some safety assessments were based 

The inspector noted that underlying this 

4 

105 
the pressure threshold for hydrogen coming out of solution in major pressure 

excursions or transients during plant operation. 

addressed with Rosemount during the inspection. 

These questions were 

3.2 Entrance and Exit Meetinqs 

During the entrance meeting for this inspection, held April 5, 1995, at 

Rosemount’s Eden Prairie, Minnesota, facility, the inspectors met with 

Rosemount management and discussed the scope and objectives of the inspection. 

During the exit meeting on April 7, 1995, with Rosemount management, the 

inspectors summarized the inspection findings. 

3.3 Inspection Details 

3.3.1 Rosemount Root Cause Analysis 

3.3.1.1 SwRI Results 

nor 

oi 1 

the 

the 

3 -3 

The inspectors reviewed Rosemount’s root cause analysis. SwRI test reports 

which were issued on March 21, 1995, indicated that SwRI performed analyses to 

identify the types of gases trapped under the diaphragm, determined moisture 

content of the oil, and performed electron micrography of both inner and outer 

surfaces of high pressure diaphragms showing the center and middle areas 

including weld beads, and performed dark field micrography of cross sections 

of both diaphragms. In addition, electron dispersive spectroscopy (EDS), 

semi-quantitative elemental analysis of high and low pressure side diaphragms 

along with the weld ring material and inductively coupled argon plasma (ICP) spectrography on high pressure side diaphragms were performed to identify and 

analyze the materials. Results of the SwRI tests were that (1) gas trapped 

under the isolator diaphragm was identified to be pure molecular hydrogen and 

no other gases were found; (2) moisture content was not found in the fill oil 

other evidence of process in-leakage, (3) there was no evidence found of 

decomposition, and (4) the elements (and their quantities) identified in 

weld rings were consistent with 316L stainless steel and the elements (and 

r quantities) in the diaphragms were consistent with Monel Alloy 400. 

1.2 Source of Hydrogen 

There are several postulated sources of the entrapped hydrogen. 

prevalent theories postulate an external source of the hydrogen entrapped in 

the transmitters: (1) hydrogen in the process fluid (coolant) diffusing or 

leaking into the sensor cell and (2) hydrogen generated by corrosion 

reactions. In addition, the possibility of internally generated hydrogen by 

fill-oil decomposition reactions or reactions with coolant that might have 

leaked into the transmitters were investigated by SwRI for Rosemount. 

The two most 

If the primary source of monatomic hydrogen available for diffusion through 

the Monel isolating diaphragm was from the process fluid, then the first 

occurrence of this particular failure being in transmitters exposed to coolant 

from the pressurizer is consistent with the expected hydrogen content of that 

5 

106 
coolant. 

a minority constituent of the diatomic (molecular) hydrogen in the coolant. 

The molecular hydrogen (H2) comes from chemical addition to the coolant and 

radiolytic decomposition of the coolant water. In this case, coolant in the 

pressurizer and especially in the steam space during operation would be the 

only plant location expected to contain sufficient excess hydrogen to cause 

significant hydrogen diffusion and resulting transmitter failure. Therefore, 

transmitters with Monel isolating diaphragms in other locations in the plant 

with less hydrogen or none at all in the process fluid, would not be expected 

to fail as a result of process hydrogen intrusion. 

source of the entrapped hydrogen is also consistent with the single reported 

instance of failures thus far because the rate of hydrogen diffusion through 

Monel is a function of temperature (relatively hotter at or near the 

pressurizer) as well as external hydrogen concentration and the time to 

failure is a function of the hydrogen diffusion length through the process 

fluid in instrument line dead legs. 

Monatomic hydrogen would be expected to be present in the coolant as 

The coolant as the primary 

However, if the primary source of monatomic hydrogen operating in this failure 

mode is from galvanic action and general corrosion, both producing hydrogen 

atoms on the surface of the Monel isolator, then other locations in the plant 

could be susceptible. 

contact of dissimilar metals - in this case, the Monel diaphragm and the 316L 

weld ring - in the presence of an electrolyte, the coolant. Such a galvanic 

cell would produce about 0.5 volt, with the Monel being the cathode. 

Rosemount’s reported commercial experience with noticeable detrimental 

hydrogen diffusion through Monel has primarily been in environments with 

relatively high hydrogen concentrations as opposed to the relatively small 

amounts that may be produced by a galvanic cell and by general corrosion. 

The galvanic action could result from the intimate 

The inspectors dizcussed with Rosemount the possibility that as moisture may 

intrude into the fill oil at a high pressure, a chemical reaction of water 

molecules with the silicon oil would yield pure hydrogen and silicon dioxide. 

The reaction would be driven to completely consume the limiting reactant, the 

water. Tests for moisture therefore should include looking for silicon 

dioxide precipitate on internal surfaces, particularly the di pnhr3gm. The 

inspectors inquired whether SwRI testing revealed any amount ot silicon 

dioxide deposits in oil or on the inner surface of the diaphragm. Prompted by 

this question, Rosemount reported having contacted SwRI during this inspection 

and discussed the issue with the SwRI test technician who had inspected the 

internals of the sensor cells. 

Rosemount that although SwRI did not look specifically for silicon dioxide; if 

present, it would most likely have shown up in the tests conducted by SwRI. 

Rosemount considered this further evidence (and the inspectors agreed) that 

moisture did not penetrate into the oil in the failed St. Lucie transmitters. 

Therefore, the pure hydrogen found in the oil was not likely to have been 

generated internally by reaction of the oil with moisture. 

The SwRI technician was reported to have told 

SwRI’s dark field micrographs indicated some small fractures in the interior 

of the weld in addition to the heat-affected grain boundary zones. 

corrosion, other fractures, or other indications of possible leakage paths 

were found. 

provided a path for hydrogen leakage into the fill oil, but Rosemount 

No 

The inspectors questioned whether these fractures might have 

6 

107 
determined that such a leakage path was highly unlikely because the weld 

fractures did not appear to penetrate the weld. 

fractures could provide an in-leakage path, there likely would have been 

evidence of moisture intrusion and certainly a penetrating crack would also 

provide a leakage path out. Rosemount concluded (and the inspectors agreed) 

that leakage out would be inconsistent with the amount of entrapped gas being 

sufficient to cause the deformation or distention of the isolator diaphragms 

observed on the failed transmitters at St. Lucie. 

Furthermore, if weld 

The SwRI testing also revealed that since refurbishing of the transmitters, 

specific quantities and material properties of the fill oil of the failed 

transmitter had changed very little, compared to a fresh sample of the fill 

oil, indicating that fill oil breakdawn did not occur. Therefore, SwRI and 

Rosemount concluded that the hydrogen found in the oil was not a result of the 

fill oil breakdown. 

3.3.1.3 Comparison with Similar Applications 

With regard to the few cases in which transmitters identified with Monel 

diaphragms did not fail, such as the third transmitter on the affected 

pressurizer at St. Lucie or transmitters at Crystal River, Unit 3, which were 

also exposed to operating conditions similar to the failed transmitters at St. 

Lucie, Rosemount stated that if transmitters in similar situations did not 

exhibit signs'of failure, that dots not mean that they may not be close to 

failure. Rosemount further explained that there may be other factors specific 

to individual installation configuratisns which influence the failure rates of 

transmitters otherwise exposed to similar operating conditions. An example of 

these factors, cited by Florida Power Corporation in its RRG response, would 

be the longer instrumt lines at Crystal River as compared to the corresponding shorter lines at St. Lucie. 

3.3.1.4 Factors Affecting Failure Mode and Probability 

The inspectors inquired whether Rosemotfnt had generated a mathematical model 

to simulate conditions of hydrogen permeation in the sensor cell to provide 

information about the direction and amount of signal shift for each ot various 

connection configurations and ahut any precursors of fai 1 ure . responded that development of mathematical models to predict transmitter 

failure probability was not being considered at the time of the inspection. 

With respect to solubility of hydrogen (both H and H,) in the Dow-Corning 704 

silicone-based oil that Rosenaount uses as transmitter fill oil, Rosemount 

stated that Dow-Corning had solubility data for helium in this oil, but not 

for hydrogen. 

Rosemsunt 

However, Rosemount had performed an analysis to determine the direction of 

transmitter drift for various transmitter configurations. 

that for pressure (absolute or gauge) transmitters, entrapped gas wsuld always 

cause transmitter output to fail high because (1) gas displacement of fill oil 

during repressurization would cause hyckalic &flection of the sensing 

diaphrqpn (in the center of the cell) in a direction tbet would produce hi 

output, and (2) hydrogen in the dielectric fill oil would reduce the effective 

dielectric constant of the oil, producing the sgnre effect. 

The conclusion was 

In differential 

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pressure transmitters, the sign or direction of the error introduced by 

entrapped gas would depend on which side of the sensor cell had more hydrogen 

and on the relative pressures. 

differential pressure cells having isolators of different materials on either 

side, Rosemount stated that irt its manufacturing process, a single lot of 

sensor cells may be built using isolator assemblies from different lots (as 

was the case with the mixed lots), but that individual cells are supposed to 

be built from only one lot. In other words, the last isolator assembly in a 

lot would not be welded to one side of a cell; rather use of a new isolator 

assembly lot would be started and the single isolator scrapped. This practice 

would preclude having a cell with a Monel isolator on one side and another 

material on the other. Therefore, Rosemount concluded that different isolator 

materials in the high and low pressure sides of a differential pressure 

transmitter sensor cell was not a factor to be concerned with. 

To address the question of the possibility for 

3.3.1.5 Selection of Isolator Diaphragm Materials 

The inspectors also reviewed design information relating to selection of 316L 

for hydrogen environment applications and to the basis for excluding Monel 

from those environments. The inspectors concluded that although 3161 was not 

selected for nuclear applications on the basis of low hydrogen permeability, 

Monel was normally excluded from high hydrogen environments because of known 

high hydrogen permeability. The inspectors determined that Rosemount has 

considerable commercial experience with Monel in hydrogen environments and 

this experience compelled Rosemount to avoid using Monel in such applications. 

Rosemount's problems with Monel in hydrogen environments is documented in 

Rosemount Technical Report 282106, "Transmitter Damage by Hydrogen Generation 

and Diffusion," dated March 10, 1982. In addition, although it is not certain 

what references, research papers, or other information, formed the basis for 

the original design decision not to use Monel in hydrogen applications, the 

current design engineer has several such references in his files which confirm 

the commonly held notion or conventional wisdom (also cited in Rosemount 

Report 282108) that nickel alloys tend to have a high hydrogen permeability. 

Rosemount 's col1 ect i ve know1 edge on corrosion (including hydrogen problems) is 

also published jn Ror.emount Technical Data Sheet (TDS) 3045A00, "Corrosion and 

Its Effects" (current edition dated January 1995). 

3.3.2 Root Cause Conclusions and Recommendations 

At the time of this inspection, Rosemount was not planning any experiments or 

other research into the mechanisms of these transmitter failures with the 

exception of a technical evaluation of potential stress cracking at the Monel 

disc-to-stainless steel weld ring weld due to corrosion and/or differential 

expansion. At the time of preparation of this inspection report, Rosemount 

had not been able to obtain solubility data for hydrogen in the silicone oil 

used in Rosemount nuclear transmitters, nor has it obtained any new information on sources of hydrogen most likely to cause the type of failures 

experienced at St. Lucie. 

Rosemount maintained that its root cause analysis, as it had been developed at 

the time of the inspection, was consistent with that provided in the Rosemount 

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notification to its affected customers pursuant to 10 CFR 21.21(b), i.e. that 

monatomic hydrogen (whether from the coolant or generated by corrosion on the 

isolator surface) diffused through the Monel i sol ators, recombined to form 

molecular hydrogen, became trapped in the fill oil, and following a 

depressurization and repressurization, resulted in failed-high transmitter 

output signals. 

that its customers replace all potentially affected transmitters (i.e., those 

with known or suspected Monel isolator diaphragms), but it could not yet 

provide any concrete information that would enable 1 icensees to justify 

leaving certain transmitters in service longer than others in order to 

prioritize the replacements. Rosemount further stated that its current 

strategy was (through increased production and some stock diversion) to focus 

on supporting replacement of sensor modules with Monel diaphragms thus far 

identified as needed by customers. 

Rosemount believed this to be sufficient basis to recommend 

3.3.3 Rosemount Methodology for Scoping and Bounding the Problem 

To aid in understanding the search methodology employed by Rosemount to bound 

the problem (i.e., identify all the potentially affected sensor modules and 

transmitters), the basic construction of a Rosemount pressure or differential 

pressure transmitter and the manufacturing process (and its documentation) for 

a sensor cell is described below. 

3.3.3.1 Basic Transmitter Construction 

A complete transmitter consists of a sensor module, an electronics module, and 

two process flanges. A sensor module consists of a cylindrical stainless 

steel housing containing a sensor cell and printed circuit boards with 

discrete electronic components. 

housing such that the housing totally encloses the cell and the electronics, 

except that the isolator diaphragms (which for the units in question were made 

of Monel) and the adjacent inner surfaces of the weld rings that surround the 

isolator diaphragms remain exposed at either end face of the housing cylinder. 

The sensor module is sandwiched between the two transmitter process flanges 

that are bolted together using metal O-rings to seal the process chambers in 

flanges to the weld rings around the isolator diaphragms in the faces of the 

sensor module. The electrical leads from the sensor cell are connected to the 

circuit boards, and wire leads from the module originate at the circuit boards 

and extend from the threaded neck of the module housing. The leads are then 

connected to a terminal block in one chamber of the transmitter electronics 

module housing after the housing is screwed onto the threaded neck of the 

sensor module. 

The sensor cell is welded into the module 

3.3.3.2 Search Methodo’logy and Manufacturing Process Documentation 

The inspectors wal ked through the search methodology Rosemount employed to 

bound the problem thus far. 

serial numbers from St Lucie, 408929A and 411711A. The A suffix is used for 

transmitters with a serial number under 500000 that have been repaired (sensor 

module replaced) for the oil loss problem. To begin the search, this number 

(without the suffix) was entered into the repair records database which 

yielded among other information, repair house order (HO) number 767765. 

The incoming information was the two transmitter 

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110 
Microfilm records were searched for this repair HO number and the associated 

documents filed under the HO number. These documents include all the 

production or manufacturing travellers and traceability information. 

A manufacturing traveller is a document that accompanies the parts being built 

and lists the various attached drawings (DWGs), bills of material (BOMs), and 

manufacturing instructions (MIS) required to produce the part number identified as the finished assembly level designated on the traveller. 

signoffs for step completion and blocks to record traceability data. 

Traceability data consists of the part numbers and lot numbers or, if 

applicable, serial numbers or heat numbers, of the raw stock, purchased parts, 

or Rosemount-built parts used in building the assembly designated on the 

travel 1 er. 

There are 

The production of a sensor cell follows two principal paths from raw materials 

to finished cell. In the primary path, bar stock of a special alloy used by 

Rosemount is fabricated through several process steps into the so-called cell 

cups which then go through the glassing process and other steps. 

cell halves are welded together with a center (sensing) diaphragm and are then 

ready to have the two isolator assemblies welded onto each side of the cell. 

Finished 

The secondary path, in which the isolator assemblies are fabricated, was of 

particular relevance to this inspection. Accordingly, the inspectors walked 

through this process in the factory at Chanhassen, Minnesota, and examined it 

in detail. In this path, rolls or coils of metal foil strip stock, of various 

materials and thicknesses (each thickness of each material with a unique part 

number) are received by the receiving clerks, who inspect them for damage and 

packaging compliance with the invoice or packing slip, and assign each box 

(containing a single roll) a sequential lot number (sequenced for that part 

number), recorded in a computer database. The foil strip stock then undergoes 

receiving inspection, which, for nuclear part numbers, is done in accordance 

with Nuclear Engineering Department drawings and procedures. After the foil 

strip stock successfully completes the various examinations and tests during 

receiving inspection, the receiving inspectors must affix a stock tag or label 

to each box. It is also usually the practice, although not :r- "ically 

required by procedures, to mark the reel flange inside the box as well. The 

boxes of inspected strip stock are then stored in a segregated area and loaded 

periodically onto the designated ready service slide-down racks (one for each 

part number) next to the first production station area. Individual reels of 

strip stock are taken from these racks (as called for by the bill of materials 

for the part number of the disc assembly being produced) and placed on the 

disc punching or "blanking" machine. The bill of materials is attached to the 

traveller along with the manufacturing instruction and the drawing for 

blanking or punching out the foil discs (called "disc assemblies") from the 

strip stock. 

For this path in the production of a nuclear (1152, 1153 or 1154) sensor cell, 

the first traveller is for the disc assembly. 

01153-0252-0042 disc assembly, used in various nuclear transmitter models, 

including those affected by the problem in question, governs fabricating 

(blanking) disc assemblies from 2-inch-wide foil strip stock made of type 316L 

stainless steel; in this case, for disc assemblies using 0.004-inch thick 

The traveller for a Part No. 

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diaphragms among Range Codes 6 through 10. 

isolator assembly for which the principal production steps consist of cleaning 

and polishing the foil disc, laser welding it to a weld ring (a purchased 

part), which is supposed to be of the same material, an inspection, and a 

helium leak check of the assembly. Upon satisfactory completion of these 

operations, the part is now an isolator assembly, which for the 1153 

transmitter sensor modules in question is designated Part No. 01153-0262-1042. 

to the outside of each of the two cell cup halves of the sensor cell. 

3.3.3.3 Detailed Record Search Specifics 

The next assembly level is the 

I The final sensor cell assembly level is where an isolator assembly is welded 

During the inspectors' walkthrough of the Rosemount record search, the 

inspectors noted that the serial numbers of the failed 1153 transmitters, 

408929A and 411711A, led to records that indicated that these two transmitters 

had been refitted with Part No. 01153-0221-0192 sensor modules, Serial No. 

2328086 and 2328094 respectively. 

and 2328094 indicated that the modules had been assigned these serial numbers, 

as is standard Rosemount procedure, because they were built with sensor cells 

of the same serial numbers. During manufacture at Rosemount's Chanhassen, 

Minnesota, factory, sensor cells are identified by part number, by lot number, 

and by the heat number of the bar stock of the proprietary alloy from which 

the cell cup halves are made. Once a cell has been completed, it is assigned 

a complete cell part number (for each range code) and a unique serial number. 

When nuclear part number sensor cells are received at Rosemount's Eden 

Prairie, Minnesota, facility, each cell is filled with fill oil, sealed, and 

initially tested. It is then used to build a sensor module, which is assigned 

the same serial number as the cell it contains. 

The travellers for sensor modules 2328086 

The travellers for sensor cells 2328086 and 2328094 were retrieved from 

microfilm, printed, and reviewed. However, to initially determine the lot 

numbers of isolator assemblies used on those cells, Rosemount (and the 

inspector) reviewed another quality record called the Weld Log which also 

listed other sensor cells using the same lots of isolator assemblies. The 

Weld Log showed that these sensor serial numbers were among those of s?nsor 

assembly Lot 75, which consisted of Serial No. 2328077 consecutively through 

2328104. The Weld log showed that this entire lot (Range Code 9) had been 

built from Lot 55 of Part No. 01153-0262-1042 isolator assemblies (weld ring 

and foil disc or diaphragm). Next, the traveller for isolator assembly Lot 55 

was retrieved, printed, and reviewed. It showed that the entire Lot 55 of 

these isolators assemblies had been made from Lot 20 of Part No. 01153-0252- 

0042 foil disc assemblies. 

Finally, the traveller for Lot 20 of these foil disc assemblies, dated May 30, 

1989, was reviewed. The traceability block data showed that strip stock, Part 

No. C10181-0014, Lot (coil) No. 016 had been recorded (and presumably used). 

This part number was for the correct size (Dash No. -0014 indicates 0.004-inch 

thickness by 2.00-inch width), but the incorrect material (C10181 indicates 

commercial Monel Alloy 400). However, this original part and lot number had 

been lined out and Part No. CO9851-011 (indicating 316L, 0.004" X 2.00"), Lot 

23 and 24, written in, initialed by "V.K." and dated October 30, 1989. The 

annotation in the margin cited DR 491585, also dated October 30, 1989, and was 

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112 
initialed by the Nuclear Document Coordinator, as authority for the part 

number and lot number correction. Rosemount. stated that their initial 

investigation had led them to this point when qncestiawd as to the 

manufacturing histwy of these sensors by FP&L. 

3.3.3.4 Confirmation of Incorrect Material 

Review of the DR, which indicated that the Material Review Board had 

apparently concluded that a paperwork error had been made and corrected, 

reportedly did not suggest to Rosemount at the time that there was a definite 

material problem. However, as the chronology of events (Appendix C) confirms, 

after the Rosemount tecbician found the distended isolators at St. Lucie and 

after SwRI identified the entrapped gas as pure hydrogen, Rosemount ordered 

the material analysis of the isolator diaphragms and weld rings because, as 

explained by Rosemount, these facts strongly suggested that the isolators were 

actually made of Monel, the part and lot number of which, C10181-0014 and 

Lot 16, the blanking machine operator had recorded. 

that the identification of the hydrogen also prompted it to continue its 

search of records to determine the scope of the problem and the use of Monel 

was shortly thereafter confirmed by the SwRI: analyses. 

Rosemount further stated 

3.3.3.5 Identification of Affected Disc Assembly Lots 

Once the material of the Lot 20 disc assemblies was confirmed, Rosemount 

reviewed the travellers for five disc assembly lot numbers above and below Lot 

20 to determine which, if any, also may have been made with Monel strip stock, 

particularly from Lot 16. The inspector repeated this review and found, as 

had Rosemount, no other disc assemblies for nuclear transmitters in this 

series that did not have the correct part number recorded. None of the disc 

assembly lots in this series (for the disc assembly part number in question) 

used 316L strip stock Lots 22, 23, 24, or 25. Therefore, it was not clear on 

what basis the person with the initidls V.K. had selected Lots 23 and 24 of 

the CO9851-0011 316L strip stock that had been annotated on the Lot 20 disc 

assembly traveler to correct the discrepancy noted in DR 491585. Rosemount’s 

only plausible oxplaqation was that the person who selected 316L strip stock 

Lots 23 and 24 most likely chose 316L lots in use at about the same time 

frame. The inspectors reviewed the receipt inspection records and found that 

Lots 23 and 24 of the 316L strip stock had been received on April 18, 1989, 

and June 20, 1989, respectively, and both coils had been purchased under 

Rosemount purchase order (PO) No. EF3082. The also inspectors reviewed the 

certified material test reports for these lots of 316L strip stock with no 

discrepancies noted. 

However, because Lots 22 through 25 of the :316L (CO9851-0011) strip stock were 

evidently not used for Lot 20 disc assemblies, the inspectors’ asked Rosemount 

to find out where its records showed these lots were used. 

inspection, Rosemount reported that it had found that Lots 22 through 25 had 

been used for other disc assembly part numbers in process at the time. 

example, Lots 23 and 24 were used for disc assemblies intended for Model 1151, 

non-nuclear transmitters (See Appendix C to this report). 

Subsequent to the 

For 

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The receipt inspection record for Lot 16 of the Monel strip stock indicated it 

had been received on December 8, 1988, on PO ED1909 and had initially been 

rejected for slightly out of specification hardness on Rejection Document No. 

46902. This discrepancy was dispositioned "use as is" because the hardness 

was only slightly out of specification, all other physical properties and 

chemical composition were within tolerances and hardness itself was not 

considered a critical attribute for the application, but was used as a 

consistency overcheck for chemical composition and other physical properties. 

The inspectors concluded, particularly on the basis of the timing involved, 

that it was not likely that this discrepancy contributed to the inadvertent 

use of this Monel strip stock instead of 316L. 

3.3.3.6 Identification of Affected Isolator Assembly Lots and Sensor Modules 

The remainder of the Rosemount scoping review (repeated by the inspectors) 

consisting primarily of searching through the Weld Log, identified all the 

lots (serial number ranges) of sensor cells that were made from Lot 55 of the 

isolator assemblies which had been made with Lot 20 disc assemblies. Then the 

microfilm records of the isolator assemblies were searched to identify all of 

the other lots, if any, of isolator assemblies that may have been made, wholly 

or in part, with Lot 20 disc assemblies. This search, also repeated by the 

inspectors, revealed that one other isolator assembly lot, Lot 54, was built 

using Lot 20 Monel disc assemblies. Another review of the Weld Log identified 

all the groups of sensors (serial number ranges) that were built with Lot 54 

isolator assemblies. While determining which sensors had been made from Lot 

54 isolator assemblies, Rosemount noted that two sensor groups within Lot 65, 

Range Code 7, had been made from both Lot 54 and Lot 55 isolator assemblies. 

However, during this review, Rosemount discovered that some of the sensors of 

sensor Lot 76, Serial Nos. 2336503 through 2336530, were made from Lot 55 

isolator assemblies with Monel discs (diaphragms) and some from Lot 56 

isolator assemblies, which records (traveller traceability data) showed were 

made with 316L stainless steel discs as required. Traceability data for this 

so-called mixed lot of sensors indicated that Lot 56 isolators were made from 

Lot 21 disc assemblies, which were made from Lot 25 of 316L foil strip stock). 

In addition, at the lower end (by serial number) of the range of affected 

sensors (those potentially made with Lot 54 and 55 isolators), the Weld Log 

showed that in Sensor Lot 73, Serial Nos. 2290445 through 2290472, some 

sensors were made from Lot 54 Monel isolator assemblies and some from Lot 53 

isolator assemblies, which, like lot 56, records confirmed to be made of 316L 

as they were supposed to be. Therefore, among the approximately 450 sensors 

potentially affected, subsequent review of the individual travellers in the 

mixed lots indicated that about 50 of them actually had 316L isolators and not 

Monel. However, Rosemount reported all the 450 as suspect because its records 

did not indicate which serial numbers within these mixed groups used Lot 54 or 

55 (Monel) and which used Lot 53 or 56 (316L). 

Material analysis by X-ray fluorescence (XRF) was performed by Rosemount on 

three returned suspect (Part 21-listed) sensors from Northern States Power. 

The XRF identified two sensor modules as having stainless isolators and one as 

having Monel isolators. In fact, it was this analysis that led Rosemount to 

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114 
discover the second mixed lot earlier than their methodical search would have. 

Finally, Rosemount indicated in its compiled list of affected serial numbers 

of sensor modules those that were scrapped before shipment and others that 

were not logged as having completed pressurization aging and so also would not 

have been shipped. Appendix A to this report was taken from Rosemount’s list 

showing the names of potentially affected licensees and other purchasers and 

the quantities supplied. 

Subsequent to the inspection, in May 1995,, the Rosemount QA Manager informed 

the inspector that Rosemount had completec! a review of all travellers for Part 

No. 01153-0252-0042 disc assemblies from 1981 to December 1994 (which 

represents 49 documents), and in a different database, from January 1995 to 

the present, during which time 10 travellers were on file. The Rosemount QA 

Manager explained that the apparent difference in the rate of traveller use 

was because during the previous period, lot sizes were generally more than 

1000; whereas, recently, lot sizes have been typically fewer than 100. The 

corresponding travellers of the other nuclear transmitter foil disc assembly 

part numbers were still being reviewed. Rosemount reported that no 

discrepancies were identified during this review and that no other DRs had 

been identified for 01153-0252-0042 disc assembly travellers. 

There are five other configurations with unique part numbers used in nuclear 

transmitters of other range codes or for special applications. 

QA Manager later reported a review of travellers of the other disc assembly 

part numbers used in nuclear transmitters from 1981 to the present (a total of 

123 documents). Two DRs had been written. One in 1987 identified the wrong 

dash number recorded in the traceability block on a traveller, a part number 

suffix that indicated that the wrong thickness of strip stock (although 

correct material) may have been used. 

appropriately, to check the thickness of completed diaphragms from affected 

lots, which were found to be correct, indicating that the error had been in 

recording the wrong dash number used. The other DR, written in 1988, 

indicated that the wrong strip stock material had been used for a lot of disc 

assemblies, and all affected parts were scrapped. No other discrepancies were 

reported by Rosemount. 

The Rosemount 

The disposition of this DR was, 

3.4 Investiqation of Oriqinal Error 

To learn more about the original error, the inspectors examined in detail the 

processes of receiving, inspecting, and issuing strip stock, and the 

manufacturing processes for isolator assemblies at the factory in Chanhassen, 

Minnesota. The NRC had initially believed that the original error was welder 

selection of the incorrect disc assemblies to make isolator assemblies. 

However, the inspectors determined that the original detectable error actually 

occurred when a punch press or blanking machine operator used the incorrect 

material spool (Monel instead of 316L stainless steel) with which to make Lot 

20 of Part No. 01153-0252-0042 disc assemblies. According to Rosemount 

procedures, machine operators are supposed to verify that the material called 

out on bills of materials attached to travellers is used, and the 

traceability data from the Rosemount stock tag or stock label (part number and 

lot number), and only from the stock tag or label, is to be recorded on the 

traveller in the block provided. 

14 

115 
In this case, the blanking machine operator correctly recorded the information 

on the material used, but failed to verify that the material used was correct 

according to the bill of materials. In doing so, the non-nuclear part number, 

C10181-0014, for strip stock of Monel Alloy 400, lost its material identity 

and became, for all anyone further down the manufacturing line would know, 

part of a nuclear transmitter sensor cell subassembly. The welder later 

selected Lot 20 disc assemblies identified with the correct part number for 

the Lot 54 and 55 isolator assemblies being made, but did not know that the 

Lot 20 disc assemblies were made of the wrong material. The inspectors also 

learned however, that the welder may have compounded the error by failing to 

recognize certain anomalies in the welding process (discussed later). 

It could not be determined why or how the wrong (Monel) strip stock had been 

used initially to make the Lot 20 disc assemblies. However, the inspectors 

determined that due to the part numbering system employed since 1990, using 

"N" numbers for material to be used in nuclear transmitter parts (there are no 

Monel "N" numbers), the error would be much less likely to occur after that 

time. Also, according to Rosemount records, there were no returns or 

complaints other than from St. Lucie attributable to this problem. 

facts support the conclusion that Rosemount's search methodology was adequate 

and that the probability of Monel isolators being used inadvertently in 

nuclear transmitters since the time of this incident is very low. 

These 

In examining boxes of material on the rack next to the punch press, the 

inspector noted that the Rosemount computer-printed stock label was placed on 

the outside of the box, but not on the spool flange holding the coiled foil 

strip stock. As discussed above, it was the practice sometimes, although not 

required by procedure, to mark the spools as well with part number, lot number 

and sometimes heat number. The inspector asked for copies of the purchase 

orders for the material in question to determine what marking requirements 

were imposed on the vendor. The inspector reviewed Rosemount Purchase 

Specification PS-25, called out on the drawings for the strip stock (C101, 81 

for Monel 400 and C09851 for 316L). The inspectors found that PS-25 required 

that all packaging (taken by Rosemount to mean exterior) be marked with the 

Rosemount part number. 

After reviewing PS-25, the inspector noted that some packaging previously 

observed by the inspector may not have been marked in accordance with PS-25. 

One box of N09851-0011 strip stock had a printed white label taped onto the 

box with only the PO number hand written in. 

Rosemount stock tag (label) put on by the receipt inspectors, but the 

inspector did not recall noticing the part number elsewhere on the box. 

boxes examined (e.g., for Monel and commercial 316L (C09851)) had the 

Rosemount part number stencilled on the box by the vendor in addition to heat 

number, and purchase order number. 

boxes of strip stock may not conform to PS-25 was pointed out to the 

accompanying cognizant Rosemount staff for investigation and appropriate 

disposition. 

The part number was on the 

Other 

The possibility that some markings on 

At the time of the inspection, Rosemount was interviewing supervisors and 

reviewing records. Subsequent to the inspection, Rosemount reported that it 

had identified the blanking machine operator from the initials on the 

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116 
traveller in question. This operator was an employee who had about 11 years 

experience at the time, but is no longer with the company. 

not attempt to locate and interview this person during this inspection. 

Rosemount has not yet been able to determine the root cause of the mistake. 

The inspectors were told that the ready service supply racks next to the 

blanking machine were the same as they were in 1989. The inspectors noted 

that they were marked with standard locations for each part number, and the 

Monel and stainless part sections are not adjacent. The inspectors did not 

identify any apparent working condition or situation that would have been (or 

would now be) conducive to selecting incorrect material. The inspectors 

concluded however, that the use, since just after this incident, of N part 

numbers for "raw" materials to be used in assemblies that have nuclear (Le., 

1152, 1153 or 1154 prefix) part numbers would be the most effective mans of 

preventing use of non-nucl ear parts (such as Monel ) i n nucl ear transmitters . 

The inspectors did 

The inspectors concluded that Rosemount's failure on flay 30, 1989, to follow 

Blanking Procedure 01153-3036 constituted a nonconformance with respect to the 

requirements of Criterion V, "Instructions, Procedure, and Drawings," of 

Appendix B to Part 50 of Title 10 of the Code of Federal Resulations (10 CFR 

Part 50). Criterion V requires that activities affecting quality be 

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

appropriate to the circumstances and that they be accoaspl ished in accordance 

with these instructions, procedures, and or drawings. Instructions, 

procedures, or drawings are required to include appropriate quantitative or 

qualitative acceptance criteria for determining that important activities have 

been satisfactorily accomplished. Rosemount Manufacturing Instruction 01153- 

3036, "Blanking Procedure," instructed the operator: "Obtain the proper stock 

coil material for the part number to be blanked." The bill of material (w attached to the traveller for Part Number 01153-0252-0042 foil disc 

assemblies, Lot 20, specified using Part Number C89851-0011, 3161. stainless 

steel foil strip stock. 

However, while blanking Lot 20 of Part Number 01153-0252-0042 foil disc 

asselabl ies, the operator apparently used Part Mer C10181-0014 Mae 

400, Lot 16, strip stock for this process instead @f the required Par 

CO9851-0011 316L; although the sperator did dscment use of lot 16 of 

Mom1 in the traceability data block on the traveller. 

then redesignated as 01153-0252-0042 disc assemblies, the identity of the 

material was lost and was henceforth, as would be expected, presumed t~ be the 

correct material. This discrepancy was not detected during subsequen 

operations, despite the apparent considerable difficulty in successfu 

welding the Monel discs to 316L weld rings. The difficulty was indic 

part on the basis of discussions with factory personnel (instrument 

who explained that (1) the laser welding machine would have had to b 

cgwtly adjusted from its nominal settings for stainless steel in ord 

welds of the 316L weld rings to Monel disc assemblies to pass the 1 

(2) laser welding Monel produces a characteristic green glen, and (3) 

actual recorded yield from Lots 54 and 55 totaled only $80 isolator 

; whereas, the potential yield available was 2500 as derived f 

r of disc assemblies (2500) in Lot 20. The discrepancy was ca 

ar Department Document Coordinator review cJf the Bcncusrent pack 

csntaining the travellers. Because this error was caught arrd propa 

#en the parts were 

16 

117 
documented by the Rosemount QA system's routine review and screening process, 

this error is not being cited as a nonconformance. 

3.5 Review of Discrepancy ReDort DisDosition 

The inspectors reviewed Discrepancy Report (DR) No. 491585 written by the 

Nuclear Document Coordinator on October 30, 1989, when she discovered the 

error during final record review. The coordinator, who still held this 

position at the time of the inspection, stated that she turned the report over 

to the Material Review Board (consisting of the production supervisor, the 

cognizant design engineer, and the cognizant quality engineer) for 

disposition. Subsequent to the inspection, the Rosemount QA manager informed 

the inspector that two of the three members of the Material Review Board that 

dispositioned DR 491585 (the quality engineer and the design engineer) had 

been interviewed and had submitted written statements regarding their 

recollections of the rationale for their disposition of DR 491585. 

inspectors noted that current Rosemount procedures required written 

justification for DR dispositions, but this had not been required at the time 

the DR in question had been inappropriately dispositioned. 

The 

Rosemount reported that the quality engineer first explained that he had 

believed that use of Monel was highly unlikely because to the best of his 

recollection, all the correct piece parts were (and are) kept in a locked 

cabinet near the isolator assemb'ly welding station. The inspector and 

Rosemount QA Manager noted that this indicated a lack of understanding on his 

part of how the mistake occurred. 

belief at the time that Monel discs could not have been successfully welded to 

stainless steel weld rings. 

leak checked with helium and it was believed that isolator assemblies with 

stainless steel weld rings and Monel discs would not have passed the test. 

The quality engineer finally added the rationale that no punch marks were 

visible on the weld rings, indicating that they were 316L. 

logic, if Monel had been used for the discs, the weld rings would have also 

been Monel and would show the characteristic identifying punch marks. The 

inspector and the Rosemount QA Manager both noted that this was circular 

reasoning because the weld rings would only be expected to be Monel and show 

Monel punch marks if Monel was being used intentionally for the discs. This 

reasoning, again, showed lack of full understanding of the process and the 

problem by the quality engineer. The QA Manager stated that he would examine 

this issue further, including talking with the quality engineer about the 

discrepancies in his reasoning. In addition, the QA Manager would attempt to 

find out if the instrument builders involved had been interviewed to determine 

if they had noted unusual welding settings or adjustments being required to 

successfully weld the Monel discs to the stainless weld rings, if anyone had 

noticed the characteristic green glow of Monel being laser-welded, or if 

anyone had questioned the apparently unusually high difference between the 

number of disc assemblies and the yield of isolator assemblies indicating a 

high scrap rate from inspection and tests of finished welds. 

The quality engineer further stated his 

He also stated that the welds are examined and 

According to this 

Rosemount reported that the statement of the design engineer Material Review 

Board member indicated that'she had concurred in the disposition of simply 

"correcting the part numbers on the traveller" because she stated that the 

17 

118 
discrepancy was characterized on the DR as "incorrect part number recorded" 

and the QA Manager added that he had been told that there were "a large number 

of paperwork errors at the time." The inspector noted that the design 

engineer's recollection of the language of the DR was inaccurate, suggesting a 

misinterpretation. In fact, the DR stated "wrong part number.. .used to build 

product." 

basing of her concurrence on (1) simply noting that the production supervisor 

and quality engineer had concurred and (2) an inaccurate understanding of the 

implication of the DR and nature of the mistake. 

The QA Manager agreed to discuss with the design engineer the 

Rosemount later reported (June 1995) that the third member of the Material 

Review Board, the production supervisor, had been interviewed, but could not 

recall his rationale for the disposition of the DR; although he did not 

disagree with the statements of the other two board members. 

the reported statements by the Material Review Board, the inspectors concluded 

that the inappropriate disposition of the DK resulted from incomplete 

understanding of the problem and a perfunctory review. 

factory personnel gave the inspectors the sense that a climate of low 

tolerance for identifying problems that may have existed at the time of this 

incident (which it was emphasized no longer existed) and perhaps some 

production pressure, may have contributed to the lack of recognition of 

unusually high scrap rates during welding a:; well as quick acceptance by the 

Material Review Board of an easy explanation for the wrong part number issue 

identified in the DR. 

On the basis of 

Discussions with 

Criterion XV, "Nonconforming Material, Part!; or Components," of 10 CFR Part 

50, Appendix B, states: "Measures shall be established to control materials, 

parts, or components which do not conform to requirements in order to prevent 

their inadvertent use or installation. These measures shall include, as 

appropriate, procedures for identification, documentation, disposition, and 

notification to affected organizations." 

Criterion XVI, "Corrective Action," of 10 CFR Part 50, Appendix B, states, in 

part: 

quality, such as failures, malfunctions, deficiencies, deviatinnr defective 

materi a1 and equipment, and nonconformances are promptly identi tied and 

corrected. 

measures shall assure that the cause of the condition is determined and 

corrective action taken to preclude repetition.. ..'I 

"Measures shall be established to assure that conditions adverse to 

In the case of significant conditions adverse to quality, the 

Contrary to the above, measures established by Rosemount to control certain 

nonconforming material did not prevent its inadvertent use. In addition, 

action taken to correct the related documented condition adverse to quality 

was inadequate in that Discrepancy Report 491585, written October 30, 1989, 

was dispositioned inappropriately by the Material Review Board. The 

discrepancy report documented the use of strip stock material that was not in 

accordance with the bi17 of materials to make Lot 20 of Part No. 01153-0252- 

0042 disc assemblies. 

what material was actually used in making Lot 20 of these disc assemblies, 

then taking the steps necessary to capture any incorrect material and prevent 

its use in transmitters designed for nuclear safety-related service, the 

Material Review Board dispositioned the discrepancy by directing that the 

Instead of investigating the condition and verifying 

18 

119 
strip stock part number on the traveller be “corrected.“ As a result, Lot 16 

Uf C10181-0014 hnel , documented as having been used in the data block on the 

traveller for Lot 20 of the disc assemblies, was actually so used, yet the 

aceability data on the traveller was erroneously changed to read Lots 23 and 

ainfess steel) (95-02-01). 

of the-part nuntber called out on the bill af materials, CO9851-0011 (316L 

3.6 Review of Rosemount #aterial ldenti fieation Tests 

inspectors reviewed traces of X-ray fluorescence (XRF) nondestructive 

ination (#ID€) of isolators returned from Northern States Power (NSP). 

printout from XRF equipment, XRF traces of material known to be Monel or 

stainless steel are very distinctive, both qualitatively (pattern) and 

titatively (amplitudes, distribution) Traces for XRF tests on isolators 

eturned modules wre crpnsisemt with thes w19 ts be either Monel or 

nless respectively. Hewever, instead of f the three returned NSP 

In 

ntificatiotl of the second mixed lot described in Paragraph 3.3.3.6 above, 

s one made from both Lot 54 (hnel) and Lot 53 (316L). Further review of 

travellers of these mdules confirmed the information in the Weld Log. 

iBspectws concluded that material analysis performed thus far has 

firmed the re1 i abi 1 i ty of Rosemount records. 

emocont Part 21 report, indicated that the serial number of an affected 

lacement traesmitter provided by Rosemunt had serial number 413060A; 

eas, the Part 21 report listed this transmitter as having serial nimber . Rosmovnt had detemined, wid the inspectors confirmed, that this 

error in transferring the serial number from Rosemount production 

s, which agree with Dresden procurmnt/installation records, to the 

publishd with the Part 21 customer nzttification. At the time of 

ring this report, there have been ~18 &her similar occurrences reported. 

t the St. Lucie plant that is 

e ~4f transmitters, specifically their 

tihe capability of identifyi anmi ttms with Monel i sol ators. 

%Bitters that were not on Rosemount’s 

St. lueie believed, OR the s of data from this AMS equipment, 

a1 for expanding the scope of 

shed by Rosemount. During this 

19 

120 
inspection, the inspectors asked for the results of material analyses that 

Florida Power & Light was performing on certain transmitter isolators to 

confirm their theory. 

material of the isolators in one transmitter not listed by Rosemount that had 

exhibited a so-called "Monel-like" AMS signature was evaluated as stainless 

steel using what the licensee described as a gamma-backscatter test. This 

result renders the AMS method for detecting Monel inconclusive at present. 

However, based on the! latest report from St. Lucie, the 

3.9 Customer ComDlaints/Returns Involvlncr Other Than Oil-Loss SymDtoms 

Rosemount reported reviewing customer complaints/returns documentation and 

stated that prior to the St. Lucie event, it; has had no returns or complaints 

regarding nuclear (1152,1153, or 1154) transmitters attributable to the 

Monel/hydrogen intrusion problem. The review was described as a computeraided search of the listed final disposition failure modes field of the 

customer return/complaint database, using pertinent keywords such as gas, 

intrusion, isolator, Monel, and hydrogen. The inspectors did not make an 

independent check of this process during this inspection. 

4.0 PERSONS CONTACTED 

Rosemount Nuclear Instruments, Incorporated 

Mark Van Sloun, VP & General Manager 

Ken Ewald, Business Unit Manager 

Jerry Valley, Quality Assurance Manager 

Stuart C. Brown, Engineering Supervisor 

Timothy J. Layer, Product Marketing Manager 

Paul Roepke, Receiving Department Technician 

Lori Majerus, Receiving Department Inspector 

Esther Pollard, Receiving Inspection Supervisor 

Jeff Bracken, Nuclear Inspector 

Bonnie Strawberry, Production Line Supervisor 

Jan Bockman, Inqtrumcnt Bui 1 der 

20 
APPENDIX A 

ORGANIZATIONS IN THE U.S. TO WHOM ROSEMOUNT REPORTED SENDING 

TRANSMITTERS OR SENSOR MODULES WITH MONEL ISOLATORS 

Organization 

Arizona Public Service 

Baltimore Gas & Electric 

Bec h tel 

Boston Edi son 

Carolina Power & Light 

Commonwealth Edi son 

Consumers Power 

Duke Power 

Duquesne Light Company 

Ellis & Watts 

Florida Power Corp. 

Florida Power & Light 

Georgia Power 

GPU 

Gulf States Utilities 

Houston Lighting & Power 

Illinois Power 

Maine Yankee Atomic Power Company 

New Hampshire Yankee, Inc. 

New York Power Authority 

Niagara Mohawk Power Corp. 

Northern States Power 

Omaha Public Power District 

Pacific Gas & Electric 

Pennsylvania Power & Light 

Phi 1 adel phi a El ectri c Company 

Portland GE 

Public Service Electric & Gas 

South Carolina Electric & Gas 

Southern Calf Edison 

Systems Energy 

To1 edo Edi son 

TU Electric 

TVA 

Vermont Yankee 

Virginia Power 

Washington Public Power 

Westinghouse 

Wolf Creek NOC 

Yankee Atomic 

Supply System 

Faci 1 i ty 

Palo Verde 

Calvert Cliffs 

Quanti ty 

Pilgrim 

Brunswick, Harris, Robinson 

Byron, Braidwood, Quad Cities 

LaSal 1 e, Dresden, Zion 

Palisades, Big Rock Point 

Oconee, Catawba, McGuire 

Beaver Val 1 ey 

Crystal River 

St. Lucie, Turkey Point 

Hatch, Vogtle 

Oyster Creek, Three Mile Island 

River Bend 

South Texas Project 

C1 i nton 

Maine Yankee 

Seabroo k 

Fitzpatrick, Indian Point 3 

Nine Mile Point 

Monticello, Prairie Island 

Ft. Calhoun 

Di ab1 o Canyon 

Susquehanna 

Limerick, Peach Bottom 

Trojan 

Salem, Hope Creek 

Summer 

San Onofre 

Grand Gulf, Waterford 

Davis-Besse 

Watts Bar, Sequoyah, Browns Ferry 

Vermont Yankee 

Surry, North Anna 

Comanche Peak 

WNP-2 

Wolf Creek 

122 

9 

2 

1 

13 

4 

2 

16 

4 

1 

1 

16 

33 

2 

2 

5 

3 

1 

2 

8 

5 

6 

12 

4 

2 

1 

8 

5 

7 

4 

2 

7 

2 

2 

5 

1 

1 

3 

5 

3 

2 
I APPENDIX B 

CHRONOLOGY 

ROSEMOUNT DIAPHRAGM ISSUE 

11/22/94 2137 Failure of two Rosemount pressurizer pressure 

transmitters generate an SI signal during 

repressurization from cold shutdown at St. Lucie 1 

11/23/94 0125 ESF actuation reported by FP&L under 10 CFR 50.72; no 

I mention of root cause 

12/15/94 

12/94-02/95 

02 /03/95 

I 02/22/95 

Rosemount tech at St.. Lucie finds no oil loss, but 

identifies gas in fill oil as probable failure mode 

due to distention/depressibility of diaphragms. 

Rosemount continues internal investigation and 

searches for laboratory capable and willing to handle 

tests on radioactively contaminated sensor modules. 

Failed sensor modules sent from St. Lucie to SwRI for 

examination and testing. 

Hydrogen identified as only entrapped gas by SwRI. 

SwRI finds no evidenc:e of corrosion, water leakage, or 

fill oil breakdown. 

I i 03/14/95 SwRI identifies diaphragm material as Monel. 

03 / 17 /95 

1 03/20/95 

03/21/95 

Number of transmitters potentially involved identified 

by Rosemount. 

Rosemount transmits to NRC preliminary list of 

affected organizations and the serial numbers of 

potentially affected modules provide? 1 '+em along 

with preliminary technical evaluation and root cause 

analysis. 

Rosemount issues 10 CFR 21.21(b) notification to 

affected customers. 

03/22/95 NRC issues Information Notice 95-20 

03/23/95 NRC activates Regulatory Response Groups of owners 

groups (Westinghouse, BWR, B&W, CE) to obtain 

information on transmitter location, operability, 

safety assessment, corrective action, etc. 

03/31/95-04/03/95 .I ,. RRGs submit written resDonses to NRC 

04/05-07/95 NRC conducts inspection at Rosemount. 

123 
APPENDIX C 

TRACEABILITY DATA SUMMARY 

Foil StriD Stock or StriD Stock (FSS) 

Part Number C10181-0014: C=commercial , lOl8l=Monel alloy 400, -0014=coiled, 

spooled metal foil strip stock, 0.004" thick x 2.00" wide 

Lots of Interest: 16 (assigned by computer during receiving) 

Part Number CO9851-0011: commercial, 09851=316L, 0011=0.004" X 2.00" Lots 22 

through 25 not used for D/As of interest, D/A Lots 15-19, used 316L FSS Lots 

11, 12, 12, 13, and 21 respectively, D/A Lots 21 and 22: 316L FSS Lots 26 & 

27, D/A Lots 23,24 1st to'use nuclear N09851-0011, Lots 001 and 003, D/A Lot 

25: NFSS Lots 003 & 005 

Part Number N09851-0011: N=nuclear grade, 09851=316L, 0011=0.004" X 2.00" (Nnumbers used for part numbers involving 01152,3,4 since 1989 

Disc Assembly (D/AL 

Foil Disc "Assembly" (punched out disc) Part Number: 01153-0252-0042 

Lot 19 traceable to 316L (CO9851-0011) strip stock Lot 21 

Lot 20 traceable to Monel (C10181-0014) strip stock Lot 16 (erroneously 

determined to be traceable to 316L Lots 23 and 24) 

Lot 21 traceable to 316L (CO9851-0011) strip stock Lot 26 

Is01 ator Assembl y ( I /A1 

(consisting of a cleaned, polished disc "assembly" welded to a weld ring 

I/A Part Number 01153-0262-1042, I/A Lot 53 traceable to 

D/A Lot 19 -Identified as last I/A Lot to use Lot 19 D/As. (Checked back to 

Lot 49 to confirm Lot 19 or below and not Lot 20), I/A Lot 54 traceable to D/A 

Lot 20 traceable to FSS Lot 16-identified by search of records of I/A 

travellers to see which I/A Lots used Lot 20 D/As, I/A Lot 55 traceable to D/A 

Lot 20 traceable to FSS Lot 16-identified through Weld Log by module serial 

numbers 2328086 and 2328094 (same as sensor serial numbers) installed in FP&L 

(St Lucie) transmitters 408929A and 411711A respectively according to 

Rosemount records filed under repair HO # 767765. 

Lot 21 - first I/A Lot to use Lot 21 D/As. (Checked forward to Lot 62 to 

confirm use of D/A Lots 21 and above and no more 20s) 

Lots of interest: 

I/A Lot 56 traceable to D/A 

Sensor Cell s 

Sensor cells and modules have same serial numbers. Sensor cells for range 

codes 6-10 of interest IAW Weld Log only these RCs used I/A Lots 54 or 55 

Part Numbers, e.g., 01153-0264-0092 for range code (RC)-9. Lots of Interest: 

Lots 73, 63, 66, 74, 14, 42, 43, 65, 75, 68, 76 made entirely or in part with 

Lot 54 or 55 I/As. Lot nos. contain serial number ranges in Part 21 report. 

Transmitters 

Model number, e.g., 1153GD9PB for RC-9, guage pressure, records trace 

cells/modules to XMTRs in which used (some supplied as modules only). 

124 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. -1 

August 17, 1995 

Mr. Martin R. Benante 

President and General Manager 

Target-Rock Corporation 

1966E Broadhol 1 ow Road 

Farmingdale, NY 11735-0917 

SUBJECT: NRC INSPECTION NO. 99900060/95-OIL AND NOTICE OF NONCONFORMANCE 

Dear Mr. Benante: 

This letter addresses the inspection of your facility at Farmingdale, New 

York, conducted by Mr. Stephen Alexander and Mr. Paul Narbut of this office on 

July 11-12, 1995, and the discussion of their. findings with the members of 

your staff identified in the enclosed report at the conclusion of the 

inspection and in subsequent telephone conversations. The inspection was 

conducted to examine Target Rock's actions arid conclusions regarding three of 

its pilot-operated main steam safety relief valves (SRVs) which failed to 

operate when the valves were tested at Nebraska Public Power District's Cooper 

Nuclear Station. 

Areas examined during the inspection and our findings are discussed in the 

enclosed report. 

surrounding the SRV failure to provide a basis for assessing the validity and 

completeness of your root cause analysis and your investigation to determine 

the extent of the problem with corroded SRY pilot valve solenoids in the 

industry. The inspectors also reviewed your procedures adopted pursuant to 

Section 21.21 of Part 21 of Title 10 of the _Code of Federal Regulations and 

your actions regarding the solenoid valve failures in question prescribed by 

those procedures. Within these areas, the inspection consisted of an 

examination of procedures and representative records, interviews with 

personnel, and observations by the inspectors. 

The inspection included a review of the circumstances 

During this inspection, we determined that the implementation of your quality 

assurance (QA) program failed to meet certain NRC requirements. The failure 

to prevent inadvertent supply of nonconforming components (i.e., solenoid 

valves apparently containing residual hydrostatic test water) to an NRClicensed facility constituted a nonconformance with respect to the requirements of 10 CFR Part 50, Appendix B. The specific findings and references to 

the pertinent requirements are identified in the enclosures to this letter. 

You are requested to provide us within 30 days from the date of this letter a 

written statement in accordance with the instructions specified in the 

encl osed Notice of Nonconformance. 

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

this letter, its enclosures, and your response will be placed in the NRC 

Public Document Room (PDR). 

include any personal privacy, proprietary, or safeguards information so that 

To the extent possible, your response should not 

125 
H. Benante -2- 

it can be placed in the PDR without redaction. 

necessary to include such information, you should clearly indicate the 

specific information that you desire not to be placed in the PDR, and provide 

the legal basis to support your request for withholding the information from 

the public. 

However, if you find it 

The responses requested by this letter and the enclosed Notice are not subject 

to the clearance procedures of the Office of Management and Budget as required 

by the Paperwork Reduction Act of 1980, Public Law No. 96-511. 

The cooperation of your staff in this matter was greatly appreciated. Should 

you have any questions about the enclosed report, we would be glad to discuss 

them with you. 

Sincerely, 

I- Robert M. Gallo, Chief 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Docket No. : 99900060 

Enclosures: 1. Notice of Nonconformance 

2. Inspection Report 99900060/95-01 

126 
NOTICE OF NONCONFORMANCE 

Target Rock Corporation 

Farmingdale, New York 

Docket No. 99900060 

Report No. 95-01 

Based on the results of an NRC inspection conducted on July 11-12, 1995, it 

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

NRC requ i remen t s . 

A. Criterion XV, "Nonconforming Material, Parts or Components," of 10 CFR 

Part 50, Appendix €3, states: "Measures shall be established to control 

materials, parts, or components which do not conform to requirements in 

order to prevent their inadvertent use or installation. These measures 

shall include, as appropriate, procedures for identification, 

documentation, disposition, and notification to affected organizations. 

Nonconforming items shall be reviewed and accepted, rejected, repaired, 

or reworked in accordance with documented procedures. I' 

Criterion V, "Instructions, Procedures and Drawings," of 10 CFR Part 50, 

Appendix 6, states, in part: "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 instructions, procedures, or drawings. 

Instructions, procedures, or drawings shall include appropriate 

quantitative and qualitative acceptanc:e criteria for determining that 

important activities have been satisfactorily accomplished. 

Contrary to the above, Target Rock failed to properly control certain 

activities to prevent the inadvertent supply of nonconforming material 

to an NRC-licensed facility in that Procedures or procedural compliance 

were inadequate to ensure that three Model 1/2-SMS-S-02 solenoid pilot 

valves for safety relief valves supplied to the Cooper Nuclear Station 

were properly and completely dried following a hydrostatic test. As a 

result, the solenoid valves failed to operate due to corrosion in the 

core tube caused by storage for about 1-1/2 years prior to installation 

with residual hydrostatic test water trapped in the core tubes. 

(95-01-01) 

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

Regulatory Commission, ATTN: Document. Control Desk, Washington, DC 

20555 with a copy to the Chief, Special Inspection Branch, Division of 

Inspection and Support Programs, 0ffic:e of Nuclear Reactor Regulation, 

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

Nonconformance. 

Notice of Nonconformance" and should include for each nonconformance 

This reply should be clearly marked as a "Reply to a 

Enclosure 1 

127 
(1) the reason for the nonconformance, or if contested, the basis for 

disputing the nonconformance, (2) the corrective steps that have been 

taken and the results achieved, (3) the corrective steps that will be 

taken to avoid further noncompliances, and (4) the date when your 

corrective action will be completed. Where good cause is shown, 

consideration will be given to extending the response time. 

Dated at Rockville, Maryland 

this 17th day of August, 1995 

2 

128 
REPORT NO.: 

ORGANIZATION: 

ORGANIZATIONAL 

CONTACT : 

TELEPHONE : 

NUCLEAR INDUSTRY 

ACTIVITY: 

INSPECTION DATES: 

LEAD INSPECTOR: 

OTHER INSPECTORS: 

REVIEWED BY: 

APPROVED BY: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF INSPECTION AND SUPPORT PROGRAMS 

99900060/95-01 

Target Rock Corporation 

1966E Broadhol 1 ow Road 

Farmi ngdal e, NY 11 735-091 7 

James D. White 

Sales and Service Manager 

(516) 293-3800, EXT 647 

Target Rock's activities within the scope of this 

inspection include manufacturing and supplying valves, 

primarily relief valves, and replacement parts and 

service to the nuclear industry. 

July 11-12, 1995 

&- 

Vendor Inspection Section (VIS) 

Special Inspection Branch (PSIB) 

Paul P. Narbut 

Special ,Inspection Section, PSIB 

Enclosure 2 

129 
1.0 SUUMRY OF INSPECTION FINDINGS 

The inspection was conducted to examine Target Rock Corporation's (Target 

Rock's) actions and conclusions regarding three of its pilot-operated safety 

relief valves (SRVs) which failed to operate when the valves were tested at 

Nebraska Pub1 ic Power District's (NPPD's) Cooper Nuclear Station (Cooper). 

The inspection basis consisted of the following: 

Appendix 6, "Qual i ty Assurance Criteria for Nuclear Power P1 ants 

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

of Federal Resulations (10 CFR Part 50) 

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

1.1 Violations 

None 

1.2 Nonconformance (99900060/95-01-01} 

Contrary to the requirements of Criteria V and XV of 10 CFR Part 50, 

Appendix B, Target Rock failed to prevent the supply of nonconforming material 

(i .e., SOWS containing residual water) to an NRC-1 icensed facility (Cooper) 

due to inadequacy of and/or noncompliance with procedures governing 

manufacture and testing of the material (See Section 3.1.3 of this report). 

1.3 BPen Item (99900060/95-01-02} 

Review of Target Rock's actions regarding a problem it identified involving 

damage to SRV main disc return springs cause by repeated SRV functional 

testing using the standard reduced flow test setup. 

damage includes apparent spring relaxation spring end chipping. 

3.2 of this report). 

The cycle-dependent 

(See Section 

2.0 STATUS OF PREVIOUS INSPECTION FINDINGS 

No previous findings were reviewed during this inspection. 

3.0 INSPECTION FINDINGS AND OTHER COMENTS 

3.1 Failed Tarclet Rock SOVs at COOD er 

3.1.1 Background 

On February 10, 1995, NPPD declared a Notice of Unusual Event at Cooper when 

three Target Rock Model 7457F-600 main steam SRVs failed to open upon receipt 

of a manually initiated electric signal during startup testing. Cooper has 

eight SRVs, six associated with the automatic depressurization system (ADS) 

2 

130 
and two with the Low-Low setpoint safety relief logic. 

attributed the failure of the two non-ADS SRVs to open upon command to a 

failure of the SRVs’ solenoid-operated pilot air valves (SOVs) to port air or 

nitrogen to the SRVs’ pneumatic operating cylinders. 

examined one of the failed SOVs and found internal corrosion had caused 

binding of the solenoid actuating mechanism. (REF: PNO-IV-95-005) 

NPPD and Target Rock 

NPPD disassembled and 

3.1.2 NRC Preliminary Inquiries 

During separate NRC telephone conversations with Cooper and Target Rock on 

February 15, 1995, the NRC was told that the three SOVs, two of which later 

failed (Model 1/2-SMS-S-O2, Serial Numbers 376, 377, and 378), were del ivered 

to Cooper with three replacement SRV’s, Model 7567F-600, purchased from GE 

Nuclear Energy (GE NE) by NPPD for Cooper. The SRVs were obtained by GE NE 

from uninstalled spares at the cancelled Shoreham Power Plant (Shoreham) and 

were refurbished by Target Rock at its facility in Farmingdale, New York, 

under contract with GE NE. The SRVs had not been in service at Shoreham. 

Target Rock installed new, improved-design, Model 1/2-SMS-S-02, SOVs on the 

three SRVs at its facility, then shipped the completed SRVs to Cooper. 

Target Rock concluded that only the three SOVs supplied for Cooper were 

affected, because they were the only ones ever built entirely with new parts 

by field service personnel, based on personnel recollections of the field 

service manager. Additional information in support of this conclusion was 

reviewed by the inspectors during this inspection at Target Rock. 

Initially, Target Rock stated that it did not intend to issue a report to the 

NRC pursuant to 10 CFR 21.21(a) because it had determined that it did not have 

the capability to evaluate this deviation for existence of a defect. 

Rock stated its intention to identify potentially affected licensees or 

purchasers (although Target Rock believed the problem to be isolated to 

Cooper) and to inform them of the deviation pursuant to 10 CFR 21.21(b). 

Target 

The Target Rock Sales and Service Manager stated during another telephone 

conversation kith the NRC, on March 30, 1995, that he had traveled to Cooper 

and disassembled and inspected the second failed solenoid. He found rust in 

the bonnet tube which would inhibit the plunger movement and concluded that 

the assembly had been improperly or ineffectively dried (due to installed 

internals) at the time of manufacture. 

made in the manner described above, and on1.y those three required the second 

hydrostatic test, Mr. White concluded that the problem was limited to the 

three assemblies provided to Cooper. 

to reporting these conclusions. 

The initial NRC assessment, based on the above information, was that a generic 

concern did not exist and immediate notification of industry was not required. 

Since only three such assemblies were 

Target Rock has sent a letter to Cooper 

3.1.3 Inspection at Target Rock 

The inspectors examined the actions that Target Rock had taken since the 

February 15 ana March 30, 1995 telephone discussions. 

changed its conclusions regarding that the problem was limited to the three 

Target Rock had not 

3 

131 
valves at Cooper. 

second of the failed valves. 

disassembled valve internals. 

representative stated that the movable core was restricted from movement by 

rust and that the rust was in the area which would contain the post-hydro 

water. The core showed a rust pattern that was consistent with the storage 

position of the valve. 

valve to be necessary since the evidence from the first two valves was 

consistent. 

Target Rock had been to the Cooper site and examined a 

The inspectors examined photographs of the 

The rust was evident and the Target Rock 

Target Rock did not consider disassembly of the third 

As is occasimally Target Rock’s practice in the case of special orders, the 

work on these SRVs, including new SOV shell assembly, hydrostatic testing (for 

strength of pressure-retaining parts and fasteners) , final assembly, 

installation and final tightness and functional testing, was done at the 

Farmingdale facility by Target Rock field service personnel instead of 

production assembly and test personnel. Although field service personnel 

routinely service SRVs in the field, including disassembly, cleaning, software 

renewal and reassembly of the SOVs, they do not perform hydrostatic tests for 

strength in the field because all pressure retaining parts, including any new 

parts they use, have all been hydrostatically tested during production at 

Farmingdale. Field service personnel only perform leak tightness tests on 

reassembled SOVs using air or nitrogen gas in the field. 

however, the field service personnel were building new SOVs at Farmingdale 

from new factory parts that had not yet been hydrostatically tested. 

hydrostatic test and its associated preparations and restorations are included 

in Section 3.2 of Target Rock Procedure TRP-4754 (currently Revision D, dated 

August 3, 1993). Accordingly, the field service technicians reportedly 

assembled the pressure-retaining parts including a valve body, bottom plate, 

bonnet/core tube assembly, and solenoid assembly (includes the bo1 ting flange) 

into what is called a shell assembly (without the internals, i.e., plunger and 

adjusting rod) in accordance with Step 3.2.1 of TRP-4754 in preparation for 

performing the hydrostatic test. After performing the hydrostatic test itself 

per Step 3.2.2, the procedure step then called for cleaning and drying the 

shell assembly (without internals) in accordance with TRP-1595. The traveller 

had a signoff block for completion of the hydrostatic test, but since t:ie 

drying requirement was included with the hydrostatic test step, there was no 

separate signoff on the traveller for drying. 

In this case 

The 

The work performed by the field service personnel on the SOVs in question was 

not documented completely on the factory traveler. 

the field service personnel apparently used the field procedure instead as 

overall guidance. 

hydrostatic test had to be repeated on the SOVs because the records of the 

original hydrostatic tests were misplaced, although they were found later. 

Target Rock concluded that the SOVs were not properly dried by the field 

service personnel after the second set of tests, most probably because the 

circumstances strongly suggested that the internals (solenoid plunger and 

adjusting rod assembly) were not removed for the second hydrostatic tests. 

Hydrostatic testing of SOVs with their internals installed is not prescribed 

by factory procedclres because !t would trap water in the air gap between the 

plunger and the fixed core at the end of the core tube and along the annulus 

formed by the plunger and core tube. Target Rock’s standard drying procedure, 

Target Rock stated that 

Target Rock records from October 1993 show that the 

4 

132 
using compressed air or nitrogen on SOVs without their internals (i.e, socalled shells), would not be effective in removing all moisture from a fully 

assembled SOV. The records do not explicitly state that the drying step was 

completed, but even if it had been, normal drying methods would not likely 

have been able to remove all the trapped water. 

Target Rock concluded that the internals were not removed because that would 

have meant having to reperform all the functional adjustments and tests on the 

SOV. However, performing the hydrostatic tests on the SOVs with the internals 

installed would be contrary to Target Rock procedures which would constitute a 

nonconformance with respect to Criterion V, "Instructions, Procedures, and 

Drawings," of Appendix 6 (quality assurance) to Part 50 of Title 10 of the 

Code of Federal Reaulat ions (10 CFR Part 50, Appendix 6). In addition, 

measures established to ensure that nonconforming material is not used were 

not adequate. Although Target Rock's postulated root cause scenario fits all 

the evidence and is the most logical explanation for the failure of the SOVs, 

it is not known for certain that the events took place in the manner 

postulated. Nevertheless, nonconforming material (i.e., SOVs containing 

residual water) was supplied, later used, and failed in an NRC-licensed 

facility. Therefore, the activities affectsng quality related to the 

preparation of this material for supply were not adequately controlled due 

either to procedural inadequacies, or noncompl iance, or both (95-01-01). 

As corrective action, Target Rock prepared ii problem report in accordance with 

its procedures. The inspectors reviewed the problem report, PR-018 dated 

February 16, 1995. 

failure of the field personnel to follow the procedure which was routinely 

used successfully by its factory personnel. 

identified four actions: 

Target Rock concluded that the problem was caused by the 

To prevent recurrence Target Rock 

e Revise the field procedure TRP 3959 to specify that work done in the 

factory would be done fully in accordance with established factory 

procedures regardless of who was doing the work, production personnel or 

field service personnel. 

Target Rock personnel stated that they would require factory work to be 

done only by factory procedures and not a combination of field and 

factory procedures. 

that the procedural corrective actions were completed as of August 1, 

1995. 

Target Rock reported subsequent tc the inspection 

e Train/Retrain field service personnel for proper assembly and test of 

production hardware (most field servic:e personnel have had factory 

experience) . 

The inspector reviewed the completed training record dated June 2, 1995, 

and found it acceptable. 

e Revise the assembly and test procedure, TRP 4574, to add a specific 

sign-off for cleaning and drying. 

The inspector reviewed the procedure c:hange and found it acceptable. 

5 

133 
0 Advise all affected customers to perform a "click" test of the Model 

1/2-SUS-S-02 valves prior to installation of the valves in plant. A 

click test consists of a bench energization of the SOV and listening for 

a click, Le., the sound of its actuation indicating free movement of 

the SOV's solenoid plunger/stem and disc assembly. 

The inspector reviewed a sample letter dated May 1, 1995, typical of the 

letters sent to all Target Rock customers, advising them of the problem, 

Target Rock's view of its limited nature, and the recommendation for a 

click test and found it acceptable. 

The inspectors considered the Target Rock problem report corrective actions to 

be appropriate and adequate. 

The inspectors noted that the Target Rock evaluation of reportability in the 

problem report concluded that the problem was not reportable under 10 CFR 

Part 21. The inspectors agreed that the problem was not necessarily 

reportable to the NRC under 10 CFR 21.21(a), but may still be reportable to 

all affected licensees or purchasers under 521.21(b). The inspectors 

considered that the rationale for Target Rock's reportability determination 

was in error. The stated rationale was that the failure of the solenoid did 

not affect the "automatic or safety function" of the relief valve. The 

inspectors pointed out that although the automatic function of the valve 

(pressure relief on high pressure) was not affected, the valve's other safety 

function, to respond to a manual signal (or in some installations, to the 

automatic depressurization system signal, although not at Cooper) to open and 

quickly reduce system pressure (in certain reactor event scenarios) was 

affected. 

that the problem was not reportable for the reason of its safety function not 

being affected. The inspectors pointed out that 10 CFR 21.21(b) requires the 

suppl ier to notify purchasers when the suppl ier cannot perform the safety 

evaluation. 

requirement. Further, since only Cooper had been affected and since Cooper 

had notified the NRC, no further action except documentation appeared to be 

required under Part 21. 

Therefore, it was not accurate for the problem report to conclude 

Since Target Rock had notified its customers, it had met that 

The inspectors further questioned Target Rock personnel regarding their 

rationale for considering that the problem was limited to the three valves 

supplied for Cooper. Target Rock representatives stated that they had 

interviewed the field service supervisor and his recollection was that no 

other similar job, requiring field service personnel to hydrostatically test a 

solenoid valve, had been performed. Further, in their field work, which 

involved disassembly of about ten solenoids a year at different facilities, 

field service personnel reportedly had seen no other evidence of corrosion in 

the solenoid valves. The inspectors also interviewed the Field Service 

Supervisor who confirmed this information. Additionally, Target Rock reported 

that it had experienced no other test failures during its routine performance 

of periodic relief valve testing for the facilities with the type of MSSRVs in 

question and stated it had received no other reports from customers regarding 

any other failures to operate. 

reviewing slightly more than one year's worth of work records to further 

verify that no other solenoid valves had been hvdrostatically tested by field 

The field service supervisor reported 

6 

134 
I personnel. Also, consultation with cognizant NRC staff members and review of 

NRC historical information did not reveal any other instances of similar 

failures of Target Rock SRVs. 

Target Rock rationale for considering the problem to be an isolated case was 

reasonable. 

Therefore, the inspectors concluded that the 

~ 

3.2 Re1 i ef Val ve Nai n SDri ng Re1 axat i on and Ti D Breakacre 

The inspectors also discussed with Target Rock representatives a problem with 

relief valve main spring apparent relaxation and spring tip breakage. 

problem was described in Target Rock Problem Report PR-20, dated June 14, 

1995. The problem involved apparent spring relaxation and tip breakage that 

was observed after extensive testing. The 1 imited steam generation capacity 

of the Target Rock test facility requires that the SRV discharge flow be 

restricted in order to maintain the pressure necessary to confirm proper SRV 

operation through its operating cycle. The discharge flow is reduced to 

within the steam flow capacity of the test facility by installing a restrictor 

in the discharge port of the SRV under test. The problem revealed itself when 

an SRV under test (after about 3000 actuations) failed to close. Disassembly 

showed that the main disc return spring free length was reduced, the normally 

circular cross section spring had slightly flattened faces between coils, and 

some small pieces of the spring tip (on the end away from the disc) had broken 

off. Target Rock examined the circumstances under which the failure occurred 

and determined that under test conditions, without the normal flow, the 

differential pressure normally felt across the main disc is significantly 

reduced, and in fact immediately goes nearly to zero as soon as the disc first 

lifts off its seat. With little flow-generated differential pressure to 

retard its motion, the disc, stem and main operating piston assembly literally 

slam open with the high-pressure opening force across the piston. 

piston’s motion is terminated when it contacts the head of the main operating 

cylinder. 

annular recess provided for it in the cylinder head, but the tremendous 

momentum imparted to the return spring coils or turns during testing causes 

them to bunch up at the non-piston end of the spring until most of them slam 

into one another. 

relaxation (reduction of free length), flattened inter-coil surfaces, and tip 

breakage experienced on the affected test SRV. 

The 

The 

The spring is not normally completely collapsed because of an 

According to Target Rock, this caused the apparent 

The Target Rock engineer determined that without the discharge flow 

restriction (as in the plant) the differential pressure across the valve 

generated by unrestricted flow would allow the valve to close normally in an 

operating environment. 

cause scoring in the operating cylinder, but; that the phenomenon had not been 

seen or reported. 

first observed, a fragment of a spring tip was passed through the pneumatic 

actuator of the SRV under test, but the Target Rock engineer conceded that it 

was possible for such a fragment to become lodged in the valve internals 

during testing. In this case, the Target Rack engineer explained, although a 

lodged fragment would not be likely to affect automatic self actuation of the 

valve on high pressure in the plant (the safety relief function), and it would 

not affect reclosing under normal flow conditions as it did under the 

restricted flow during bench testing, it could possibly affect the ability of 

7 

He further stated that spring tip fragments might 

During the Target Rock testing in which the problem was 

135 
the valve to be opened remotely (electro-pneumatic operation), which could 

impact operability of a valve used for the ADS. 

Target Rock initially identified this potential problem several years ago, but 

had not yet reported it because Target Rock had determined that it was aware 

of the condition of all its valves because Target Rock services the valves and 

monitors their condition in the field and the condition had not manifested 

itself until after 3000 test operations of an SRV using reduced flow. The 

Target Rock problem report on this issue indicated that Target Rock planned to 

issue a service bulletin to all potentially affected facilities by August 16, 

1995, to alert them to the problem. To be conservative, the bulletin will 

recornraend replacement of SRV main disc springs after 100 actuations (bench 

test, restricted flow actuations). Target Rock explained that it had decided 

to issue the service bulletin (and possibly also to recomnend that GE NE issue 

a service information letter (SIL) pursuant to the requirements of 10 CFR 

21.21(b). The NRC will follow up on Target Rock's action on this issue and 

may review it in a future inspection (Open Item 99900060/95-01-02). 

3.3 10 CFR P art 21 Proc edures 

The inspector reviewed the effective revision of procedures adopted pursuant 

to 10 CFR 21.21(a), contained in QCI-1306, Revision Cy dated August 10, 1994, 

and found that QCI-1306 was generally up to date and consistent with the 

requirements of the regulation. However the inspector noted some weaknesses 

in the procedure, which left uncorrected, could fail to prevent violation of 

certain provisions of 10 CFR Part 21. In one instance, the weakness would 

have previously constituted a Severity Level V violation of 10 CFR 21.21(a) in 

accordance with Supplement VI1 of Appendix C to 10 CFR Part 2, the previous 

enforcement guidance of the NRC's Rules of Practice. However, under the new 

NRC enforcement policy (as promulgated in NUREG 1600) that became effective on 

June 30, 1995, this weakness constituted a violation of minor significance and 

is being treated as a Non-Cited Violation, consistent with Section IV of the 

new NRC Enforcement Policy. The specific weaknesses noted in QCI 1306 were as 

follows: 

3.3.1 

of the Energy Reorganization Act of 1974, and 10 CFR 21.21. 

sections of Part 21, also relevant to Target Rock scope of activities are 

omitted. 

Paragraph 6.2 requires reading and understanding QCI-1306, Section 206 

However, other 

3.3.2 Apparently pursuant to 10 CFR 21.6(b), Paragraph 7.l(a) of QCI-1306 

calls for posting a notice that describes the instruction, identifies those to 

whom reports should be made, and states where the instruction and its socalled "implementing" reports may be found and examined. However, including 

in the notice the location of implementing reports is not consistent with 

10 CFR 21.6(b) which requires that the notice posted in addition to Section 

206 and in lieu of the regulation and the procedures adopted pursuant to the 

regulation include a description of the regulation as well as the procedures. 

In addition, §21.6(b) requires that the notice state where "they" may be 

examined. The language of Paragraph 7.l(a) of QCI-1306 indicates that Target 

Rock interpreted the word "they" in §21.6(t) to be referring to its immediate 

8 

136 
antecedent notin, i.e., the reports. 

§21.6(b), in addition to identifying persons to whom reports should be made, 

is to make employees aware of 10 CFR Part 21 and the procedures adopted 

pursuant to it, the purpose and function of the regulation and the procedures 

and where the regulation and the procedures may be examined. 

states where the procedures and regulation may be examined because the notice 

is posted in lieu of the entire regulation and the entire procedures when 

posting them is not practicable. Making employees aware of where Part 21 

reports may be examined was not intended to be a required function of the 

notice. 

reports. " 

The purpose of the notice allowed by 

The notice 

Finally, it was not clear what was meant by the term "implementing 

3.3.3 Paragraph 7.3.2 of QCI-1306 required that the General Manager or 

designee(s) be "informed.. .within 5 working days after completion of the 

evaluation." Although it appeared that the General Manager was to be informed 

of the conditions stated in Paragraph 7.3.1, it was not stated what the 

General Manager was to be informed of, nor if the evaluation being referred to 

was the one described in Paragraph 7.3.3. While it is within Target Rock's 

prerogative to require that its General Manager be informed of so-called 

potential defects, deviations, or failures to comply, QCI-1306 lacked a 

provision required by §21.21(a) to be included in procedures adopted pursuant 

to the regulation, to implement §21.21(a)(3) which explicitly required that a 

director or responsible officer be informed of a defect or a failure to comply 

associated with a substantial safety hazard within 5 working after completion 

of the evaluation described in 521.21(a)(l). 

In response to this concern, Target Rock explained that its General Manager is 

also its President (and also a Vice President of Curtis-Wright Corporation, of 

which Target Rock is a wholly-owned subsidiary). The General Manager being 

also the President of Target Rock satisfied the director or responsible 

officer requirement, but the inspector pointed out that using the words "or 

designees" without making it clear that such designees must also meet the Part 

21 requirement of being themselves directors or responsible officers may lead 

to failure to comply with §21.21(a)(3) should a defect be reported only to an 

i ne1 igi bl e designee. 

The inspector also noted that other language in QCI-1306 was quite effective, 

particularly Paragraph 7.3, and noted the significant strength that Target 

Rock's QA procedures governing control of nonconforming materials and 

corrective action were integrated or fed into the Part 21 process. The 

inspectors found no instances in which Target Rock had failed to handle and 

document deviations or failures to comply in a manner inconsistent with the 

requirements of Part 21. 

Finally, the inspectors had noted in the review of problem reports, the one 

associated with the MSSRV issue in particular, that Target Rock made a 

practice of making what amounted to a recomnendation to its customers 

regarding reportability of a given issue pursuant to Part 21. The inspectors 

pointed out that the statement of this type in the MSSRV solenoid problem 

report was misleading because the evaluation that formed the basis for the 

conclusion of -0nreportabil ity did not consider all possible safety functions 

9 

137 
of the electropneumatic controls of the MSSRV. Although the failed valves at 

Cooper were not in the ADS group, the inspectors pointed out that, in general, 

Target Rock as a vendor (other than an NSSS or A/E), would only be expected to 

provide affected licensees or purchasers with all pertinent information it has 

relative to a deviation or failure to comply as provided for in 921.21(b) and 

would not be expected to suggest that something is not reportable unless 

Target Rock were sure that it is qualified to perform the evaluation described 

in $21.21(a). 

4.0 PERSONS CONTACTED 

Andrew L. Szeglin, 

Edward Champey, Jr 

Robert E. Glazier, 

James D. White, Sa 

Senior Design Engineer 

, Director, Qual i ty Assurance 

Manager, Quality Engineering 

es and Service Manager 

10 

138 
' ***** '- 

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555.oo01 

September 14, 1995 

Mr. Ronald H. Koga 

General Manager 

Comnercial Nuclear Fuel Division 

Westinghouse Electric Corporation 

P.O. Box 355 

Pittsburgh, PA 15230-0355 

SUBJECT: NONPROPRIETARY VERSION OF NRC INSPECTION REPORT 99900005/95-01 

Dear Mr. Koga: 

This letter transmits the nonproprietary version of the U.S. Nuclear 

Regulatory Conmission's (NRC's) Inspection Report No. 99900005/95-01. Our 

letter to you dated July 25, 1995, transmitted the original (proprietary) 

version of the report. On the basis of our discussions and review of the 

information in your August 23, 1995, letter (NTC-NRC-95-4535), and its 

enclosures (Application For Withholding Proprietary Information From Public 

Disclosure AW-95-874 and Affidavit AW-95-874), we have concluded that the 

specific values identified in your letter could be regarded as proprietary 

and, as such, were removed from the inspection report. In the revised 

nonproprietary (public) version of the report, the NRC has briefly summarized 

the deleted text. 

Your response to either this letter or our letter dated July 25, 1995, and 

their enclosures are not subject to the clearance procedures of the Office of 

Management and Budget as required by the Paperwork Reduction Act of 1980, 

Public Law No. 96-511. 

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

this letter and its enclosure will be placed in the NRC Public Document Room. 

If you have any questions concerning this matter, please contact Steven M. 

Matthews at (301) 415-3191. 

Si nce re1 y , n 

Robert. M. Gallo, Chief 

Special Inspection Branch 

Division of Inspection 

and Support Programs 

Office of Nuclear Reactor Ljulation 

Docket No.: 99900005 

Enclosure: Report No. 99900005/95-01 

139 
REVISED NONPROPRiebAKY VERSION 

REPORT NO. : 

ORGANIZATION: 

ORGANIZATIONAL 

CONTACT: 

NUCLEAR INDUSTRY 

ACTIVITY: 

INSPECTION DATES: 

LEAD INSPECTOR: 

OTHER INSPECTORS: 

REVIEWED BY: 

APPROVED BY: 

U. S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF INSPECTION AND SUPPORT PROGRAMS 

99900005/95-01 

Westinghouse Electric Corporation 

Energy Systems Business Unit 

Commercial Nuclear Fuel Division 

Pittsburgh, Pennsylvania 

Michele H. DeWitt, Manager 

Qual ity Pol icy Deployment 

Qual ity and Strategic Management 

Energy Systems Business Unit 

Westinghouse Comercia1 Nuclear Fuel Division provides 

pressurized-water reactor reload core designs and 

reload safety analysis, fuel assembl ies, fuel-related 

core components, zirconium alloy fuel clad tubing, and 

fuel-related inspection services to the U.S. nuclear 

i ndustry . 

CNFD/Westinghouse Energy Center - February 5-10, 1995 

CNFD Specialty Metals Plant - February 6-10, 1995 

CNFD Columbia Plant - February 27-March 10, 1995 

CNFD Western Zirconium Plant - March 20-24, 1995 

Steven M. Matthews Ddte/ 

Vendor Inspection Section (VIS) Special Inspection Branch (PSIB) 

David H. Brewer, VIS/PSIB 

Daniel W. Brite, Battelle Pacific Northwest Laboratory 

Dr. John F. Carew, Brookhaven National Laboratory 

Ramon L. Cilimberg, Parkter, Inc. 

Rodney 1. Grow, Parhter, Inc. 

Edward D. Kendrick, Reactor Systems Branch, Division 

of Systems Safety and Analysis 

Fatrick S. Lacy, Parimeter, Inc. 

Roger kndez, VIS/PSIB 

Dr. Partha Neogy, Brookhaven National Laboratory 

/ 

Gregory /ea C- Cwalina. Chief, VIS/PSIB Date 

Robert M. Gallo, Chief, PSIB/DISP Date 

ENCLOSURE 

140 
CONTENTS 

1 

2 

3 

SCOPE AND SUMMARY OF INSPECTION FINDINGS 

1.1 Violations ............ 1.2 Nonconformances .......... 1.3 Weaknesses and Observations .... 1.4 Open Item ............. 1.4.1 Open Item 95-01-01 ..... 

STATUS OF PREVIOUS INSPECTION FINDINGS . 

..... 

..... 

..... 

..... 

..... 

..... 

.. 

.. 

..... 

2.1 CNFD Specialty Metals Plant . 99900005/86-01 

2.1.1 Nonconformance 86-01-01 (CLOSED) . 2.1.2 Nonconformance 86-01-02 (CLOSED) . 2.1.3 Nonconformance 86-01-03 (CLOSED) . 2.1.4 Unresolved Item 86-01-01 (CLOSED) . 

.. 

.. 

.. 

.. 

2.2 CNFD Columbia Plant . 99900005/92-01 . . 2.2.1 Nonconformance 92-01-01 (CLOSED) . 

.. 

.. 

INSPECTION FINDINGS AND OTHER 

3.1 

3.2 

3.3 

3.4 

Background ...... Entrance Meetings. Inter 

Meeting ........ 

CNFD/Westinghouse Energy 

3.3.1 Reload Core Design 

OMMENTS .... .. 

........ 

m Exit Meetings. 

........ 

Center ..... 

.. 

and 

.. 

.. 

.. .. .. 1 

.. 

Final 

... 

... 

and Reload Safety Analysis 

3.3.2 Fuel Assembly Mechanical Design . 3.3.3 Error Free Performance Team ... 3.3.4 Training ............ 3.3.5 Fuel-Related Inspection Services 

3.3.6 Conclusions ........... 

CNFD Speciilty Metals Plant ...... 

3.4.1 Product Assurance ........ 3.4.2 Customer Requirements ...... 3.4.3 Fuel Clad Tubing Production ... 3.4.4 Quality Services Lab ...... 3.4.5 Instrument Lab ......... 3.4.6 Corrective Actions ....... 3.4.7 Conclusions ........... 

... 

€xi t 

... 

... 

Process 

2 

2 

2 

2 

3 

3 

3 

3 

4 

4 

5 

6 

6 

24 

29 

30 

30 

31 

32 

32 

33 

33 

36 

37 

38 

38 

141 
3.5 

3.6 

3.7 

APPENDIX A 

APPENDIX B 
1 SCOPE AND SURUARY OB INSPECTION FINDINGS: 

During this inspection, the NRC inspection team (team) evaluated the 

Westinghouse Electric Corporation (W), Energy Systems Business Unit (ESBU), 

Commercial Nuclear Fuel Division (CNFD) mdagement, staff, and quality 

programs and the implementation of those programs related to pressurized-water 

reactor (PWR) reload core designs and reload safety analysis, fuel assemblies, 

fuel-related core components, zirconium a1 loy (zircaloy) fuel clad tubing, and 

fuel-related inspection services. These inspections were conducted to provide 

a basis for confidence that these items and services supplied to the U.S. 

nuclear industry would perform their safety function. 

consisted of the following: 

The inspection basis 

(a) General Design Criterion (GDC) 30, "Reactor Design," and GDC 12, 

"Suppression of Reactor Power Oscillations," of Appendix A, "General Design 

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

Federal Requlations (10 CFR Part 50). 

(6) Appendix B, "Quality Assurance Criteria for Nuclear Power Plants 

and Fuel Reprocessing Plants," to 10 CFR Part 50. 

(c) Part 21, "Notification of Failure tn Ccnply or Existence of a 

Defect," of 10 CFR. 

(d) W ESBU Topical Report, documented in WCAP-8370, "Quality Assurance 

Plan," Revision 12A, dated April 1992, approved by the NRC on April 23, 1992, 

as meeting the requirements of Appendix B to 10 CFR Part 50, and as amended by 

the updated organizational charts submitted by W on June 14, 1994, hereafter 

referred to as the "QA topical report," 

1.1 Violations 

No violations were identified during this inspection. 

1.2 Nonconformances 

No nonconformances were identified during tlhis inspection. 

1.3 Weaknesses and Observations 

A few weaknesses, primrily in the procedural conformance of certain 

activities that affect quality, were identified during this inspection. 

Neither the weaknesses nor the observations described in the inpsection report 

require any specific action by or written response from CNFD. 

-1- 

143 
1.4 ODen Item 

1.4.1 Open Item 95-01-01 

As described in Section 3.5.10.2, "Chemical Laboratory," of this report, the 

team observed weaknesses in certain calibration practices of CNFD COLA. The 

team requested CNFD to notify the NRC when its analysis of the calibration 

practices, and corrective actions taken if any, have been completed. 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

2.1 CNFD SDecialtv ktals Plant - 99900005/86-0€ 

During an inspection of CNFD Specialty Metals Plant (SMP) conducted on 

August 18-20, 1986, an NRC inspection team determined that certain CNFD SMP 

activities were not conducted in accordance with NRC requirements. The 

following nonconformances, issued by the staff on October 3, 1986, and the 

associated corrective actions taken by CNFD SMP were evaluated by the team 

during this inspection. On the basis of its evaluation of the CNFD SMP 

corrective actions, the team determined that each of the nonconformances was 

cl osed. 

2.1.1 Nonconformance 86-01-01 (CLOSED) 

Contrary to Section 17 of WCAP-8370/7800, Revision 10A/6A, dated August 1984, 

the standard used in ultrasonic testing (UT) to validate tube ovality was not 

serialized and maintained under the equipment calibration control system nor 

were the dimensions of the standard traceable to the National Bureau of 

Standards. 

During its evaluation of the CNFD SMP tube reduction process, the team found 

that for the inspections observed, CNFD SMP had appropriately implemented its 

inspection requirements with the correct standards. Therefore, the team 

concluded that CNFD SMP corrective actions taken appeared adequate to ensure 

that the appropriate inspection standards were used to deterr;l!i:; "re 

acceptability of tube hollows. 

2.1.2 Nonconformance 86-01-02 (CLOSED) 

Contrary to Section 17 of WCAP-837017800, Revision 10A/6A, dated August 1984, 

the data for vertical 1 inearity, horizontal 1 inearity, and calibrated 

attenuation were not recorded on the calibration data reports, dated 

February 24, 1986, and March 27, 1986, for two UT flow detectors (Nos. 33 

and 40). 

During its evaluation of the CNFD SMP tube reduction process, the team 

observed that CNFD SMP had appropriately implemented its inspection 

requirements with the correct UT data recorded on calibration data reports. 

Therefore, the team concluded that CNFD SMP corrective actions appeared 

adequate to encure that the appropriate UT data were recorded on calibration 

reports to suppc-t the acceptability of tube hollows. 

-2- 

144 
2.1.3 Nonconformance 86-01-03 (CLOSED) 

Contrary to Section 17 of WCAP-8370/7800 Revision 10A/6A, dated August 1984, 

Quality Services Lab Procedures QS-213, 3s-249, QS-261, and QS-262 contained 

references to QS-118, which became an obsolete procedure on July 7, 1985, when 

it was superseded by Produce Assurance (PA) procedure PA-103. 

The team determined that CNFD SMP maintained current documents in computer 

files, thereby permitting the documeqts to lbe searched for specified words Or 

phrases by the search routine in the program. This technique provided the 

basis for searching current documents for references that had been changed. 

Based on evaluation of CNFD SMP document management, the team considered this 

nonconformance closed. 

2-1.4 Unresolved Item 86-01-01 (CLOSED) 

Paragraph 4.5.3 of material specification NFD-31008, "Seamless Zircaloy-4 

Tubing," Revision 28, dated May 16, 1986, required, in part, that tube ovality 

not exceed 0.0013-inch total indicator reading (TIR). 

procedure QC-301 , "Final Inspection - U1 trasonic Dimensional Setup and 

Calibration," Revision J (QC-301), outlines the method used to certify that 

this dimensional requirement is met. Although Frccsdure QC-301 is accurate, 

it is unclear whether the actual requirement of TIR is being met. 

The team determined that Revision L of procedure QC-301, dated July 6, 1989, 

requires that the TIR reading be made in a helical pitch rather than a 

stationary plane. 

had assumed that the TIR reading was to be taken in a stationary plane. 

Paragraph 6.4 of 

This customer requirement clarifies this concern; the team 

2.2 

During an inspection of CNFD Columbia Plant (COLA) conducted from 

January 13-17, 1992, an NRC inspection team determined that certain CNFD COLA 

activities were not conducted in accordance with NRC requirements. 

following nonconformance, issued by the staf,f on February 6, 1992, and the 

associated correccive actions taken by CNFD COLA were evaluated bq the team 

during this inspection. 

corrective actions, the team determined that the nonconformance was closed. 

CNFD Columbia Plant - 99900005/92-01 

The 

On the basis of its evaluation of the CNFD COLA 

2.2.1 Nonconformance 92-01-01 (CLOSED) 

Contrary to Criterion V of Appendix 6 to 10 CFR Part 50 and Section 7 of CNFD 

COLA Administrative Procedure CA-006, "Columbia Plant Training Policy," 

Revision 3, dated October 12, 1990 (CA-006), an operator performed the prepl ug/pre-weld operati on for several weeks in accordance with an outdated 

revision to the governing procedure before acknowledging the correct revision 

to the procedure. 

-3- 
As part of the corrective actions, CNFD COLA removed references to the 5 day 

rule. The revised procedure CA-006 required that operations personnel review 

and sign-off a new or revised procedura before performing the operation 

covered by the procedure. In addition, the team was informed by the operation 

supervisor that no revision to a procedure would take effect until the 

Thursday of any week, allowing sufficient time to inform operators of pending 

changes to the procedures. Duri ng the inspection , several minor observations 

were noted regarding the sign-off of some procedures as discussed in this 

report; however, product quality was not affected. 

evaluation of the CNFD COLA corrective actions, the team determined that the 

nonconformance was closed. 

On the basis of its 

3 INSPECTIOW FINDINGS AND OTHER COMENTS 

3.1 Backwound 

Westinghouse is a diversified, technology-based corporation founded in 1886. 

The CNFD, a charter recipient of the Malcolm Baldrige National Qualiiy Award, 

is one of several divisions and business areas of the !f ESBU. Other 

organizations within the y ESBU that perform fuel-related activities evaluated 

during this inspection are the Nuclear Technology Division, Nuclear Services 

Division, and Quality and Strategic Management. 

In addition to completed fuel assemblies, CNFD supplies zircaloy fuel clad 

tubing, fuel-related core components, and engineering services and training to 

licensees, other fuel vendors, and utilities worldwide. 

Over 2,000 people were employed by CNFD at four locations and CNFD was the 

only fully integrated supplier of nuclear fuel products and services in the 

United States. At the CNFD Western Zirconium Plant (WZ) in Ogden, Utah, 

zircon sand was converted to zircaloy, which was used to fabricate fuel clad 

tubing and other fuel-related core components. 

was shipped to the CNFD SMP in Blairsville, Pennsylvania, for finishing. 

Finished fuel clad tubing was then shipped from CNFD SMP to CNFD COLA in 

Columbia, South Carolina, where completed fuel assemblies and frre1-related 

core components were fabricated. The CNFD reload core design and reload 

safety analysis engineering personnel were located at the division’s 

headquarters, the Westinghouse Energy Center, in Monroeville, Pennsylvania. 

Exttwded zircaloy from CNFD WZ 

The plant managers of CNFD WZ, CNFD SMP, and CNFD COLA all report to the 

general manager of CNFD. 

philosophy; that is, product flowing from one plant site to the other was 

treated as though it came from another department within the same 

organization. As implemented by CNFD, this philosophy meant that product 

parameters were not reinspected when products arrived at the subsequent plant 

site; material was checked for shipping damage and placed in production. 

CNFD was managed with a “one-roof manufacturing” 

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146 
The "one-roof manufacturing" philosophy was implemented in several other ways. 

For example, the Fuel Performance Technology group (located at the 

West inghouse Energy Center) of the Produc ,./Process Development & Design 

(P/PD&D) group of CNFD (P/PD&D management and most of its staff are located at 

CNFQ COLA) includes a Product Design/Development group that consists of 

materials engineers and mechanical engineers that spend most of their time at 

CNFD SHP. The P/PD&D Product Design/Devel opment group was responsible for 

material specifications, maintained at CNFU SMP, and drawing definition, 

maintained at CNFD COLA. This group interfaced with other organizations at 

CNFD SMP and CNFD COLA by regularly scheduled, informal telephone conferences, 

in which the P/PD&D Product Design/Development group cgllectively dealt with 

questions arising at any of CNFD plants. 

The Error Free Performance Team (EFPT), described in Section 3.3.4 of this 

report, was another way in which the CNFD "one-roof manufacturing" philosophy 

was implemented. A1 though EFPT management was located at the Westinghouse 

Energy Center, the EFPT membership and participation was from CNFD WZ, 

CNFD SMP, and CNFD COLA. 

3.2 Entrance Beetinas. Interim Exit Heetinas. and Final Exit Beeting 

For each of the following inspections, the team conducted an entrance meeting 

on the first day of the inspection. 

e 

e 

e 

e 

February 5-10, 1995 

February 6-10, 1995 

February 27 - 

March 10, 1995 

March 20-24, 1995 

CNFD/Westinghouse Energy Center 

4350 Northern Pike 

Monroeville, Pennsylvania 

CNFD Specialty Metals Plant (SMP) 

Westinghouse Road, R.D.4 

Blairsville, Pennsylvania 

CNFD Columbia Plant (COLA) 

5801 Bluff Road 

Col umbi a, South Carol i na 

CNFD Western Zirconium Plant (WZ) 

10,000 West 900 Street 

Ogden, Utah 

During each of the entrance meetings, the team met with members of the CNFD 

management and staff, discussed the scope of the inspection, reviewed the 

team's and CNFD's responsibilities for handling proprietary information, and 

established contact persons for the team within the management and staff of 

the applicable CNFD organization. 

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147 
During the inspection periods described above, the team conducted a 

performance-based inspection of CNFD through technically directed observations 

and evaluations of processes, activities, ad documentation. 

(a) examined technical documentation, procedures, and representative records, (b) interviewed CNFD personnel, (c) held discussions with CNFD personnel, 

(d) listened to presentations by CNFD personnel, and (e) made other 

observations. The specific areas examined, the documentation reviewed, and 

the team findings are described in this report. 

in and who were contacted during this inspection are listed in Appendix A to 

this report. 

The team 

The persons who participated 

On the last day of each of the inspection periods describea above, the team 

conducted an interim exit meeting to outline to CNFD management and staff 

major concerns, weaknesses, strengths, and observations identified by the team 

during that portion of the inspection. 

During its closing exit meeting at the Westinghouse Energy Center in 

Monroeville, Pennsylvania, on April 13, 1995, with CNFD management arld staff, 

the team sunmarized the inspection findings, weaknesses, strengths, open items 

and observations. 

3.3 CNFD/Mestinahouse Enerw Center 

In inspecting the CNFD activities at the Westinghouse Energy Center, the team 

evaluated (a) the reload core design and reload safety analysis process, 

(b) the fuel mechanical design process, and (c) fuel-related inspection 

services. 

3.3.1 Reload Core Design and Reload Safety Analysis Process 

CNFD produces 30 to 40 reload core designs and related engineering services to 

support licensing and plant operation? each year; and this process was 

facilitated through a high degree of automation, including an automated Calc 

Note system that was geared towards error reduction and uniform documentation. 

Each reload core design requires a complete core design and safety ana1:jsis. 

The results of the re:oad core design and reload safety analysis were 

documented in the Reload Safety Evaluation Report (RSE) provided to the 

licensee. The reload analyses were performed with methods that were 

documented in NRC-approved topical reports. Both full-scope and spl it-scope 

reload design evaluations were performed. 

licensee performs selected parts of the reload core design (typically the core 

neutronics analysis) and CNFD performs the remaining analyses required to 

complete the reload core design. 

computer codes and methods to licensees and provided the training required for 

the proper application of these methods. 

In split-scope evaluations, the 

CNFD had released some of its core design 

In inspecting the CNFD reload core design and reload safety analysis process, 

the team evaluated the activities of (a) the CNFD, performed by the Core 

Engineering group located at the CNFD/Westinghouse Energy Center and by the 

Product/Process De-:elopment and Design (P/PD&D) group located at CNFD COLA, 

and (b) the Nuclear Technology Division (NTD) performed by the Nuclear Safety 

Analysis (NSA) group. Where the activities of these groups related to the 

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148 
reload core design and the reload safety analysis process, the team conducted 

a detailed evaluation of those activities by selecting certain reload core 

design packages. 

software controls, design verifications, design change controls, interface 

controis, and documentation and reccrds. 

core design packages a1 so covered the steady-state neutronics, thermalhydraulics (T/H) design, transient analysis, fuel mechanical performance, core 

monitoring, and set point analysis. 

The team evaluated the design inputs, design processes, 

The evaluation of selected reload 

Evaluation of the INCORE code (used to perform on-line core surveillance using 

in-core flux measurements) included the input preparation, transmittal of the 

INCORE data sets to the licensee, comparisons of predicted and measured 

neutron flux distributions, and licensee feedback. The performance of a 

reJoad core is evaluated during the operating cycle by monitoring the results 

of startup tests, critical boron concentration and core power distribution 

measurements, and coolant chemistry data obtained and transmitted by the 

licensee. 

core design. The team also evaluated the RSE report and the related Core 

Operating Limits Report (COLR). 

analysis computer codes and verified that NRC-approved codes were being used 

and that NRC-developed Safety Evaluation Report (SER) restrictions and 

ljmitations were being observed. The team found that the reload core design 

activities were generally being adequately performed, with the few exceptions 

noted below. 

These comparisons and test results were used to evaluate the reload 

The team evaluated the reload core design 

In order to optimize the reload core design evaluation, CNFD employed a 

bounding analysis approach in which the cycle-specific core design was bounded 

by a previously analyzed reference core design. In this approach, many of the 

cycle-specific safety analyses were not required and the reload core design 

evaluation was greatly simplified. Where a bounding reference analysis could 

not be identified for a particular reload core design, a cycle-specific 

analysis was performed for those aspects that were not bounded by a reference 

core design. 

The team examined certain reload core design packages by evaluating the reload 

core design and reloAd safety analysis process, starting from the end products 

(deliverables to the licensee). 

RSE and COLR. From these documents, results were selected and traced to their 

source to datermi,,? if the analysis process was perlormed in accordance with 

the procedures and was adequately documented. 

The two key deliverables examined wsre the 

In inspecting other reload core design packages, the team began at the front 

end of the design process by interviewing the cognizant project engineer, and 

then examining the project interface documents. These interface documents 

included those internal to the CNFD, the NTD, CNFD COLA, and the NRC 

licensees. 

To select the piant-specific reload core designs to be evaluated, the team 

reviewed current reload core design issues of special importance. The team 

identified more than 20 reload core design issues of special interest, among 

them the following: (a) vendor/l icensee ,interface concerns (e.g., spl it-scope 

designs) , (b) recent operational problems, (c) special-puryse fuel designs 

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149 
(e.g., flux suppression fuel assemblios), (d) new fuel designs, (e) recent 

licensing issues (e.g. , asymnetric roo cluster control assembly (RCCA) 

withdrawal), and (f) issues identified in recent Licensee Event Reports (e.g. , misalignment of wet annular burnable absorber (WABA) rods). These issues and 

their treatment in recent reload core designs were discussed during the 

in-itial meetings with CNFD. 

selected five reload core design packages to evaluate the reload core design 

and reload safety analysis process, and CNFD’s response to the most 

significant reload issues: 

On the basis of these discussions, the team 

0 South Carolina Electric 8 Gas Company, 

Virgil C. Sumner Nuclear Station Cycle 9 

0 Wisconsin Electric Power Company, 

Point Beach Nuclear Plant Unit 2 Cycle 21 

Zion Station Unit 1 Cycle 14 

0 Commonwealth Edison Company, 

0 Public Service Electric & Gas Company, 

Salem Nuclear Generating Station Unit 2 Cycle 9 

0 Houston Lighting & Power Company, 

South Texas Project Unit 2 Cycle 4 

3 3.1.1 Re1 oad Core Des i gn Process 

The CNFD reload core design process comprised four steps: 

Automated Linkage of the PHOENIX-P and Advanced Nodal Code (ANC), the 

PHOENIX-P code used to generate cross-sections, the ANC code used for twodimensional (20) radial and three-dimensional (3D) core analysis, and the 

PHIRE post-processing code for PHOENIX-P data banks); 

(a) Core Design and Steady-State Analysis (utilizing the ALPHA code for 

(b) Operational Strategy and Analyses (utilizin, the AYOLLU code for 1D 

(axial) core analysis), the VENUS code used to perform peaking factors 

synthesis for Constant Axial Offset Control (CAOC) analysis and Relaxation of 

Constant Axial Offset Control (RAOC) analysis, the ALUCARD code used to 

generate INCORE constants, and the INCORE code); 

(c) Fuel Management (utilizing the Advanced Loading Pattern Search 

(ALPS); and 

(d) Core Monitoring (utilizing the SPNOVA code and BEACON, the 4! OnLine Core Monitoring System code). 

The core design process was described in detail in the Methods Conawnication 

manual (METCOM) as well as in the individual code manual. The team found the 

four volume METCOH document to be comprehensive and rich in analytical and 

procedural detail. The METCOM documented the design and quality objxtives, 

responsibilities, and requirements for the reload process. The manual 

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150 
included summaries of certain codes, associated detailed modeling 

instructions, and selected reactor systems data. In addition, detailed 

procedures are provided for determining the, accident-specific input to the 

RSAC. 

in the y topical reports. 

The AETCOIM manual provides substa<itial iJ more detail than is included 

The METCOM manual is reviewed every 3 years and updated as necessary 

(typically every 6 months), as required by Engineering Procedure (EP) 

procedure EP-105, “Design Manuals,” Revision 5, dated February 1, 1993 

(EP-105). In response to this requirement, CNFD established a METCOn team 

that continuously reviews and updates the manual and responds to users’ needs. 

This activity typically results in approximately 10 METCOM revisions per year. 

During the course of the team’s review, several minor METCOM omissions or 

errors were identified and discussed with the Core Engineering staff. 

Westinghouse CNFD stated that these would be eva?uated and considered for 

possible inclusion in future METCOM updates. 

The reload core design process began when CMFD received the Reload Schedule 

and Energy Requirements (RSER) document from the 1 icensee. 

determined the fuel enrichment, the integral fuel burnable absorber (IFBA) 

design, the fuel rod design, burnup limits, number of fuel assemblies, and 

loading patterns. Fuel rod design limits were co,il”irmed by the Fuel Analysis 

group in Core Engineering, and this confirmation was documented in the RSE. 

The Core Engineering, Core Design group determined the boron concentration, as 

documented in the Boron Design Requirements (BORDER). 

determined the peaking factors and power shapes and provided them to the Core 

Engineering, Fuel Analysis group for T/H design. The core design process also 

involved an extensive physics database including design data and hot zeropower (HZP) and hot full-power (HFP) data. The Fuel Analysis group provided 

the departure from nucleate boiling ratio (DNBR) and fuel temperature inputs 

to the Reload Safety Analysis Checklist (RSAC), an input to NTD. The Core 

Design group also determined the reactivity parameters, RCCA rod worths, and 

kinetics parameters, all of which were also inputs to the RSAC. 

P/PD&D group prepared the Design Evaluation Verification List (DEVL) and the 

Fuel Parameters Checklist (FPC) and provided these documents to Core 

Engineerins. T:- e Core Engineering, Fuel Licensing Integration grmp 

sumnarized any mechanical design changes identified in the DEVL and FPC and 

provJed that data to NTD. 

completed and transmitted to NTD by Core Engineering. The CNFD groups that 

provided input to the RSE were Core Engineering, Core Design, P/PD&D, and Core 

Engineering, Fuel Analysis (T/H analysis and fuel rod design). 

Licensing Integration group was responsible for completing the final RSE and 

incorporating the inputs from CNFD and NTD. 

Core Engineering 

The Core Design group 

The CNFD 

Also, the reload-specific RSAC document was 

The Fuel 

3.3.1.2 Reload Safety Analysis Process 

The CNFD reload safety evaluation methodology was based on WCAP-9272, 

”Westinghouse Reload Evaluation Methodology,” dated March 1978, and made 

extensive use of the RSAC. 

rout i ne eval uat ion of re1 oad core safety. I t uti 1 i zed a perturbation approach 

to determine whether key safety parameters for the reload core design (Le., 

design parameters that have non-negl igible jmpact on the safety sorformance Of 

The methodology provided the basic methods for 

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151 
the core) were bounded by values used in the reference reload safety analysis. 

The approach tended to minimize the effort spent in reanalysis at the expense 

of being somewhat over-conservat i ve . Uncertainties were included in most, 

though not a1 1 Because the overall conservatism 

of the reload safety evaluation methodology, the team judged that the neglect 

of explicit accounting of uncertainties in a few safety parameters did not 

alter the net conservatism of the approach. The RSAC did not contain explicit 

values of the key safety parameters for the reload core design, only a 

determination that the values were bounded (or not bounded) by those assumed 

in the reference safety analysis. CNFD argued that this process ensured that 

the appropriate safety margins were managed and controlled by a single group 

(the CNFD Core Engineering group). 

current values of the key safety parameters on the RSAC increased the chances 

of an error in the comparison to the reference values. However, the team did 

not uncover an instance where such an error had been made. This team 

observation requires no specific action nor written response. 

key safety parameter Val ues . 

The team felt that not recording the 

The reload safety analysis process began when the Reload Safety and Licensing 

Checklist (RSLC) document was received from the licensee. 

Initialization Questionnaire (RIQ) document was prepared by CNFD and 

transmitted to the licensee to confirm the current status of the plant, 

specifically with respect to safety-related operations and des.ign input 

values. NTD Fluid Systems group, using the BORDER input from CNFD Core 

Engineering, confirmed that the boron system design requirements were met. 

Using the identified fuel mechanical design changes and RSAC inputs from CNFD 

Core Engineering, NTD NSA performed any necessary loss-of-coolant accident 

(LOCA) analysis and Non-LOCA transient analysis, confirmed the safety 

analysis, and defined Technical Specification changes, if any. NTD NSA also 

provided its input to the RSE through CNFD Core Engineering Fuel Licensing 

Integration group. 

A Reload 

Many of the reload safety analyses evaluated by the team appeared to be the 

result of comparing the current reload core design parameters with earlier 

bounding analysis values. In many instances, the earlier bounding analysis 

was completed a number of years ago, in a different culture with a less 

controlled process, by engineers who are no longer with NTD. Thus, the team 

observed the potential for interface gaps between past work and present work; 

current work may be based on a weak understanding of the earlier work. This 

team observation requires no specific action nor written response. 

3.3.1.3 Virgil C. SuRlaer Nuclear Station Cycle 9 

In evaluating South Carolina Electric and Gas Company (SCE&G), Virgil C. 

Sumner Nuclear Station (Sumner) Cycle 9 reload core design and reload safety 

analysis, the team began by evaluating the end products; the Reload Safety 

Evaluation (RSE) and the Core Operating Limits Report (COLR). The team 

examined these documents to determine key or representative results to be 

traced to their source. Sumner Cycle 9 was a split-responsibility reload core 

design with the licensee, SCE&G, having responsibility for the nuclear design. 

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152 
The CNFD engineer functioned as verifier for the nuclear design Calc Notes. 

Interface documents were examined to determine the control of the split 

responsibility and the flow of information to and from the licensee. There 

were no Technical Specification changes fw Sumner Cycle 9 reload. 

The technology transfer and the control of software was examined by first 

reviewing the appl icable procedures in the CNCD Design Engineering Procedure 

Manual and the Software Engineering Methodology manual, Revision 11, dated 

October 25, 1994. The process was inspected by examining (a) the original 

technology transfer to SCE&G, (b) the most recent technology transfer to 

SCE&G, (c) the development and release of a new code, (d) the updating of an 

old code, and (e) examples of error reporting. All material inspected was 

found to 5e in compliance with the relevant procedures. A demonstration of 

the STATEPOINT software for configuration monitoring by the supervisor 

engineer of CNFD Core Engineering Techno1 ogy Product Services group showed 

that the configuration control process had been automated. 

(1) Thermal-Hydraulic Analysis 

The Thermal-Hydraulic (T/H) Design documented in Section 2.3 of the RSE was 

examined by discussions with the fuel analysis engineer and evaluating the 

relevant Calc Notes. A weakness was identified as a result of this 

inspection. 

CNFD Engineering Procedure (EP), as documented in EP-302, "Documentation and 

Verification of Design Analyses," Revision 5, dated November 1, 1992 (EP-302), 

Revision 42, November 30, 1994, required that analysis such as the T/H 

analysis for the reconstituted fuel, be documented. The team observed that 

the RSE for Summer Cycle 9 stated that a T/H evaluation for the fuel rod 

reconstitution had been performed. 

94-109-0, did not contain any discussion or !analysis to support the Cycle 9 

reconstituted fuel. 

without documenting the rational and analysis that showed the relevance of the 

topical to the current reload core design. 'The team observed that the 

referenced topical report, WCAP-13060-P-A, "Vest i nghouse Fuel Assembly 

Reconstitution Evaluation Methodology," dated July 1993, was applicable to the 

Summer Cycle 9 reload core design. 

cycle-specific justification was judged by the team to be a poor method that 

did not conform to CNFD engineering practice. 

This weakness was discgssed with CNFD personnel and they responded by adding 

to Calc Note T/H 94-109-0, Part 14 which identified the reconstituted fuel, 

the relevant CNFD Core Engineering methodology, and the engineering analysis 

and rationale for the Cycle 9 reconstituted fuel. The CNFD actions taken 

during the inspection satisfied the team's concerns. 

However, the relevant CNFD Calc Note, TJH 

The responsible engineer had relied on a topical report 

The reliance on topical reports without 

(2) Fuel Wechanical Design 

The Mechanical Design, documented in Section 2.1 of the RSE, was examined by 

discussions with the fuel analysis engineer, and the Calc Note was reviewed 

relative to the Fuel Rod Design Procedure Manual, Revision 4, dated April 

1993. This review identified a weakness. 

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I 153 
I 

The team observed that no’single document showed that the 11 fuel rod design 

criteria, specified in the Fuel Rod Design Procedure Manual, were satisfied 

for the Sulmner Cycle 9 reload core design. In contrast, the team found that 

checklists were used frequently in most otbr aspects of the reload core 

design evaluation process. Further inspection showed that three different M 

organizations had performed calculations for this specific reload and that 

collectively the 11 design criteria were satisfied. 

This lack of a single document, such as a checklist, that could be referenced 

in the RSE approval documentation was a weakness discussed by the team with 

Core Engineering personnel. CNFD responded by creating a fueJ design criteria 

checklist and stated that the checklist would be included in the next revision 

of the Fuel Rod Design Procedure Manual. Memo FA-95-052, dated February 9, 

1995, was issued instructing engineers to utilize the new checklist and to 

reference the checklist in the RSE sign-off documentation. 

taken during the inspection satisfied the team’s concerns. 

The CNFD actions 

3.3.1.4 Point Beach Nuclear Plant Unit 2 Cycle 21 

In evaluating Wisconsin Electric Power Company (WEPC) , Point Beach Nuclear 

Plant Unit 2 (Point Beach) Cycle 21, the team began at the front end of the 

reload core design process by interviewing the cognizant project engineer. 

This interview led to an examination of project interface documents. On the 

basis of its evaluation of these documents, the team determined that the Point 

Beach Cycle 21 reload core design and reload safety analysis was challenging 

in several ways : 

(a) Cycle 21 energy requirements increased from 11.2 gigawatt-days per 

metric tonne of initial uranium metal (GWD/MTU) to 11.8 GWD/MTU. This 

from 28 feed fuel assemblies to 29 assemblies. 

I required changing the reload core design during the Cycle 21 design process 

(b) The design of integral fuel burnable absorber (IFBA) (uranium 

dioxide (UO ) fuel pellets coated with zirconium diboride (ha,)) used a 

i longer length fuel rod. 

(c) Some of the IFBA loaded fuel assemblies have an asymmetrical IFBA 

loading gattern with a “half moon“ design. 

(d) Fuel rods with IFBA loadings had a lower weight percent (w/o) U, 

enrichment than the stirrounding fuel rod enrichment; however, of the two fuel 

assembly designs with the same enrichment one did not include IFBA and the 

other included IFBA. 

(e) The reload pattern included 12 fuel assemblies with hafnium (Hf) 

flux suppressor rods and four fuel assemblies with water displacement rods. 

Each of the four water displacement fuel assemblies contained 12 dumy 

zircaloy rods in the guide tubes. The Hf flux suppressor rods to be used were 

already at the Point Beach Unit 2 site, having been used in Cycle 20. The 

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154 
water displacement rods were to be shipped by CNFD COLA with the reload fuel 

assemblies. The licensee had the responsibility to load both of these fuelre1 ated core components (Hf flux supprEssor and water displacement rods) per 

the loading plan. 

(f) The existence of a qua Cycle 21 startup. The size of th t varied with Cycle 21 burnup. A tilt 

had also been observed during Cyc 

(< 2%). 

The team evaluated a number of Ca 

12 requiremeats for Calc Notes stated in procedure EP-302. The team 

determined that all Calc Notes evaluated, with two exceptions, were prepared 

and verified in accordance with the requirments of procedure EP-302. 

However, the team identified three weaknesses. One of the two exceptions 

noted above, Calc Note T/H-94-086-0, contained fuel rod design results, did 

not contain a checklist. The team examined this Calc Note to follow up the 

Sumer Cycle 9 reload core design evaluation and confirm whether the same 

weakness existed in the Point Beach Cycle 21 reload core design. 

because of the CNFD action taken (Memo FA-95-052) as a result of the team 

evaluation of the Summer Cycle 9 reload core design. taken with regard to this weakness. 

power ti 1 t which was observed during 

but of a significantly smaller size 

es and related documents, using the 

It did not 

No further action was 

The team evaluation of the related Calc Notes determined that all of the 

unusual design features were properly addressed in the reload core design and 

reload safety analysis. 

Examination of Calc Note WII-94-056-0, "Point Beach Unit 2 Cycle 21 

Reevaluation of the RSE due to a HFP tilt of 3.3%," confirmed that measured 

results obtained during startup and early-cycle operation were evaluated for 

impact on the pre-operation RSE results. 

potential causes of the tilt, e.g., the steam generator plugging imbalance 

between the two steam generators and a small burnup difference in reinserted 

fuel assemblies. Corrective actions for Point Beach Cycle 22 were described 

in the Calc Note. 

differences in fuel assemblies located symmetrically across the core could 

exacerbate a small tilt driven by steam generator imbalance or could mitigate 

the tilt. 

This reevaluation identified 

One of these was based om the recognition that small burnup 

Beta-effecti ve 

Beta-effective (Beff) is the isotopic and importance-weighted delayed-neutron 

fraction in core (delayed neutrons are neutrons emitted by fission products 

sometime after a fission). 

"Beta Effective for WIS21 RSAC," identified the following procedural and 

technical weaknesses: 

The team evaluation of Calc Note WII-94-017-0, 

(a) The Beff uncertainty at the lower band was applied incorrectly. 

METCOM procedure 6.7, "Beta-effective and Prompt Neutron Lifetime," paragraph 

6.7-1, clearly required that rinimum Beff is calculated by mu1tiplyir.g the 

best-estimate value by [Deleted pursuant to 10 CFR 2.790 - Document describes 

- 13 - 

155 
I 

a specific value]. However, the Calc Not6 

[Deleted pursuant to 10 CFR 2.790 - Docm 

thereby obtaining a slightly larger minin Beff. Although this methodology 

was judged by the team to be nonconserv 

significant because the nonconservatisa 

compared to the avai 1 ab1 e margin . 

(b) The verifier noted on the che ,t that, contrary to the 

requirements of METCOM procedure 6.7, t 

was not applied. However, the team detr .d that the multiplier had been 

applied but that the verifier had not clc 

defect with the author of the Calc Note b4 

ivided the best-estimate value by 

describes a specific value], 

, the nonconservatism was not 

he error (0.25%) was small 

I'actor mu1 tip1 ier of 0.97 on Beff 

out this mistakenly observed 

rather left the defect unresolved. 

(c) The total number of pages were not noted and the last two pages of 

the Calc Note were not numbered, including the checklist page. 

EP-302, these provisions were intended to ensure that the Calc Note is 

complete and auditable. 

As required by 

The team discussed these weaknesses with the author and verifier, and again 

later with the author, verifier, and manager. 

to these weaknesses by correcting the Calc Note and reissuing it as 

Revision 1. 

concerns. 

CNFD Core Engineering responded 

The CNFD actions taken during the inspection satisfied the team's 

(2) Doppler Effect 

The team evaluation of the Point Beach Cycle 21 inputs into the rod ejection 

accident evaluation led to investigating the Doppler effect, which is negative 

reactivity insertion due to an increase in neutron absorption by fuel. This 

reactivity effect occurs when the fuel temperature is elevated, thus 

increasing the absorption cross-section of fuel. The Doppler effect is the 

primary core physics inherently negative fast feedback to power changes and 

plays a large role in fast reactivity transients. CNFD methodology emphasized 

the Doppler defect representation as opposed to the Doppler coefficient. 

METCOM and other documents such as topical report WCAP-7308-L-P-A, "Evaluation 

of Nuclear Hot Ciicrnnel Factor Uncertainties," specified uncertainties rin 

safety-re1 ated parameters. An internal memo, WIN 249-5142, "Recomnended 

Design Limits for Reactivity Coefficients for Safety Analysis," provided for 

an uncertainty to be applied to the Doppler effects. 

procedure 6.6, "Doppler Coefficient and Defect," Revision 0, dated February 

1994, did not provide a Doppler defect uncertainty factor. 

The best estimate value of the Doppler defect for this reload core design, if 

adjusted in the conservative direction by [Deleted pursuant to 10 CFR 2.790 - Document describes a specific value] in accordance with the value in WIN 249- 

5142, would be 3% from the bounding analysis value for the rod ejection 

reactivity insertion accident. This was not significant for the Point Beach 

Cycle 21 reload core design since margin (3%) would have remained and the team 

considered [Deleted pursuant to 10 CFR 2.790 - Document describes a specific 

value] uncertainty to be conservative. However, it was postulated by the team 

that future reload core designs could have smaller Doppler defects which, if 

adjusted by an uncertainty factor, would not meet the bounding analysis value. 

The 

However, METCOM 

- 14 - 

156 
The team determined that this may not be apparent if an uncertainty factor was 

not applied since it may not be clear that the Doppler defect va:ue provided 

by CNFD to NTD was a best estimate and not a value including uncertainty. 

The team discussed this weakness with CNFD. 

pursuant to 10 CFR 2.790 - Document doscrlbes a specific value] uncertainty 

factor value in YIN 249-5142 was larger than currently warranted. 

stated that METCOM procedure 6.6 would be modified to make designers aware 

that Doppler defects calculated by the ANC code were best estimate values 

without uncertainty factors incorporated. For future evaluations, the 

designer will factor this into ewluations of margin for Doppler affected 

transients. 

CNFD actions taken during the inspection satisfied the team's concerns. 

It was agreed that the [Deleted 

CNFD also 

CNFD assigned METCOM work item 95-06 to address this issue. The 

3.3.1.5 Zion Station Unit 1 Cycle 14 

The fuel projects engineer was the 8 interface for the Comnonwealth Edison 

Company (CEC), Zion Station Unit 1 (Zion) Cycle 14 reload core design and 

reload safety analysis and provided the Contract and Technical Data ICATD), 

RSER, and the RSLC to Core Engineering for the core design analyses. The 

Core Engineering reload core design activities were performed using 

methodologies that were documented in (a) the METCOM, (b) the Software 

Engineering Methodology manual, and (c) the Engineering Services Manuals, as 

required by procedure EP-105. 

carried out following the METCOM procedures found in Volumes 1-4 and 

documented in the Calc Notes Report prepared by the core designer. 

the review of the Zion Cycle 14 reload core design process, the METCOM 

procedures, and their application, as documented in the Zion Cycle 14 Calc 

Notes, were reviewed in detail. 

The Zion Cycle 14 reload core design was 

As part of 

Westinghouse CNFD had established a special F-configuration classification for 

computer codes that received sufficient verification and validation to be used 

in the reload core design analyses. 

Cycle 14 reload core design analysis were F-configured. 

close interaction and good communication between the computer software 

developers and the Core Engineering code users. 

identification and timely correction of the erroneous xenon (Xe) yield data in 

the PHOENIX-P library. 

PHOENIX-P data were appropriately evaluated and incorporated in the Zion 

All computer codes used in the Zion 

There appeared to be 

This was il1tlctrated in the 

The effect of this modification on the F-configured 

' The Zion Cycle 14 reload core design was initially a single-scope 

design with CNFD providing the complete reload core design and reload safety 

analysis. However, after the initial CNFD reload core design was completed, 

the licensee, CEC, requested a split-scope and assumed responsibility for the 

reload core design only. 

addresses the initial CNFO reload core desigln. After the CNFD initial reload 

core design was completed, CEC performed the! reload core design analysis with 

the W PHOENIX-P code and the AN6 code system (installed on a computer system 

similar to that used by CNFD), with the METCOM manuals, and with training 

provided by M. For several cycles, CEC has also provided the cycle-specific 

startup data and constants for the INCORE code. 

The team evaluatioln described in this report 

- 15 - 

157 
Cycle 14 reload core design analysis (RSAC Calc Note, CWB-93-013-0, 

"Evaluation of the Impact of the Xe Yield Error on the Zion 1 Cycle 14 

Models") . 

The Zion Cycle 14 reload core design analysis, documented in the four volumes 

of Calc Notes, was reviewed in detail. The Zion Cycle 14 calculations of the 

reload core design parameters input to the RSAC (e-g, reactivity coefficients, 

power peaking factors, RCCA bank worths) were also reviewed. 

were carried out and documented in a manner consistent with the METCOn 

calculational procedures and requirements. Each Calc Note was verified by an 

independent reviewer and all comments satisfactorily resolved, as required by 

METCOF( procedure 1.9. 

These analyses 

Vessel F1 uence Reduction 

CEC's core design specification, documented in letter ZIC141003, "Final Energy 

Specification for Zion 1 Cycle 14," dated September 22, 1992, required that 

the reload core design employ the strategy of the LQP (low leakage loading 

pattern) plus Hf flux suppressor rod flux reduction as the vessel fluence 

reduction option for Zion Cycle 14. The average fuel assembly power for the 

peripheral fuel assemblies should approximate, or be bounded by, the h4P 

target assembly powers. The review of the Zion Cycle 14 Calc Notes found no 

documentation of the calculation made to demonstrate that the proposed 

Cycle 14 reload core loading provided the required vessel fluence reduction. 

The lack of documentation confirming the ability of the Zion Cycle 14 reload 

core design loading pattern to provide the required vessel fluence reduction 

is considered a weakness in the Calc Note system. 

When this omission was brought to the attention of the CNFD staff, a new Calc 

Note, CW6-95-001-0, dated February 10, 1995, was developed which demonstrated 

that the Cycle 14 core loading pattern provided the vessel fluence reduction 

required by the CEC core design specification. The CNFD actions taken during 

the inspection satisfied the team's concerns. 

Several Zion Unit 1 operational issues were discussed during meetings with the 

CNFD Core Engineering staff. A relatively small core power t~rc i: =P% at 

HFP) had been observed at Zion Unit 1. 

were suggested, including (a) steam generator tube plugging, (b) pump 

maintenance, or (c) asymmetric fuel shuffle. 

Several possible causes of the tilt 

During the startup tests for Zion Cycle 13, a quadrant power tilt was observed 

and ultimately traced to the failure to load the Hf flux suppression rods in 

the selected peripheral fuel assemblies. The correct core loading pattern, 

including the location of the IFBA and the Hf flux suppression rods, was 

transmitted by W CNFD to CEC in Figures 2 and 9, respectively, of the Fuel 

Loading Pattern Letter, 93 CW-G-0030, "Commonwealth Edison Company Zion 

Nuclear Power Plant, Zion 1 Cycle 14 Burnable Absorber Requirements and 

Candidate Loading Pattern," dated March 1, 1993. However, the locations of 

the Hf rod inserts were apparently not adequately communicated to the Zion 

Unit 1 site personnel responsible for the core loading. 

core loading pattern in Figure 9 and concluded that the indication of the Hf 

flux suppression rods was adequate. Nevertheless, CNFD stated that the core 

The team revSewed the 

- 16 - 

158 
loading maps are produced by the ANCHOR code (used to aid in verifying ANC 

loading patterns) and, that to improve the CNFD - CEC interface, ANCHOR had 

been modified to simplify and improve the presentation of the data on the core 

loading maps. 

indicated by this corrective action was d strength in the CNFD Core 

Engineering reload process. 

The team concluded that tbe flexibility and responsiveness 

(2) Thermal-Hydraul ic Analysis 

The Zion Cycle 14 T/H analysis was performed by the CNFD Core Engineering Fuel 

Analysis group and documented in Calc Note T/H-93-012-0. 

included the determination of fuel rod densification and temperatures, DNBR 

limits, 3nd axial offset 1 imits. Analyses of the loss-of-flow, locked-rotor, 

rod misalignment and steam1 ine break transients are also included. 

analyses used the THINC-IV (thermal-hydraul ic interaction, analysis, code) and 

PAD (fuel rod performance code). 

bases, and assumptions for these analyses were given in the Thermal-Hydraulics 

Design procedure manual. 

these procedures. 

that reload core design information required for the steamline break T/H 

analysis was obtained in a telephone conversation with CEC. No followup 

documentation of this communication and data transmittal was available. 

This undocumented transmittal of design input data did not allow independent 

verification and was not consistent with the T/H design procedures. 

considered this instance a weakness in the Calc Note system. 

The analysis 

These 

The procedures documenting the methods, 

The Zion Cycle 14 T/H analyses generally followed 

However, in reviewing the T/H Calc Note, the team noted 

The team 

(3) LOCA and Non-LOCA Transient and Safety Analyses 

The reload safety analyses were performed by the NTD. 

analyses were performed by the Transient Analysis group and were controlled by 

a set of Safety Analysis Standards. The LOCA analyses were performed by the 

Safeguards Engineering group and were controlled by a set of Safeguards 

Engineering Standards. The reload transient analyses were performed using a 

bounding analysis approach in which the cycle-specific input parameters, with 

a significant effect on the transient were compiled in the RSAC and compared 

to the parameters for the precalculated reference or bounding analysis. 

the input parameters were not conservative relative to the reference analysis, 

a cycle-specific transient evaluation was required. The RSAC was initiated by 

the Transient Analysis group and the cycle-specific parameters were 

contributed by the responsible engineering groups. Zion Cycle 14 was a splitscope reload, and the licensee, CEC, provided the RSAC parameters for reload 

core design neutronics in letter ZlC14/016, "Neutronics Only SPIL Transmittal - Zion Unit 1 Cycle 14: Revised," dated October 7, 1993. 

The non-LOCA transient 

If 

'It was noteworthy that letter ZlC14/0:16, "Neutronics Only SPIL 

Transmittal - Zion Unit 1 Cycle 14: Revised," dated October 7, 1993, provided 

information concerning the CEC reload core redesign relative to the current 

Zion Unit 1 reference analysis limits; however, the Zion Cycle 14 T/H analysis 

was based on a comparison of t5e CEC reload core redesign to the original CNFD 

Zion Cycle 1" analysis. 

- 17 - 

159 
As part of the reload procedures evaluation, the Safety Analysis Standards 

were reviewed with representatives of the Transient Analysis group. 

l i s t i n g o f transient parameters used fcir the Zion Cycle 14 reload was found to 

be in agreement with the approved listing in WCAP-9272-P-A, "Westinghouse 

Reload Safety Evaluation Methodology," dated July 1985. 

The RSAC 

To evaluate the safety analyses performed when the reference analysis was not 

bounding, the team reviewed the Zion Cycle 14 rod ejection accident (REA) 

analysis performed for the VANTAGE4 upgrade. 

CN-TA-90-282 documented the methods and assumptions of this analysis. 

calculations were performed with F-configured versions of the TWINKLE (core 

transient) code and the FACTRAN (fuel rod performance) code. The stand-alone 

calculations of the REA rod motion and doppler weighing factor were reviewed 

and found to be correct. The TWINKLE moderator temperature coefficient (MTC) 

input was determined by the ANC code, and the fuel rod heat flux and 

temperatures were determined by the PAD code. The team concluded that the 

Zion Cycle 14 REA analysis was carried out in a manner consistent with the 

procedures given in Safety Analysis Standard 14, "Rupture of a Control Rod 

Drive Mechanism." 

The REA analysis in Calc Note 

The 

(4) Fuel Rod Design Analysis 

The Zion Cycle 14 fuel rod design was performed by the Fuel Analysis group. 

The analysis included an evaluation of centerline temperature, internal 

pressure, clad strain, oxidation, corrosion, clad flattening, swelling, and 

gap conductance. The P/PD&D group in CNFD COLA provided input to these 

analyses, including the DEVL and the Bill of Materials/Key Sheet (BOM/KS). 

The Fuel Analysis group provided the rod back-fill pressure to the CNFD COLA. 

The methods and criteria used in these analyses were documented in the Fuel 

Rod Design Procedure Manual. The analysis was performed with F-configured 

versions of the PAD code and various stand-alone versions of the PAD modules. 

The Zion Cycle 14 fuel rod analysis documented in Calc Note T/H-93-067-0 and 

the three subsequent revisions were reviewed in detail. Revision 1 of T/H-93- 

067-0 used a cycle-specific fluence versus burnup correlation +P Astermine rod 

growth; Revision 2 incorporated the final CEC spl it-scope reload core dasign; 

and Revision 3 incorporated the revised Region-14A fluences (resulting from 

the Hf rod inserts) in the rod growth calculation. In Revision 1 to T/H-93- 

067-0, the growth calculation required a special-purpose computer calculation, 

and the team confirmed that this calculation was verified by an independent 

hand-calculation. The fuel desisn analyses satisfied all the performance 

criteria up to the Zion Cycle 14 fuel burnup 'limit. 

carried out, documented, and verified in a manner consistent with the methods 

and requirements of the Fuel Rod Design Procedural Manual. 

These analyses were 

3.3.1.6 Sal Wuclear Generating Station Unit 2 Cycle 9 

A designated project engineer acted as the contractual and technical interface 

between the licensee, Public Service Electric & Gas Company (PSE&GC), and the 

CNFD for the Salem Nuclear Generating Station Unit 2 (Salem) Cycle 9. 

team noted that the reload design process begins as an iterative process in 

which the licensee, CNFD, NTD, and the CNFD COLA participate. The RSLC plays 

The 

- 18 - 

160 
a key role in the interactions of these groups. 

design process, the RSAC played a key role in ensuring that the Core 

Engineering group and the NTD NSA group together produced a reference safety 

analysis that is valid for the reload cwe being designed. 

Throughout the reload core 

The team reviewed the interfaces and documentation of the reload analysis 

process by inspecting the key interfacing documents (i .e., the RSER, the RSLC, 

the Enrichment Requirement Letter, and the RSAC) and their transmittal letters 

to gauge the adequacy of the process. 

the external and internal interfaces governing the reload core design and 

reload safety analysis process, the technical directions provided for the 

process, the responsibilities assumed by the individual engineers in the 

execution of their tasks, and the engineers responsiveness to licensee inputs 

and requirements were adequate. 

The team judged that for this reload 

Furthermore, the team noted that in the instance described below, the strength 

of the CNFD organization was demonstrated through the effective interactions 

between members of CNFD and NTD under conditions of considerable time 

constraint. After failed fuel was detected during ultrasound inspection of 

fuel to be reinserted into the Salem Cycle 9 reload core design, the Cycle 9 

reload core was redesigned. 

Unit 2 Cycle 9 Redesign," CDB-94-253, FA-94-294, dated November 11, 1994, the 

redesign included replacing all region 8 fuel assemblies with a history of 

baffle placement. The redesign also necessitated a re-evaluation of the LOCA 

analysi s (Cal c Note SEC-SA1 I-4570-C2 , "RSAC-PNJ-Cycl e 9 Re1 oad Eva1 uat ion 

redesign," Revision 0, dated November 14, 1994). The redesign and the 

associated analyses were nonroutine activities. 

evaluation, the team determined that these activities were carried out in a 

thorough manner meeting existing procedural requirements. These activities 

also reflected effective interfacing between members of CNFD and NTD to 

successfully execute a difficult task under considerable time pressure. 

As documented in RSAC, "Evaluation for Salem 

On the basis of its 

(1) AsyRmetrlc RCCA Withdrawal 

On May 27, 1993, at Salem Unit 2, a single failure in the rod control system 

caused a single rod to withdraw 15 steps from the core while an insert signal 

was being applied. On June 17, 1993, PSE&GC requested an emergency license 

amendment involving the rod control system at Salem Units 1 and 2. The 

emersency 1 icense amendment request noted that a potenti a1 single failure 

could cause a single (or multiple asymnetric) RCCA withdrawal, and that 

explicit analyses determined the single RCCA withdrawal at power event to be 

bounded by a multiple RCCA withdrawal of two adjacent D-Bank RCCAs. On 

June 21, 1993, the NRC issued Generic Letter 93-04, "Rod Control System 

Failure and Withdrawal of Rod Control Cluster Assemblies, 10 CFR 50.54." In 

response to this issue, W CNFD published WCAP-13803-AY "Generic Assessment of 

Rod Cluster Control Assembly Withdrawal," Revision 1 (original version dated 

August 1993; approved version dated November 1994). 

- 19 - 

161 
Westinghouse CNFD had previously considered the occurrence of a mu1 tiple RCCA 

withdrawal event due to a single failure to be incredible. Consequently, the 

METCOH design manual did not analyze this event. The Calc Notes documenting 

the analysis of this event, therefore, identified the method used as a 

nonstandard method since it is not described in the design manual. The use of 

nonstandard methods was governed, in part, by CNFD procedure EP-302. 

sections of procedure EP-302 were relevant to the documentation of nonstandard 

calculations: 

TWO 

(a) Section 7.1.11.5: document assumptions, identified those 

assumptions that must be verified as design proceeds. 

(b) Section 7.1.11.8: document deviation from standard methods 

(defined in a design manual) or the use of nonstandard methods (not defined in 

a design manual) in sufficient detail to support verification. 

Calc Note PSK-93-001-0 (creation date June 5, 1993) documented the accident 

analysis for single and multiple rod withdrawal accidents. 

determined that Calc Note PSY-93-001-0 did not conform to Sections 7.1.11.5 

and 7.1.11.8 of procedure EP-302 since the assumptions regarding bounding 

asymmetric rod configuration had not been documented and verified and since 

the use of nonstandard methods had not been documented in sufficient detail to 

support verification. Since the Calc Note provided the bases for the 

emergency license amendment request, conformance with procedure EP-302 was 

necessary to ensure that these cal cul at ions meet the appropri ate CNFD 

technical and quality standards. The team concluded that CNFD failure to 

document the assumptions and the deviations from standard methods did not 

comply with the provisions of EP-302. As a result, a potential nonconformance 

was identified during this part of the inspection. 

In response to the team determination that Calc Note PSK-93-001-0 did not 

conform to procedure EP-302, the following corrective actions were taken by 

Core Engineering while the team inspected CNFD/Westinghouse Energy Center: 

(a) to clarify the methodology used in analyzing the asymmetric rod withdrawal 

accident, 

However, the team 

the coauthor modified a page of the Calc Note and added three pages 

(b) a clarification of the methodology referred to WCAP-13803-A, 

Revision 1, which provided the analyses needed to identify the bounding 

asymnetric control rod configuration; and 

(c) the additional material provided bore the signatures of the 

coauthor and the verifier. 

As a result of the additional material being incorporated into Calc Note PSK93-001-0 and the corrective actions taken by Core Engineering, the team 

determined that its concern regarding compliance with procedure EP-302 had 

been satisfied and that the potential nonconformance was closed. 

- 20 - 

I 162 
(2) LOCA and Non-LOCA Transient and Safety Analyses 

The team assessed transient, safety, and set point analysis processes by 

examining of the standards utilized in these analyses, by verifying that 

NRC-approved computer codes were used in these analyses in conformance with 

SER-specified limitations and restrictions, and by examining individual Calc 

Notes pertaining to transient and safety analyses. 

CNFD currently utilized the LOFTRAM code for systems transient analyses, the 

FRACTRAN code for fuel rod heat transfer calculations, the THlNC code for DNBR 

analyses, the TWINKLE code for fast transient analyses, and the OPTOAX code 

for set poifit analyses. The WFLASH and NOTRUMP codes were used for smallbreak LOCA analyses; and the LOCTA, SATAN, COCO, WREFLOOD, BART, and BASH 

suite of codes were used for large-break LOCA analyses. 

In the reload safety evaluation process, the values of the key safety 

parameters for the reload core were determined by CNFD and transmitted to NTD 

via the RSAC. Westinghouse NTD determined whether all "reload values" of the 

key safety parameters were bounded by the current limits. 

violations were resolved through redesign (interactively with CNFD) or through 

reanalysis of the affected transients. 

conservatism inherent in the reference analyseq, most reload cores have 

parameter values that are bounded by the current limits. 

they did occur, were usually resolved through a relatively minor redesign of 

the core. Reanalysis of a set of transients was rarely necessary. The LOCA 

analyses for the reference core tend to be bounding for reload cores unless a 

new fuel design was introduced, in which caise the LOCA analyses were redone. 

As a result, while CNFD routinely performs between 30 and 40 core design 

analyses a year, complete sets of transient. and safety analyses were much 

rarer. 

and safety analysis area compared to core engineering. 

If they were not, 

The team noted that owing to the 

Violations, when 

This Pact was reflected in the much smaller staffing in the transient 

Based on an examination of the appropriate standards, computer code 

documentation, topical reports on recent transient and safety analyses, and 

selected Calc Notes, and based on discussions with the engineers involved, the 

team determined that activities in the safety and transient analyses erea were 

excellently performed. 

3.3.1.7 South Texas Project Unit 2 Cycle 4 

The evaluation of the CNFD design process for the Houston Lighting and Power 

Company (HL&P) South Texas Project Unit 2 (STP) Cycle 4 reload core design and 

reload safety analysis started with the examination of the contractual and 

schedular requirements. 

licensing engineer. 

define the scope and length of the project., The minutes of the Design 

Initialization meeting on July 15, 1992, were reviewed for design constraints. 

The minutes of the Production Initialization meeting on October 30, 1992, were 

also reviewed for consistency with the interface documents (DEW) and the 

original manufacturing schedule received from CNFD COLA. STP Cycle 3 was 

originally scheduled to shut down on February 28, 1993, with SIP Cycle 4 to 

start up on April 25, 1993. Due to a lengthy outage, the actual STP Cycle 4 

The overall process was out1 ined by the fuel 

The project engineer correspondence file was examined to 

- 21 - 

163 
start up (Mode 2) was on April 24, 1994. 

underwent several redesigns from July 1992 to February 1994 to respond to 

changing licensee conditions (i .e. , reduced energy requirement, fuel vibration 

problems, and rotated grid issue). A STP mid-Cycle 4 redesign effort was also 

reviewed, regarding a STP plant-specific evaluation which was submitted for 

NRC review in April 1993 and approved for incorporation during Cycle 4. 

Setpoint changes for the midcycle redesign were discussed in a meeting with 

HL&P on September 14, 1994, in which the RIQ was revised, and the RSE was 

scheduled for December 1994 to allow midcycle incorporation by April 7, 1995. 

The STP Cycle 4 reload subsequently 

The detailed STP Cycle 4 inspection began with a review of the deliverables to 

the licensee: the final nuclear design report (NDR), RSE, and COLR documents. 

These were evaluated to determine the key reload design parameters, which were 

then followed backwards through the design process to the initiating contract 

and 1 icensee requirement documents. 

review and approval of these documents were also examined. 

The process and documentation for the 

(1) LOCA and NOR-LOCA Transient and Safety Analyses 

Section 3.2 of the RSE documents the Accident Evaluation analyses. 

reviewed through discussions with the responsible NTD Safeguards Engineering 

group and Transient Analysis group engineers. 

Calc Notes for the STP Cycle 4 reload LOCA analysis were reviewed relative to 

the Safeguards Engineering Standards. The referenced standards were RSAC-01 

"Overview of the Reload Process," Revision 3, dated June 13, 1994, and 

RSAC-02, "Reload Safety Analysis Checklist (RSAC) Parameters and their 

Review," Revision 4, dated March 18, 1994. 

and customer interfaces with NTD, were examined for appropriate documentation. 

RSAC-02 defined the key LQCA-related parameters and listed the code models 

used. 

It was 

The analyses process and the 

RSAC-01, which defined the CNFD 

The non-LOCA analyses were reviewed relative to the Safety Analysis Standards 

(SAS). The application of SAS-17, "RSAC Preparation and Evaluation," 

Revision 5, dated Harch 21, 1994, to specify the allowed reload key safety 

parameter ranges which define the current limits of RSAC was examined fw 

consistency with WCAP-9272-P-A. Other standards referenced in the Cak Notes 

were examined. 

(2) Thermal-Hydraul le Analysis 

Section 2.3 of the RSE, "Thermal and Hydraulic Design," was evaluated by 

discussions with the responsible fuel analysis engineer. 

Notes (T/H-92-179) were reviewed relative to the Thermal Hydraulic Design 

Procedure Manual. Previous Calc Motes (T/H-91-124 and T/H-92-084) were also 

referenced as unc anged for the analyses of record. 

meeting minutes, including the RIQ, as the primaery input interface document 

for the T/H design. 

January 15, 1993, documented the reload redesign. The fuel rod design effort 

involved review of the mechanical design against the RSAC limit list to 

confirm applicability for the DYB events or to determine re-analysis, as for 

the steamline break. 

The STP Cycle 4 Calc 

The Design Initialization 

Fuel Design Data List (FUDDL) memo CDC-93-014, dated 

- 22 - 

164 
(3) Wucl ear Design 

Section 2.2 of the RSE, "Nuclear Design," was reviewed through discussions 

with the responsible CNFD Core Engineering Core Design engineer. 

became partially a spl it-scope reload cm! ae:, ign, with the 1 icensee 

participating in the nuclear design to reduce the number of feed bundles 

required. The primary interface documents were reviewed to determine the 

control of the split responsibility and the flow of information to and from 

the licensee. Calc Notes supporting the key parameters for the NDR, the RSE 

and the COLR were reviewed. 

reload were reviewed and representative interface documents were examined and 

found satisfactory. The eight volumes of Calc Notes were reviewed for 

consistency and completerress with respect to the significant redesign, the 

extended schedule, and personnel turnover. The final reload core design was 

reviewed in detail with the currently responsible Core Engineering Core Design 

engineer, including the Cycle 4 final design model sumnary/checklist from 

METCOM Table 1.7-9. The overall flow of information between CNFD and HL&P was 

reviewed with the project engineer. 

(4) fuel Hechanical Design 

STP Cycle 4 

The Technical Specification changes for the 

Section 2.1 of the RSE, "Mechanical Design,," was evaluated by discuss ons with 

the fuel analysis engineer and by the review of the Calc Note relative to the 

Fuel Rod Design Procedure Manual, Revision 4, dated April 1993. STP used the 

VANTAGE-5 fuel assembly design referenced iin WCAP-10444-P-AY with a 14-foot 

active fuel length. The 36 Cycle 4 Region 6 reload fuel assemblies specified 

in the final core design incorporated the following features differing from 

the previous Region 5 reload: 

low-pressure drop (LPD) Zircaloy mid-grids, 

I FBAs , modified top grid assembly, and 

modified top nozzle assembly. 

The 36 Region 7 fuel assemblies comprising the remainder of the Cycle 4 reload 

also incorporated the following additional features: 

extended burnup bottom grid, 

fuel rod repositioning, 

keyless top nozzle assembly, and 

rotated a1 ternate mi xi ng virne LPD mitl-grids. 

As was observed for the Sumner Cycle 9 reload, no single document existed 

which showed that all fuel rod design criteria, as specified in the Fuel Rod 

Design Procedure Manual, were satisfied for the core design. 

- 23 - 

165 
(5) Technical Specifications 

Section 4.0 of the RSE references the STP Technical Specification changes 

required for Cycle 4 operation. The Technical Specification changes involved 

the implementation of increased boron concentrations in the refuel ing water 

storage tank and the safety injection accumulator, and it was verified that 

these changes were accounted for in the reload design process. 

(6) Core Operating Limits Report 

Section 5.0 of the RSE references the STP COLR for Cycle 3 and the updated 

Cycle 4 COLR that was delivered to the licensee along with the RSE. 

verified that the values of MTC, control rod insertion limits, the peaking 

factors (F, and FAH), and the allowable axial flux difference as listed 

corresponded to those determined in the reload design process. 

It was 

3.3.2 Fuel Assembly Hechanical Design 

The fuel assembly mechanical design functions were performed by the 

Product/Process Development 81 Design (P/PD&D) group of CNFD, located at CNFD 

COLA. 

inputs to and interface with the reload core design and reload safety analysis 

process during this portion of the inspection at CNFD/Westinghouse Energy 

Center. Therefore, in the interest of both overall readability and 

convenience, this portion of the inspection report described the team 

evaluation of the P/PD&D fuel assembly mechanical design functions that were 

performed at both the Westinghouse Energy Center and CNFD COLA and evaluated 

during both inspection periods, as described in Section 3.2 above. 

Although P/PD&D was located at CNFD COLA, the team evaluated P/PD&D 

Westinghouse CNFD provided a wide range of fuel designs, including the 

- W 14x14, 15x15, 16x16, and 17x17 lattice fuel assembly arrays and fuel 

assembly designs for nuclear power plants designed by both ABB Combustion 

Engineering and Babcock & Wilcox. The primary fuel mechanical design 

responsibilities of P/PD&D included (a) new hardware product development and 

testing, (b) process development to support fabrication of new hardware 

product designs, (c) reload-by-reload specification of the hdrawe product 

design for manufacturing to meet licensee-specific design requirements, 

(d) design support for manufacturing, and (e) collection, evaluation, and 

dissemination of product performance data. Both the Fuel Performance 

Technology and the Product Performance groups 3f P/PD&D were located at the 

Westinghouse Energy Center, and tne following P/PD&D groups were located at 

CNFD COLA: (a) Product Development 81 Testing, (b) Thermal-Hydraulic Testing 

Analysis, (c) Materials & Mechanical Process Development, (d) Design 

Specification & Drafting, and (e) Product Design. 

3.3.2.1 Mechanical Design Process 

The Fuel Projects Organization provided the Job Order and CATD document, which 

included licensee fuel and component design and operating data. In a typical 

reload core design, the P/PD&D Design Specification and Drafting group 

received design and fabrication data from Core Engineering and produced the 

BOM/KS, which compiled the fabrication specifications for the fuel assemblies 

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166 
and core components. The reload-specific fuel and component designs were 

selected by the Design Specification and Drafting group from the W Current 

Design List (CRDL), which was a compilation of available compatible fuelre1 ated core components (e. g . , pyrex gl ass burnabl e absorbers , WABAs , hol ddown 

assemblieb, and thimble plugs) and des ,gn features (for Westinghouse VANTAGE5, VANTAGE-SH, and VANTAGE+ fuel assembl ies, these design features include 

axial and radial blankets, IFBAs, intermediate flow mixer (JFM) grids, lowpressure drop (LPD) zircaloy grids, removable top nozzles (RTNs), debrisfilter bottom nozzles (DFBNs), certain assembly modifications, and ZIRLO" fuel 

clad tubing). 

The Core Engineering inputs to the BOM/KS included the (a) fuel and IFBA 

enrichment, (b) pellet stack length and IFHA pattern, (c) core loading plan, 

and (d) the fuel rod pressure. The P/PD&D Product Design group evaluated 

design deviations and special designs. The P/PD&D Product Development & 

Testing group evaluated major redesigns and new fuel designs and did special 

mechanical testing (e.g., grid compressibility, clad burst, and vibration 

tests). These design evaluations were led by a project manager; the size and 

composition of the design team were determined by the scope of the design. 

Recent examples of designs evaluated includle the Maine Yankee Atomic Power 

Company, Maine Yankee Atomic Power Station (Maine Yankee) Cycle 15 reload core 

design (ABB Combustion Engineering designed1 pl3nt: , Hf flux suppression rod 

designs, and fuel assembly vi bration compensatory designs. These design 

evaluations may require qualification of special processes (e.g., debrisresistant coating) and T/H testing. The P/PD&D Thermal-Hydraulic Testing 

group performed pressure drop tests and evaluated DNB test data. 

multirod-array DNB tests were performed at Columbia University and the sideby-side tests were performed in Canada. The results of this process included 

the DEVL, prepared by CNFD P/PD&D and the FPC, as described in Section 3.3.1 

of this report, which were provided to CNFDi Core Engineering. The CNFD Core 

Engineering Fuel Licensing Integration group summarizes any mechanical design 

changes identified in the DEVL and FPC and gave that data to NTD. 

The 

To evaluate the CNFD fuel mechanical design process, the team selected the 

following reload design packages. 

(1) Zion Station Unit 1 Cycle 14 

The Zion Cycle 14 reload fuel consisted of 44 assf .,blies at 3.6 w/o and 

32 assemblies at 3.4 w/o VANTAGE-5 fuel assemblies without IFM grids. 

fuel rod analysis for the reload core design employed NRC-approved methods . The3 

%einer, R.A., et a1 . , WCAP-10851-P-A, "Improved Fuel Performance Models 

for Westinghouse Fuel Rod Desiqn and Safety Evaluations, " dated August 1988; 

and Davidson, q.L., (Ed.), el al. , WCAP-10125-P-A, "Extended Burnup Evaluation 

of Westinghousc Fuel," dated December 1985. 

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167 
The mechanical design activities for the Zion Cycle 14 reload fuel were 

controlled by a detailed design specification contract process. The specific 

design tasks were initiated by Job Orders, and the schedule was determined by 

the Project and Design Milestone Schedule System (PDMS). 

were based on the information provided in the CATD for Zion Unit 1, dated 

March 2, 1992, which was provided by the project engineer. The available and 

compatible VANTAGE4 fuel assembly design features, part numbers, and build 

instructions were selected from the CRDL 15A, "Current Design list 15x15 Fuel 

Rod," Revision 9, and PELS100, "Assembly, Pellet, Stack, Fuel ," Revision 14. 

A Production Initialization meeting (analogous to the Design Initialization 

meeting held by Core Engineering) was held prior to the release of the DEVL. 

The Zion Cycle 14 fuel rod pressure was provided in the "Generic Backfill 

Pressure Update," dated February 2, 1993. 

The initial BOM/KS 

The core loading pattern and burnable absorber requirements were provided by 

Core Engineering in "Zion 1 Cycle 14 Burnable Absorber Requirements and 

Candidate Loading Pattern," dated March 1, 1993. 

include the axial blanket stack height, but this data was provided later in 

"CWBQ Axial Blanket Length," letter dated March 19, 1993. This omission was 

corrected in future reload core loading pattern transmittals, as part of the 

corrective actions for the Florida Power and Light Company, Turkey Point Units 

3 and 4, WABA axial misslocation event described in Section 3.3.2.2 below. 

The shipping requirements for the WABA core components was given in the 

Component Loading Chart, dated June 24, 1993. The final BOM/KS, giving the 

Zion Cycle 14 part numbers for fabrication, were then issued for the fuel 

assemblies and core components. 

in conformance with procedure. 

This transmittal did not 

These documents were reviewed and found to be 

The changes to the fuel assembly design included a change to a controlled fuel 

rod gap and changes to the top nozzle engraving. The controlled fuel rod gap 

was described in Engineering Change Notice 26754, dated June 22, 1993, and the 

engraving was described in the Waiver Request T93-021-01, dated May 25, 1993. 

The treatment of these design changes was in conformance with procedures. The 

final Zion Cycle 14 reload safety evaluation was transmitted to the licensee 

(CEC) in a letter dated July 15, 1994. 

(2) Uaine Yankee Atomic Power Station Cycle 15 

The Maine Yankee Cycle 15 reload core design and fuel assemblies were being 

provided by CNFD. Since this is the first nuclear fuel designed and 

manufactured by 

fuel assembl ies required a series of reload-specific analyses. The required 

analyses were simplified by the fact that the licensee, Maine Yankee Atomic 

Power Company, performed the nuclear analysis, most of the T/H analysis, and 

the reload safety analysis. 

CNFD for Maine Yankee, this Cycle 15 reload core design and 

A member of the Project Manager group acted as the project engineer for the 

Maine Yankee Cycle 15 reload core design, and the P/PD&D Product Development 

and Testing group performed the mechanical design. The cycle-specific fuel 

data requirements were included in "Maine Yankee Purchasers' Fuel Data 

Requirements for Licensing," dated November 29, 1994, and "Maine Yankee (MYCQ) 

Fuel Design Drawing List - Revision-01," dated March 31, 1994. The P/PD&D 

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168 
Product Development and Testing group design act i vi ties i ncl uded (a) component 

procurement, (a) vibration testing, (c) bulge testing of thimble tubes to the 

grid sleeves, (d) fabrication drag tests of the fuel rods through the grids, 

(e) top nozzle-to-fuel handling device, (f) grid-crushing tests, and (9) RTN 

joint testing. 

The test evaluation reports' were evaluated by the team and found to be 

adequately documented and verified, in conformance with procedures. The Maine 

Yankee final design review, documented in PDT-94-098, "Maine Yankee Fuel 

Design Assembly Final Design Review Package," dated May 5, 1994, and the 

acceptance of the resolution of the four action items by the designated 

reviewers were also evaluated by the team and found to be in conformance with 

procedure. 

3.3.2.2 Mechanical Design Issues 

During the inspection of the CNFD fuel mechanical design process, the CNFD 

staff briefed the team on certain mechanical design issues and events that 

were of special interest to the team. Of these issues discussed, the team 

selected the following reload fuel mechanical design issues to evaluate the 

P/PD&D response to and evaluation of the issues. 

(1) Fuel Assembly Vibration 

In April 1993, fuel rod failures resulting from grid-to-rod fretting (GRF) 

were observed during the Salem Unit 2 Cycle 7 outage and the Duquesne Light 

Company, Beaver Valley Power Station Unit 1 (Beaver Valley) Cycle 9 refueling 

outage. The GRF occurred between the third and sixth grids (mid-grids) and in 

most cases was found in fuel assemblies that had spent at least one operating 

cycle adjacent to the reactor core baffle. Westinghouse CNFD had not 

experienced prior mid-grid failures; however, a foreign uti1 ity had similar 

failures, involving only a few rods, in fuel assemblies close to the baffle. 

Based on a detailed evaluation, the foreign utility concluded that the 

responsible mechanism was a self-induced assembly vibration having a sharp 

amplitude peak at the characteristic (close to rated) flow (WJ. 

It was noteworthy that the Salem Unit 1 andl Beaver Valley Unit 2, plants 

simil iar to Salem Unit 2, had not observed these type of failures. 

occurred in the first 17x17 zircaloy grid with a specific diameter rod design 

and in the reactor core regions with lead fuel assemblies having the LPD 

zircaloy grid. 

Salem Unit 2 and Beaver Valley Unit 1 plants had observed the GRF. 

The GRF 

Of the eight nuclear power plants with this design, only the 

PDT-94-067, "Maine Yankee Bottom Gridl Bulge Joint Strength Test Report," 

dated March 18, 1994, and PDT-94-167, "C-14 RTN Joint Test Report and Strength 

Capability," dated July 11, 1994. 

4 

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169 
Westinghouse P/PD&D believed that the plants experiencing the GRF had rated 

flows that were close to the characteristic assembly flow, where Wv = 1700- 

1900 gallons per minute (gpm), at which flow, this vibration was excited. The 

vibration frequency had been measured at the CNFD test facilities to be 

= [Deleted pursuant to 10 CFR 2.790 - Document describes specific values]. 

The team noted that, since the vibrating fuel assemblies were located close to 

the core boundary, this frequency might be confirmed by a frequency analysis 

of the measured excore detector response at Salem Unit 1 and Beaver Valley 

Unit 1. 

finite-element analyses and had concluded that rotation of alternate grids 

will eliminate the fuel assembly vibration. 

its grid designs; only this grid design experiences a significant vibration at 

operating flows. The vibration amplitude was believed to be larger for larger 

fuel assembly-to-reactor baffle gaps. 

Although rotating the grids appeared to eliminate the assembly vibration and 

the potential for GRF, tests performed at Columbia University indicated that 

the grid rotation a?so reduced the margin to DNB when the grid-span was 

10 inches. 

grid-span, and CNFD P/PD&D was evaluating these plants and the effects on DNB. 

Westinghouse CNFD P/PD&D had performed extensive vibration tests and 

Westinghouse has tested all of 

However, only two &designed plants currently have the 10-inch 

After evaluating the CNFD P/PD&D response to the fuel assembly vibration 

issue, the team determined that the analysis and testing of the vibration 

mechanism and potential fixes were excellent and demonstrated the broad 

capabilities of CNFD in responding to fuel-operational problems. 

(2) Axial Hislocation of UABA Rods 

As part of its evaluation of reload fuel mechanical design issues, the team 

evaluated the axial mislocation of the WABA rods at Turkey Point Units 3 

and 4, as documented in Turkey Point LER-001, dated January 15, 1993. The 

Turkey Point Cycle 13 reload core design fuel assemblies included the debrisresistant fuel rod design. 

shifted the active fuel upward 1.368-inches. 

corresponding axia! shift in the WABA rods. 

This design had a solid fuel rod end-cap that 

This design change required a 

The Core Engineering group identified these changes and correctly incorporated 

them into its core neutronics and T/H design analyses. However, the Core 

Loading Pattern letter from Core Engineering to P/PD&D did not explicitly 

identify the axial shift in the WABA rods and in this sense was incomplete. 

The Production Initialization meeting CNFD COLA also did not identify the 

necessary design change in the WABA rods, and the subsequent Mechanical Design 

Review (89-02) did not correct this design error. 

In response to this event, CNFD established a Corrective Action Cornittee to 

review the event and define appropriate corrective actions. 

identified the root cause as a failure to treat the relationship of the fuel 

stack to the burnable absorber stack as a specific design criterion. 

The comnittee 

The 

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170 
corrective actions involved reviewing and updating the various data 

transmittals, such as the Core Loading Pattern letter, the DEVL, and the Core 

Engineering METCOM. 

training and trmsmittal distribution lists. 

In addition, the Corrective Action Committee considered 

Although the team concluded that the corrective actions would prevent the 

reoccurrence of this particular design error, the team was not reasonable 

assured that these corrective actions (a) ad2quately addressed the inability 

of the Product Initialization Meeting and the Mechanical Design Review to 

identify the UABA design error and (b) provided reasonable assurance that more 

general design changes in the fuel and core components would be identified in 

future reload core designs. The team considered this to be a weakness in the 

CNFD problem analysis and made the following observations: 

(a) the Production Initialization meetings and Final Mechanical Design Reviews could benefit from Core Engineering (e.g., neutronics and thermalhydraulics engineers) participation, and 

(b) the Design Initialization meeting could benefit from P/PD&D 

participation in these meetings. 

These team observations require no specific action by or written response from 

CNFD. 

3.3.3 Error Free Performance Team 

The team examined error reporting in depth. The inspection of training 

material documented in FA-94-131, "Error Reporting Seminar Package," 

Revision 1, dated September 16, 1994, showed that adequate training was 

provided in software error reporting and the error-handling process. 

Engineering Non-conformance Log for 1994 was reviewed and from this log 

several recent software error reports were traced through the Reportable 

Technology Error, Non-conformance, and Request/Problem Report (R/PR) process. 

The examples evaluated were found to be in compliance with the procedures. 

The 

The notebook, "EFPT Root Cause Analysis and Support Information," Volume 1, 

was reviewed to determine the functioning of the Error Free Performance Team 

(EFPT). The documentation showed that the EFPT was est-$lished in May 1993, 

and discussions with coanizant enqineers revealed that the EFPT had been fully 

functional for about 6 ionths. Tie team evaluation of an incident pertai 

to Sumner from the Incident Status List showed that the root-cause analys 

and corrective action followup was adequately performed and documented. 

history of the EFPT was too short to determine its effectiveness, but the 

process was found to be sound. 

i ng 

he 

S 

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171 
3.3.4 Training 

The training and qualification of the engineering staff were assessed by the 

team during the course of this inspection through discussions with individual 

members of the staff. 

through discussions with the training course administrator and engineering 

managers and through examination of selected training course materials and 

training records of individual engineers. 

The adequacy of the training programs was assessed 

The team determined that CNFD had a formalized, comprehensive training 

program. At the center of the CNFD training program was a 3-week course in 

Nuclear Design Technology and Methods. A17 new members of the engineering 

staff were required to take the course. In addition, technical seminars were 

offered to the entire staff when a new methodology or a new set of computer 

codes was introduced. The instructors were certified by a training program, 

and student evaluations provide feedback on their performance. 

The team reviewed a sample of training records of training activities for 

approximately 10 engineers from both CNFD and NTD. 

the engineers reviewed were consistent with the described training. 

external training was the same as the internal training. CNFD required a 

certain level of training before an engineer was qualified to be a sole author 

of a Calc Note. An additional level was required for an engineer to be a 

verifier . 

The work assignments for 

The CNFD 

The training program at NTD was less formalized and exhaustive than CNFD’s. 

Newly hired staff members participated in an informal Mentor Program for a 

year and then qualify as analysts and reviewer/engineers on the basis of 

qualifications, experience, and on-the-job training. NTD maintains a Training 

Matrix which listed analysis topics or reactor events such as rod withdrawal 

accident. 

activities. 

that staff members remain current in methods and techniques. 

This matrix was checked off as the engineer completed training 

The NTD offers periodic Training Seminars to its staff to ensure 

Based on the its assessment described above, the team determined that the 

training and quiiiifications of the CNFD and NTD staff were adequate fo;. the 

activities in which they were engaged. 

3.3.5 Fuel-Related Inspection Services 

This portion of the CNFU inspection at the Westinghouse Energy Center also 

included an evaluation of the quality program and design activities for 

selected fuel-related inspection services, specifically the design of fuelhandling tools. 

Engineering group of the Nuclear Services Division (NSD) of the y ESBU. 

These activities are performed by the Reactor Cavity Service 

The team evaluated NSD’s design activities for a self-aligning 17x17 removable 

top nozzle (RTN) installation and removal tool. The team found that the 

initial design review, tool qualification testing, and final design review 

were extensive and precisely documented. The final design and functional 

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172 
specification for the RTN tool was also well documented. In addition to this 

multi-phase design review and tool qualification process implemented by NSD, 

the team found the operating procedure for the RTN tool to be controlled and 

we1 1 documented. 

To train and qualify its fueling operations technicians, NSD used competencybased training to develop the skills and knovledge needed to accomplish a 

given fuel-related inspection activity. 

its competency-based Fueling Operations Training Program was that training was 

efficiently matched to the needs for a particular job, and job performance was 

directly improved by such training. 

operations technician, each !$ fuel ing operations trainee must successfully 

complete a three-phase program: Primary Classroom Instruction, Secondary 

Classroom Instruction, and Practical Applications. In addition, candidates 

for senior fueling operations technologist must demonstrate proficiency in all 

phases of fueling operations during a specified minimum number of full or 

modified full-scope fueling operation days at a nuclear plant site, actively 

perform fueling operations during a specified minimum number of fueling 

operations, and pass a review by a Qualification Board. 

According to NSD, the end result of 

To bec:ome a W-certified fueling 

3.3.6 Concl us1 ons 

The team conducted a performance-based audit of CNFD to provide a basis for 

confidence that CNFD products will provide their intended safety functions. 

In order to reach that conclusion, the team evaluated the organization, 

staffing, training, and qualification of the engineering staffs of Core 

Engineering, P/PD&D, and safety and transient analyses. At the end of the 

audit, the team determined that the performance of the CNFD staff, processes, 

and products in each one of these areas was excellent. The team identified 

one instance where the strength of the W ESBU organization was demonstrated 

through the effective interactions between members of CNFD and NTD under 

conditions of considerable time pressure. 

The team observed that a strong quality culture existed within CNFD Core 

Engineering and that good quality assurance (QA) practices were routinely 

performed. The team based this observation on the following i+omr: 

'3) Excellent engineering procedures were utilized in the various 

functional areas. The procedures were rich in methodology and rationale. The 

procedures also utilized good interface control documentation such as 

check1 i sts. 

(b) The METCOM contained core analysis methodology as well as 

procedures for model inputs. 

training was provided and revisions were issued with recipient sign-offs on 

the updates. 

The METCOM was updated frequently and update 

(c) Procedures were revised frequently suggesting that changes to the 

process were being made when required. An illustration of this was found in 

the Software Engineering Methodology manual, which originated in 1989 and 

currently was in Revision 11, dated October 25, 1994. 

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173 
(d) The technology transfer to licensees 

documented via the Software Engineering Methodo 

packages for the initial technology transfer to 

update, the most recent technology transfer, an1 

was controlled and well 

ogy manual . Transmittal 

a 1 censee, a technology 

the transfer of a new code 

were examined. Ail packages were found complete per the applicable QA 

requirements. 

(e) Software error hand1 ing was traceable through the Reportable 

Technology Errors, Non-conformance reporting, and R/PR process as required by 

the error reporting procedures. 

(f) The EFPT was recently initiated at CNFD. The Root Cause Analysis 

and Support Information documentation was reviewed and an inspection of a 

recent incident from the Incident Status List showed that the root cause 

analysis and corrective action followup was documented. 

(g) The initial design review, fuel-related tool collocation testing, 

and final design review by NSD and the operating procedure for the RTN tool 

were extensive, controlled and we1 1 documented; and the competency-based 

Fueling Operations Training Program for &certified fueling operations 

technicians was judged to be comprehensive in the subjects covered and 

hppeared to ensure demonstrated proficiency by certified fueling operations 

technicians. 

3.4 CNFD SDeci a1 tu Metal s Pl ant 

The CNFD Specialty Metals Plant (SMP) in Blairsville, Pennsylvania, 

manufacturers zirconium (Zr) alloy (zircaloy) tubing for use in the nuclear 

power industry. 

(e.g., nuclear fuel pellets through 1960, stainless steel turbine blades, and 

forged bar and strip products), the manufacturing of zircaloy tubing started 

in 1967, and the manufacturing of inLonel steam generator tubing began in 

1968. In 1985, the manufacturing of inconel was discontinued and the CNFD SMP 

was comnitted completely to zirconium-alloy-based nuclear-grade (a) tubing for 

fuel rod cladding, (b) tubing for discrete burnable absorber rod cladding, and 

(c) tubing for thimble tubes, instrumentation tubes, sleeves, spacers and 

connectors. According to CNFD SMP, it has produced over 70 types and sizes of 

zirconium alloy tubing. 

ZIRLO" for longer operating cycles and higher burnups, guide thimbles, and 

burnable absorber tubes are typical products. 

Although production at the plant in Blairsville began in 1955 

Zircaloy-2 for boiling-water reactors, Zr4 for PWRs, 

3.4.1 Brsaduct 

The team conducted this inspection, in part, by interfacing with personnel 

performing specific tubing production operations and with the Product 

Assurance (PA) organization, described in Department Charter BA-700, 'Product 

Assurance Charter,R Revision 7, dated September 23, 1993. 

consisted of the following groups: (a) PA Engineering, (b) PA Operations, 

(c) Customer Projects, and (d) Equipment Reliability. 

was responsible for the finishing inspection activities for all shifts, and 

The PA organization 

The PA Operations group 

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174 
for the operations of Laboratories Physical Testing and Laboratories Gages. 

The PA Engineering group was responsible for document control activities, 

engineering , and UT i ns trument at i on. 

The team determined that PA personnel were functioning as expected; activities 

were performed by trained people according to approved written procedures. 

The team concluded that the PA organization works well and meets QA 

requirements. The team identified no current weaknesses or concerns. 

3.4.2 Customer Requirements 

Customer requirements were imposed on CNFD SMP through standard material and 

design specifications and the CNFD quality program invoked in customer or 

1 icensee purchase orders (POs) . Each customer required the CNFD quality 

program to meet Appendix B to 10 CFR Part 50. CNFD SMP was responsible for 

tubing production from the delivery of the tube reduced extrusion (TREX) to 

CNFD SMP from CNFD WZ through the delivery of finished tubing to CNFD COLA. 

The Zr4 and ZIRLO" TREXs received by CNFD SMP were produced by CNFD WZ in 

accordance with CNFD standard material and design specifications. ChFD SMP 

also produced tubing in accordance with CNFD standard material and design 

specifications. Custom specifications were not used. 

On the basis of its evaluation, the team determined that the quality program 

implemented by CNFD SMP met the requirements of Appendix B to 10 CFR Part 50 

and provided effective control over activities affecting quality. 

3.4.3 Fuel Clad Tubing Production 

CNFD SMP received Zr4 and ZIRLO" TREXs from CNFD WZ. Fuel clad tubing was 

produced by reducing the outside diameter (OD) and the TREXs wall thickness 

through the basic reduction steps to form ak final tube hollow. The final tube 

hollow was then reduced to meet the final dimensional requirements for the 

fuel clad tubing. 

The team observed that CNFD SMP produced fuel clad tubing from TREXs by the 

basic reduction steps described below. 

until the required OD and wall thickness for the final tube hollow were 

achieved. All activities were performed in accordance with Follower Cards 

(travellers) and written procedures. 

These steps would normally be xpeated 

(a) The first tube hollows were produced by cold reducing TREXs through cold pilgering. This process accomplishes tube elongation and wall reduction 

by rolling TREXs back and forth between two grooved dies. 

process, the tubes were rotated and advanced in small increments over a 

stationary mandrel. 

this process. 

During this 

Both the tube diameter- and wall thickness were reduced by 

(b) The first tube hollows were cut to lengths and deburred. 

(c) The first tube hollows were cleaned and pickled. 

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175 
(d) The first tube hollows were vacuum annealed. The tube hollow 

annealing proc'bss heated the material to a specified temperature to achieve 

recrystallization and reduction of the stresses introduced by pilgering. 

(e) The first tube hollows were cold-reduced to produce the final tube 

hollows by cold pilgering. 

(f) The final tube hollows were cut to lengths and deburred. 

(g) The final tube hollows were cleaned and pickled. 

(h) The final tube hollows were vacuum annealed. 

(i) Following the second recrystallization anneal, the third and final 

pilgering pass was performed. Contractile strain ratio (CSR) and hydride 

orientation were developed at this stage from the amount of reduction produced 

in diameter and wall thickness. 

orientation, described in Section 3.4.4 of this report, confirmed the 

effectiveness of this step. 

Subsequent measurement of CSR and hydride 

(J) After pilger reduction, the tubing was cut to specified lengths by 

removing a specified amount from the trailing end and the remainder from the 

leading end. This step removed material that had not received cold work 

consistent with the rest of the tube. 

recorded and the follow card signed by the operators. 

The weight of the tube was measured and 

(k) The tube was then cleaned. The weight was measured and recorded 

and the follow card signed by the operator. 

(7) Thermal stress relief was then performed and the furnace number 

recorded on the follow card. 

texture of the microstructure developed during the final pilgering pass while 

producing a more uniform stress level within the structure. 

measured and recorded and the follow card signed by the operators. 

This operation retained the metallurgical 

The weight was 

(RI) Straicjhtenjng was performed, the weight was moasured and recorded, 

and the follow card signed by the operators. 

(n) The Inside Diameter (ID) of the tube was grit-blasted, the weight 

was measured and recorded, the follow card signed by the operators. 

(0) The tube was cut to length, and the ends were faced, deburred, and 

checked for squareness. The cutoff machine number was recorded, the weight 

measured and recorded, and the follow card signed by the operator. 

(p) The 00 of the tube was polished, and the polisher number was 

recorded. The number of pieces accepted or scrapped was recorded and the 

follow card signed by the operator. 

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176 
(9) Final cleaning was performed by alkaline cleaning, rinsing, and 

drying. The conductivity of the final rinse was controlled to a specified 

maximum value to assure the purity of the final rinse. 

accepted or scrapped was recorded and the follow card signed by the operator. 

The number of pieces 

(r) Alloy verification was performed to assure that the proper Zr alloy 

was being supplied to the customer. 

provide traceabi 1 i ty throughout product 7 ife. 

or scrapped was recorded and the follow card signed by the operator. 

Each tube was 3dentification marked to 

The number of pieces accepted 

On the basis of its evaluation, the team determined that the metallurgical 

implications for the final tube hollow reduction process were the same for Zr4 

and ZIRLO'"; that is, CSR and hydride orientation were affected in a similar 

manner by the final pilgering parameters, and the stress equalization achieved 

was the same in both alloys during thermal stress relief. CNFD SMP produced 

no Zr2 tubing at this time. The team also determined that the manufacturing 

and inspection activities produced fuel clad tubing suitable for its 

application. 

(1) Beta-Quench 

CNFD SMP performed no beta quenching. 

CNFD SMP by CNFD WZ had been suitably beta-quenched by CNFD WZ. 

products produced by CNFD WZ were del ivered to tubing producers other than 

CNFD SMP. 

The Zr4 and ZIRLO" TREXs provided to 

Zircaloy-2 

Nondestructive Examinations 

The team reviewed nondestructive examinations (NDE) of fuel clad tubing. CNFD 

SMP performs UT for OD, ID, wall thickness, and flaw detection on 100% of the 

Zr4 and ZIRLO" finished tubing produced for nuclear application. 

observed UT Level I qualified personnel setup and calibrate UT machine No. 18. 

UT inspection of Zr4 fuel clad tubing was performed by Level I qualified 

personnel. Sorting stations were used to segregate the material tested 

according to flaw and dimensional characteristics requiring further action. 

One station received tubing with acceptable characteristics. 

material was processed in accordance with procedure QC-318, "Dispositioning of 

Fuel and WABA Tubing After Ultrasonic Dimensional and Flaw Inspection," 

Revision 37, dated November 15, 1994. 

The team 

Unacceptable 

(3) Final Inspecti on 

Final inspection consisted of the following steps. 

(a) The ID surface was examined by visually examining the tube ID 

against a lighted background. 

(b) The finish on the ends was examined. The number of pieces 

accepted, reworked, and scrapped was recorded and the Follower Card signed by 

the operator. 

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177 
(c) Length and end squareness were checked. 

(d) Straightness and ID at the ends were checked. The number of pieces 

accepted, reworked, and scrapped was recorded and the Follower Card signed by 

the operator. 

examination was performed to check for surface roughness. 

reworked, and scrapped was recorded and the Fo’llower Card signed by the 

operator. 

(e) The OD surface was visually examined and a mechanized OD surface 

(f) Tube packaging was performed. The number of pieces accepted, 

(4) 

The team evaluated the packaging and shipping of tubing with respect to the 

protection of the metal surface condition during shipping. Full sheets of 

Styrofoam contoured to match the geometry of the tubing were used tl) separate 

the full length of each layer of tubing within heavy wooden boxes lined with 

thick brown paper. The packaging appeared effective and had not resulted in 

any reported shipping damage. 

3.4.4 Quality Services Lab 

Wandl ing, Storage, and Shipping 

The requirements for laboratory physical testing were contained in 

specifications imposed on CNFD SMP by its customers, such as the CNFD COLA. 

The team reviewed specification NFD-31008, “Seamless Improved Zircaloy-4 

Tubing,” Revision 38, dated July 24, 1992, to determine the requirements for 

longitudinal tensile properties, CSR, corrosion resistance, and hydride 

orientation. Samples for testing were pulled from production. 

(1) Tensile Testing 

The team observed that tensile testing was performed. 

Certification Laboratory Test Reports for zircaloy tubing determined that room 

temperatur-9 tensile properties met the requirements of CNFD s+-;-ication 

Review of Lot 

NFO-31808. 

(2) Contractile Strain Ratio 

A relationship has been demonstrated between the contractile strain ration 

) and the crystallographic orientation or texture of the grain structure 

i rcal oy . Texture affec zircaloy tubing in several ways. Texture 

ride orientation, iodine stress corrosion 

The re1 ationships between texture and these 

re known, and h aterial will respond can be predicted. By 

controlling variables in the ~u~ing reduction processs the texture of the 

material can be controlled, thereby producing desired characteristics or 

avoiding undesirable characteristics. 

for characterizing texture and thereby assuring that desired metallurgical 

properties have been produced. 

The CSR measurement provides a method 

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The team observed CNFD SMP performing CSR testing. 

with circumferential lines over a specified gage length and various 

measurements were taken. 

manner and remcasured From these measurements , the radi a1 and 

circumferential strains were calculated and the contractile strain was 

determined. By reference to a graph showing the relationship between CSR and 

a radial texture parameter (f,), the fraction of grains exhibiting radial 

texture can be determined. Radial texture exists when the basal plane of the 

zirconium hexagonal crystal is normal, or perpendicular to the radius of the 

tube. 

A tube sample was marked 

The sample was then strained in the prescribed 

(3) Corrosion Testing 

The team found that corrosion testing was performed. 

Certification Laboratory Test reports for zircaloy tubing determined that 

corrosion test results met the requirements of CNFD specification NFD-31008. 

Review of Lot 

(4) Hydride Orientation 

Hydride formation in the zirconium hexagonal crystal has been shown to prefer 

the basal plane. Therefore, a radial crystal texture produces a 

circumferential hydride platelet orientation. Highly circumferential hydride 

platelet orientation has been shown to be necessary to produce good ductility 

in tubing after corrosion on reactor service. Hydride platelets at angles 

smaller than 40" to the radial direction are classified as radial and others 

as circumferential. The ratio of radial platelets to total platelets is 

defined as the hydride fraction (f,). Since f, can be controlled through the 

tube reduction process, a limit can be set on its value. 

s"ubject of negotiations between the buyer and the suppl ier. 

The limit may be the 

CNFD SMP used procedure QS-503, "Determination of Hydride Orientation of 

Tubing," Revision 6, dated June 8, 1S93, to make the hydride orientation 

determination. A tube sample was carefully cut to avoid introducing 

additional internal stress, flash pickled and hydrided in a controlled furnace 

for a specified time. It was then cooled and carefully prepared for 

metallographic examination; avoiding the introduction of additional internal 

stress. A photomicrograph of a selected area was prepared and f, was 

dchrmined. 

product represented by the sample and the report was returned to manufacturing 

operations for appropriate material disposition. 

A determination was made regarding the acceptability of the 

3.4.5 Instrument Lab 

The team observed that production control and test instruments such as furnace 

controllers and UT and inspection devices were identified with calibration 

instrument numbers traceable to the gage laboratory and were within the 

calibration due dates. Review of documentation and discussions with CNFD SMP 

staff who were responsible for the Instrument Lab, indicated that activities 

were being performed by qualified personnel in accordance with approved 

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procedures. QC technicians calibrated the dimensional standards used to 

perform final inspection, as described in Section 3.4.3 of this report. This 

cal ibration was performed in accordance with Product Assurance Procedure PA212, "Dimensional Standards For Zirconium Alloy Tubing," Revision S, dated 

August 22, 1994. 

3.4.6 Corrective Actions 

The team evaluated the findings of CNFD SMP internal audit ESBU-94-08, issued 

June 8, 1994, and the implementation of corrective actions. The team 

considered the concerns cited in the audit to be an indication of a thorough 

audit. 

corrective action, three of which dealt with customer product processed during 

the audit. CNFD SMP implementation of the corrective and preventive actions 

taken in response to the audit findings were observed to have been performed in a timely manner. 

CNFD SMP internal audit ESBU-94-08 itemized four issues requiring 

3.4.7 Concl us i ons 

The team conducted a performance-based audit of CNFD SMP to provide a basis 

for confidence that CNFD products will provide their intended safety 

functions. In order to reach that conclusion, the team evaluated the 

organization, staffing, training, and qualification of the operators, 

technicians, and PA staff. At the end of the audit, the team determined that 

the performance of the CNFD SMP staff, processes, and products in each one of 

these areas was adequate. 

existed within CNFD SMP and that good QA practices were routinely performed. 

The team based this observation on the following items: 

(a) The flexibility of the CNFD SMP organization. The organization 

empowered individuals to focus their expertise on the activities needed to 

produce a qual i ty product accord i ng to approved writ ten procedures. 

The team observed that a strong quality culture 

(b) Because of the integrated approach, the team could not tell 

management from labor by the dress or conversation. 

(c) Each individual interviewed during this part of the inspection 

appeared to he well trained and knowledgeable of the related production 

operations ad quality requirements. 

On the basis of its evaluations during this portion of the inspection, the 

team did not identify any weaknesses in or concerns with the CNFD SMP 

organization or its activities that affect the quality of the Zr tubing 

manufactured for use in the nuclear power industry. 

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180 
3.5 CNFD Columbia Plant 

The CNFD Columbia Plant (COLA) in Columbiai, South Carolina, was a fully 

integrated fuel fabrication facility performing (a) the conversion of uranium 

hexafluoride (UF ) to uranium dioxide (UO,); (b) fuel pelleting; (c) zirconium 

diboride integraq fuel burnable absorber (IFBA) pelleting; (d) fuel rod 

fabrication; (e) fuel bundle assembly; and (f) the various verifications, 

tests, examinations, and special processes. The PWR fuel fabricated by CNFD 

COLA, which began operation in 1969, is used in reactors originally supplied 

by Y and those supplied by other reactor vendors. CNFD COLA also supplies 

fuel-related core components, such as top and bottom fuel assembly nozzles, 

grids, burnable absorbers, and control rods to its customers about the world. 

The I? fuel assembly mechanical design function was performed by the 

Product/Process Development & Design (P/PD&D) group of CNFD, located at CNFD 

COLA, and was fully integrated with the process development and manufacturing 

activities of CNFD COLA. 

this portion of the inspection, the team also evaluated P/PD&D inputs to and 

interface with the Teload core design and reload safety analysis process 

during the CNFD/Westinghouse Energy Center portion of the inspection. 

Therefore, in the interest of both overall readability and convenience, 

Section 3.3.2 of this report describes the team's evaluation of the P/PD&D 

fuel assembly mechanical design functions that were performed at both the 

CNFD/Westinghouse Energy Center and CNFD COLA. 

The team not only evaluated P/PD&D activities during 

The inspection of CNFD COLA emphasized the manufacturing processes that re1 ate 

to the fuel rod failure mechanisms (e.g. , hydriding, fretting, pellet/cladding 

mechanical interaction (PCMI) , overheating, cladding collapse, bursting, and 

mechanical fracturing). 

3.5.1 Product Assurance 

The team evaluated the Product Assurdnce (PA) organization's independent 

oversight and the adequacy of the verifications, tests, and examinations 

performed by PA and other CNFD COLA organizations. 

the CNFD COLA Plant Yanager at the same organizational -1,evel as the 

Manufacturing Manager. 

structure provided PA with sufficient independence from cost and schedule 

concerns to focus on product safety, as required by Section 1, '"Organization," 

of the y QA topical report. 

3.5.2 Customer Requirements 

The PA Manager reported to 

The team determined that the CNFP COLA organizational 

Customer purchase orders (POs) for fuel assemblies entered through the 

Operating Plant Business Unit (OPBU) organization (vs ESBU), located at the 

Westinghouse Energy Center. 

the Domestic Sales and Customer Projects organization were assigned to 

interface with individual utilities. 

Project Sales Managers and Project Engineers of 

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181 
Domestic Sales and Customer Projects initiated Region Orders and/or Job 

Orders. 

assemblies. 

Materials, Planning and Services and the Product/Process Development and 

Design (Design Specification and Drafting) organizations. 

Region Orders are for an indefinite quantity of an item such as fuel 

Region Orders and Job Orders are issued to CNFD COLA through the 

The NRC inspection team reviewed several orders and determined that the proper 

regulatory requirements were followed in the POs to I? and in CNFD procurement 

documents rev i ewed . 

3.5.3 Procurement 

In the manufacturing of nuclear fuel assemblies at CNFD COLA, the major 

components procured externally were tubing for various appl ications, strip for 

grid assemblies, bar for end plug fittings, and top and bottom nozzles. 

materials (Zr4, ZIRLO", stainless steel, and Inconel) from which these 

components were made were defined by materials specifications and drawings 

that were controlled by CNFD SMP and maintained at CNFD COLA. 

The 

The team selected the following fuel assembly component POs to evaluate the 

CNFD COLA procurement processes: 

POs FC-99408-MGM and FD-15704-MGM issued to CNFD SMP for Zr fuel 

clad tubing per material specifications NFD-31008, "Seamless Improved 

Zircaloy-4 Tubing, " Revision 38, dated July 24, 1992, and NFD-31003, "Seamless 

Improved Zircaloy-4 Thimble Tubes," Revision 22, dated October 12, 1992. The 

tubing was supplied by CNFO SMP under the "one roof manufacturing" concept; 

that is, after being checked for shipping damage it was accepted at CNFD COLA 

and placed in production. 

(a) 

(b) POs 92777 and 92779-BBR and PO FC-92437 issued to the Vallorbs 

Jewel Company of Lancaster, Pennsylvania, for Zr top and bottom end plugs per 

specification NFD-31009, "Cold-Wound Helical Steel Springs," Revision 13, 

dated April 21, 1993. 

they were machined from bar material supplied by CNFD WZ. 

required, some end plugs were machined at CNFD COLA. 

inspected at CNFD COLA in accordance with the requirements of Quality Control 

Instruction QCI-311202, "Fuel Rod End Plugs and End Pilot - Receiving 

Inspection," Revision 89, dated January 23, 1995. 

End plugs were purchased primarily from Vallorbs, where 

In eitbr rase they were 

As occasion 

(c) PO FD-99731-MGM issued to the Associated Spring Company of Corry, 

Pennsylvania, for stainless steel alloy fuel rod springs per NFD-31006, "Cold 

Finished Zircaloy-4 Bar," Revision 18, dated March 4, 1993. 

(d) PO FD-12224-MLN issued to CNFD WZ for Zr4 strip for grid straps. 

Strip for grid strap fabrication was supplied by CNFD WZ, under the "one roof 

manufacturing" concept. 

accepted at CNFD COLA and placed in production. 

After being checked for shipping damage it was 

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182 
(e) POs FD-16063-MGH and FD-16064-MGH issued to Ulbrich Stainless Steel 

of Wallingford, Connecticut, for Inconel strip material for grid straps per 

NFD-31002, "Nickel Alloy 718 Sheet, Strip, and Plate," Revision 19, dated July 

22, 1994. 

(f) Top and bottom nozzles were primarily purchased from L8S Hachine 

Company (L&S) in Latrobe, Pennsylvania. Some nozzles were produced at CNFD 

COLA. L&S was a certified supplier of nozzles to the quality requirements of 

the American Society of Mechanical Engineers (ASME) standard NQA-1, "Quality 

Assurance Requirements for Nuclear Faci 1 ity Applications. " When receiving 

nozzles from L&S, CNFD COLA checked them for shipping damage and the presence 

of the correct information on the paperwork and placed them in production. 

CNFD COLA performed surveillance on L&S, most recently during the period Hay 

23-27, 1994. 

For each of these procurements, the team determined that (a) the requirements 

of 10 CFR Part 21 and Form 102, "QA Requirements For Purchased Material, 

Items, and Services," Revision 5, dated August 31, 1993, were imposed per 

specific drawing and specification requirements; (b) the certified material 

test reports for each PO contained data showing that the materials met all 

requirements including those in the CNFD material specifications; and (c) all 

suppliers were listed on the ESBU Qualified Suppliers List, documented in 

QSA 95-0187, dated January 23, 1995. 

3.5.4 Chemical and Ceramic Operations 

The starting material for the chemical conversion process (ammonium diuranate 

(ADU), or wet conversion process) was uranium hexafluoride (UF ), received in 

large cylinders containing up to 1505 kilograms (kg). 

sampled and analyzed for enrichment verification before being processed 

through one of the production lines. 

Each cyfinder was 

3.5.4.1 Chemical Conversion 

The UF, cylinders were placed in a steam chest and heated to vaporize the UF, 

which was conveyed through pipes to the hydrolysis column, where, through a 

highly exothermic reaction, a solution of UO,F, and HF was produced. The UO,F, 

solution was pumped into the precipitation column, along with a NH,OH 

solution. A recirculation pump continuously recycled the resulting slurry 

(containing particles of ammonium diuranate) to the top of the column. 

Another pump transferred a portion of the recycling stream to a dewatering 

centrifuge. 

into a horizontal dryer. 

heated rotating tube calciner, where H, and steam were introduced to the 

powder flow. 

the fluoride (F). 

A paste-1 i ke product from the dewatering centrifuge was pumped 

The product from the dryer is conveyed into a gasThe H, reduces the U6 to U'& and the steam aided in volatilizing 

The resul ti ng 

The hamnermil 

approximately 

sampled and a 

determine the 

UO, powder was fed directly from the calciner into a hamnermill. 

ed UO, was collected in plastic containers called polypacks in 

10-kg quantities. 

composite sample from polypacks were blended and analyzed to 

ztoichiometric oxygenluranium ratio (O/U), F, BET surface area, 

For process control , each polypack was 

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183 
Fisher Subsieve Sizer surface area, porosity, bulk density, and H,O. 

concluded that each critical process parameter was identified and control 1 ed 

for each process step. 

steps in conversion from UF, to UO, were well defined and controlled and 

ensured expected-qual i ty UO, powder. 

The team 

The team a1 so concluded that the process and sampl ing 

3.5.4.2 Powder B1 endi ng 

The polypacks were stored until analyses are completed. A "picklist" was 

generated, based on these analyses, for selecting the polypacks that will 

constitute an approximately 1500 kg UO, powder blend. The team verified that 

the selection process takes into account the various factors that determine 

pressing behavior and sinterability of the UO, pellets (e.g., bulk density, 

surface area, and O/U). 

A blend was assembled by dumping the UO, from the selected polypacks into a 

bulk blender, which is then rotated for a specified number of revolutions at a 

desired speed, specified in revolutions per minute (RPM). Because enrichment 

blending was sometimes used to achieve a final specified enrichment, the team 

reviewed qual ification report CD-FB-018-082, which establ ished the adequacy of 

the blending process by blending natural and deoleted materials and checking 

homogeneity by isotopic analysis. 

blending process, specified in X-revolutions at Y-RPMs, was adequately 

supported by the test data documented in CD-FB-018-082. 

The team concluded that the specified 

Qualification report PE-EJS-83-015, which qualified the same blender for 

enrichment blending by blending depleted UO, into enriched virgin UO, powder, 

was also reviewed. 

adequately supported by the test data documented in PE-EJS-83-015. 

The team found the specified processing conditions 

After blending, the powder was processed through a hammermill to break up 

agglomerates and tumbled again for X-revolutions at Y-RPMs. 

sampled at four locations and each sample is analyzed for enrichment, %U, O/U, 

H,O, carbon (C), nitrogen (N), F, spectrographic impurities, EBC, sodium (Na) , Zr, bulk density, porosity, and surface area (by both the Fisher Subsieve 

Sizer and BET techniques). The team concluded that the blending and sampling 

steps were well defined and controlled and ensured expected-quality UO, 

powder. 

It is then 

3.5.4.3 Fuel Pelleting 

The 1500 kg UO, blends were processed in 18 kg batches for pellet pressing. 

$0 mixlure was roll compacted and granulated to make the powder suitable for 

pressing. 

blended by rolling the polypack. 

specified rate of pellets per minute. The press operator manually checked 

pellet green density manually at intervals using pellet weight and length data 

described on a control chart. 

controlled the length of each pellet pressed. 

powder is added to each batch for density and sinterability control. The 

Before pelletizing, a die lube was added to the polypack and 

Pellets were formed in a rotary press at a 

Sensors on the press tooling measured and 

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184 
An automatic stacker placed the pellets into sintering boats. 

was generated for each boat. 

time at a specified temperature measured at the middle of the high-heat zone 

of the sintering furnace. 

and the operator manually recorded the temperature once per hour. 

temperature was over-checked twice each shift with an optical instrument. 

A follower card 

Pellets were sintered for a minimum specified 

Sintering temperatures were recorded on a chart, 

The 

Following sintering, the pellets were processed as lots. Each lot consisted 

of four sintering boats. A specified number of pellets were taken at random 

from each lot and ground to the specified diameter and density was measured. 

Based on the density, the lot was accepted or rejected for grinding. 

Low-density pellets could be resintered. Accepted pellets were ground to the 

specified diameter and accumulated into trays for drying. 

pel lets was 100% vi sual ly inspected for acceptance using vi sual standards for 

chipping and cracking. The CNFD COLA inspector's bar-coded badge was read by 

the computer in order for the inspector to enter the inspection data. 

Each tray of 

A specified number of pellets were taken at. equal intervals throughout the 

blend sample for H- analysis, which included hydrogen contained in absorbed 

H 0. 

of the fuel. 

the average H ppm content, special instructions were issued in procedure 

QCI-910210, "60 Pellet Hydrogen Sampling and Release," Revision 69, dated 

December 19, 1954, designed to ensure that the limit was not exceeded. 

Pellets were also pulled for a "high block test" per procedure QCI-910219, 

"UO Pellet Hydrogen High Block Evaluation," Revision 11, dated November 28, 

199z. For a pellet lot to be accepted the lot average must be less than or 

equal to a specified ppm H at the 95% confidence level, and no individual 

test result can exceed a sfightly larger specified ppm H,. The team examined 

this issue in great detail and concluded that CNFD COLA's hydrogen control of 

its fuel pellets was a strength of its pelleting operations. 

After the sample pellets were selected for H analysis, a specified number of 

pellets was selected for SPIDER (System for bellet Inspection Data Entry and 

Retrieval). 

density determination. Perpendicularity, dish depth, and surface roughness 

were also measured on a limited number of these pellets. 

checki;,g, a specified number of these pellets were selected as archive 

pellets. Additionally, a specified number of pellets were selected for pellet 

chemistry measurements. 

were submitted whole for F analysis. The remaining pellets were crushed for 

the remaining analyses (such as enrichment, %U, O/U, Cy N, F, total metallic 

impurities, and equivalent boron concentration). The team determined that the 

CNFD COLA UO, pellet pressing, sintering, sampling, and analyses were we17 

defined and controlled and ensured quality pellets with low H, content. 

3.5.4.4 IFBA Pelleting 

Hydrogen confent of the pellets was one of the more critical properties 

Because the precision of the hydrogen analyzer was only f 50% at 

Each of these pellets were measured and weighed for sintered 

After density 

Some of the pellets selected for pellet chemistry 

CNFD reported that, currently, approximately 25% of a reactor reload contained 

integral fuel burnable absorber (IFBA) rods;. 

rod loading were brought into the IFBA area, which was segregated from the 

regular pel let 1 ine. 

UO, pellets acceptable for fuel 

Pel lets were loaded anto a screen cage holder for 

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185 
sputtering the zirconium di boride (ZrB,) coating onto the cy1 indrical surfaces 

of the pellets. 

transferred to the pellet surfaces in a vacuum chamber. Adherence of the 

coating to the pellets was tested and samples of the coated pellets were 

analyzed for boron and H, content. The pellets were visually examined for 

oxidation using visual comparison standards (oxidation changes the pellet 

color). 

Zirconium diboride from the sputtering targets was 

The team determined that the process methods used for coating the UO, pellet 

surfaces with ZrB,, the IFBA pellet testing, and the IFBA pellet sampling 

methods were well defined and controlled to ensure quality IFBA pellets. 

3.5.5 Fuel Rod Fabrication and Inspection 

The team reviewed the flow of materials (e.g., fuel clad tubing, end plugs, 

plenum springs, and pellets) from receipt of purchased items to release for 

fuel rod fabrication. 

3.5.5.1 Fuel Rod Fabrication 

Fuel rod fabrication began by laser-marking the fuel clad tubing with a unique 

bar code entered into the Rod Accountability Monitoring System (RAMS) to 

maintain the identification and to control the processing of each fuel rod as 

it progressed through the fabrication steps. 

data processing system that uses the stored data and other information to 

control the status of each fuel rod through each step of the fuel rod 

fabrication process. 

authorized production and quality control (QC) personnel to either enter data 

or perform activities in accordance with those procedures. 

RAMS is a computer-controlled 

RAMS also contained approved procedures for access by 

For the fabrication of fuel rods for the Northeast Nuclear Energy Company, 

Millstone Nuclear Power Station Unit 3 (Millstone 3) Cycle 10 reload, the team 

observed the top end of Zr fuel clad tubing laser marked with bar codes. 

Bottom end plugs were then inserted with an interference fit before a weldipg 

operator performed the girth welds to fusion-bond the end plug to the fuel 

clad tube using ttie gas tungsten-arc welding (GTAW) process. ;tir qualified 

welding process was computer controlled and subjected to inprocess checks of 

critical parameters. Instrument calibration, weld current, weld time, postflow time, helium flow rate, argon flow rate, rotation speed, and electrodeto-seam alignment were some of the parameters that were checked. Every bottom 

end plug girth weld was ultrasonically (UT) examined to detect porosity, 

underpenetration, and undercut. An automated visual inspection for weld 

continuity was made of each weld with a magnjfied optical image. 

Inspected pellets were moved to the rod-loading line after the pellet 

identification information had been entered into the Item Control System (ICs) 

by an optical reader. 

and enriched UO, pellets into tubes according to procedures displayed on a 

RAMS monitor at the work station. 

operator to achieve the required pellet stack length and type per RAMS 

instructions. The pellet tray and fuel rod bar code were entered into the 

RAMS through an optical reader. 

The team observed a rod line operator loading natural 

Plenum length gauges were used by the 

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186 
After springs were inserted, top end plugs were inserted automatically and a 

welding operator automatically performed tlhe girth weld to fusion-bond the top 

end plug to the top of the fuel rod. 

processes were computer controlled and given the same in-process checks and 

instrument calibration discussed above to weld the bottom end plugs. Every 

top end plug girth weld was inspected and one rod was pulled from each line at 

the beginning and the end of the day for the Analytical Services Laboratories 

tests. 

The qualified automatic welding 

Each fuel rod was weighed and the data entered into RAMS. 

3.5.5.2 IFBA Rod Fabrication 

IFBA fuel rods were fabricated in a controlled area that was separated from 

the standard fuel rod fabrication. An IFBA fuel rod normally contained about 

6 inches of natural UO blanket pellets at the top and bottom; about [Deleted 

pursuant to 10 CFR 2.760 - Document &scribes a specific value] inches of 

uncoated enriched UO, pellets in the adjacent stack length at the top and the 

bottom; and enriched UO, pellets coated with zirconium diboride (ZrB,) in the 

middle of the rod. 

The team observed the fabrication of IFBA fuel rods for the Millstone 3 Cycle 

10 reload. A dual robotic stacking system collated the coated pellets with 

uncoated and natural pellets for loading into fuel tubes. 

that the IFBA rod fabrication was well defined by procedures and adequately 

controlled to ensure quality IFBA fuel rods. 

The team determined 

3.5.5.3 Rod Inspection 

Rods were inspected by either radiographic examination (RT) or UT systems that 

examined bottom and top girth welds and top seal welds. All RT and UT systems 

were developed by CNFD COLA Level 111’s and the operations of those systems 

were monitored by a CNFD COLA Level I1 and a PA engineer. 

(1) Ultrasonic Examination 

The team observed the UT examination of the top end plug welds on fuel rods. 

The Level I1 operator verified the UT system calibration by processing 

standards through the system: (a) a radial drilled for undercut and porosity, 

(b: a V-notch for underpenetration, (c) an axial drill hole in the face of an 

end plug for underpenetration in the seal weld, and (d) a tungsten inclusion 

in a seal weld for detection by X-Ray fluorescence. System verification was 

performed after a specified number of fuel rods had been inspected. 

(2) Radiographic Examination 

RT (x-ray) examinations were performed on all fuel rods when the production 

load permitted. 

was performed instead of X-Ray. 

maximum a1 1 owabl e underpenetrat i on, OD and 1[D undercut, cracks, 1 aps, seams, 

pipes, lack of fusion, porosity, and inclusions. 

During periods with greater production loads, UT examination 

The RT acceptance standards included the 

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187 
The team observed QC inspectors reading and interpreting radiographic film 

from the RT examination of fuel rods. The team determined that the RT of fuel 

rod girth and seal welds were performed according to documented procedures by 

qualified personnel. This area of the inspection was identified as a strength 

because of the expertise exercised in the development of the RT examination 

systems and the training of personnel that control the critical parameters 

associated with fuel rod weld quality. 

(3) Dimensional and Visual 

The team observed QC inspectors performing dimensional and visual inspection 

of fuel rods. This inspection compared visual examination at 5X magnification 

of all girth and seal welds to visual standards and 100% of fuel rod surfaces 

for clean1 iness, scratches, pits, gouges, arc strikes, and discoloration. All 

fuel rods were sampled for length. IFBA fuel rods were verified for end plug 

identification and all fuel rods were verified for laser mark visibility and 

absence of scatter dwell point, which sometimes occurs during laser marking. 

This condition was seen as tiny bright spots scattered throughout the bar 

code. 

(4) Helium Leak Testing 

He1 ius leak testing was performed to determine the pressure boundary integrity 

of fuel and nonfuel rods. Fuel rods were pressurized with helium prior to 

weld closure and nonfuel rods were not. 

operation. The primary system was automated and all fuel rods were processed 

through it in groups of 25. 

automated system were not helium leak-tested through the secondary system. 

The secondary system was used to identify individual leaking rods from groups 

of 25 that failed the automated system. 

test the non-fuel rods. 

Two leak detection systems were in 

Groups of 25 rods successfully passing the 

This system was also used to leakTo test their integrity, nonfuel rods were placed in a chamber in which they 

were pressurized in a helium atmosphere to drive helium into rod defects. 

They were removed from helium pressurization and placed in tk ??'ium leak 

detector and observed for helium leakage in the same manner as fuel rcds. 

Several qualification reports had been written documenting variables examined 

to establish the processes and define normal operating parameters. 

Responsible personnel were qualified to Levels I, 11, and I11 of American 

Society for Nondestructive Testing (ASNT) standard SNT-TC-lA, December 1988 

Edition. Standard leaks were traceable to the National Institute of Standards 

and Technology (NIST) and were observed to be within date limit. 

observed that activities affecting quality were being performed in accordance 

with written instructions, as required, and determined that measures had been 

established to assure that leak detection processes were controlled and 

accompl i shed by qual i f i ed personnel . 

The team 

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188 
(5) Rod Scanner 

The rod scanrsrs contained californium (C '252) neutron sources, which 

activated the U,, within a fuel rod, causing it to emit gamma radiation. 

Gama radiation was detected and evaluated to determine critical parameters 

used to ensure fuel rod quality, e.g., nonconforming pellets (in diameter 

and/or enrichment), gaps between pellets, plenum spring presence and its 

length, and the pellet stack length. Qualification reports had been written 

documenting vari ab1 es examined to estabi i sh the processes and define normal 

operating parameters. Standard defective fuel rods were periodically used to 

calibrate the scanning systems. The team observed that activities affecting 

quality were being performed in accordance with written instructions, as 

required, and that measures had been established to assure that rod scanning 

processes were control 1 ed and accompl i shed by qual i f ied personnel . 

3-5-6 Fuel Bundle Components 

The following fuel bundle components were evaluated during the inspection. 

3.5.6.1 lop and Bottom Nozzles 

Two types of top nozzles were manufactured; one using a precision casting 

welded to an adapter plate box nozzle and another made of pieces of bar, a top 

plate, and an adapter plate box nozzle welded together. In comparison, two 

types of bottom nozzles were manufactured much the same way, one also using a 

precision casting welded to an adapter plate box nozzle and another made of 

pieces of bar welded to four castings and an adapter plate box nozzle. 

Qualification reports for various welding processes and procedures described 

the parameters investigated to establish welding procedures used for 

fabricating various types of top and bottom nozzle assemblies. 

The team observed that (a) activities affecting quality were being performed 

in accordance with written instructions, (b) methods had been established to 

maintain identification of materials, parts, and components and a computer 

record keeping system was used to track their status through each 

manufacturing and inspection operation, and (c) suitable tee+- .-A 

verifications had been satisfactorily completed for aPtivities affecting 

qual i ty . 

3.5-6.2 Grid Straps 

By observing fabrication and reviewing documents, the team confirmed that grid 

strap assemblies were constructed in accordance with approved drawings, 

procedures, and specifications using materials that met design requirements. 

(I) Stamping 

The team observed operations that automatically stamped Zr grid straps from a 

coil of feed material. The stamping Operation produced both Inconel and Zr 

straps that meet the drawing requirements for many different parts 

representing all grid strap designs. During this portion of the inspection, 

the team observed part 6483E78H03, in-process grid straps, identified with a 

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189 
sheet of notebook paper that stated "Do not use any of these outer straps run 

on 3-2-95." The team was concerned that this material had been put on hold by 

a method and form that was not in the described QA system. CNFD COLA PA 

personnel subsequently revealed that Quality Control Deviation or 

Notifications (QCDNs) were being written by the QC inspector, who had told the 

operator to hold the material until the "with material" QCDN was available. 

Completed QCDNs 14637 and 14674 subsequently released the material for further 

processing. The team expressed concern about procedures that did not provide 

the operator and the QC inspector with clear instructions for holding the 

material. The team saw as a strength, the QC inspector and the operator's 

actions in taking the initiative to hold the questionable matwial. The team 

determined that procedure QCI-000112, "Quality Control Deviation or 

Notification (QCDN)," Revision 27, dated July 16, 1993, was confusing and not 

adequately specific. The team identified this procedural ambiguity as a 

weakness in the QCDN system. The PA Manager agreed that the employees had 

taken the correct actions and that the QCGN hold system will be clarified. 

The team determined that the CNFD COLA actions taken during the inspection 

satisfied the team's concern. 

Vacuum Anneal i ng 

The team reviewed the annealing of Zr straps in the vacuum annealing furnace. 

The annealing produced the properties required after removing the stresses 

induced by the stamping operation. 

parameter and was achieved through the use of an automatic control system. 

The team thought that this operation was a strength since it included an 

annual furnace profile to ensure that the temperatures of annealing loads are 

being achieved by control thermocouples. 

Annealing temperature is a critical 

Asseinbl y 

The team observed the electronically controlled welding of Zr alloy grids by 

laser welding operations. 

issuing material, assembling straps, spot-welding corners, applying braze, 

brazing and annealing, recording furnace data, inspecting grid: -ge-hardening 

grids, recording data, bead-blasting grids, and inspecting the grids. Inconel 

grids were normally brazed with the braze paste being installed automatically 

by robot after the corners of the grid had been spot-welded. The team 

reviewed the brazing and solution annealing of Inconel grid assemblies. 

brazing is done in furnaces that produce sound braze joints and create an 

initial metallurgical structure, which is then age-hardened to achieve the 

required mechanical properties. 

for the required time-at-temperature, followed by rapid cooling. 

The steps of fabricating an Inconel grid were 

The 

The annealing was done in the required vacuum 

Age hardening achieves the required mechanical propfrties with yield strengths 

greater than 155,000 pounds per square inch (lbf/in ). 

included with each furnace load were tested in the Metallurgical Laboratory to 

verify that the time at temperature age-hardening process achieved the 

Tensile specimens 

- 48 - 
required results. 

factory since temperatures and time control are the critical parameters in 

achieving braze quality and strength in the Inconel structural grid 

assembl i es . 

The team found this operation to be a strength at the COLA 

(4) Inspect i on 

Zr and Inconel straps were inspected visually by QC inspectors using hand 

instrumentation, optical comparators, and bright 7 ighting to ensure that the 

straps met drawing requirements. 

and then installed in grid fixtures with one type of Inconel grid and all 

types OC Zr alloy grids being laser-welded. 

Straps and sleeves were assembled by hand 

3.5.6.3 Skeleton Assembly 

The purpose of the skeleton assembly was to provide the structure into which 

the fuel rods could be inserted and thimble tubes into which various nonfuel 

rods could be inserted. 

were the top and bottom nozzles, thimble tubes, and various grid azsemblies. 

The major components comprising the skeletoli assembly 

The team observed (a) that activities affecting quality were being performed 

in accordance with written instructions, as required; (b) that methods had 

been established to maintain identification of materials, parts, and 

components and the computerized record-keeping system used to track their 

status through each manufacturing and inspection operation, appeared to be 

suitable to prevent the use of incorrect or defective materials, parts, and 

components, as required; and (c) that suitable tests and verifications had 

been satisfactorily completed for activities affecting quality, as required. 

3.5.7 Bundle Assembly 

The team observed the magazine loading operation used to construct bundle 

assemblies. Fuel rods, standard and IFBA, were lcaded into a magazine before 

being loaded into a skeleton. 

the required templates until loading was complete. After magazine loading, a 

bar code scanner was used to scan the rods in sequence. 

been given a unique identification that the computer status system recognized 

when the rod’s bar code was scanned. 

type registered in an unexpected location. 

The appropriate fuel rods were loaded through 

Each fuel rod had 

The system would not accept a fuel-rod 

The magazine-loading pattern was determined by the operator’s selection of the 

skeleton to be loaded. 

loading pattern had been assigned to the skeleton before being released. 

team observed the skeleton loading, installation of the bottom nozzle, 

torquing of the thimble screws, expansion of the thimble screws into the 

antirotation feature in the bottom nozzle, installation of the top nozzle, 

strain gage measurement of the force required to remove the top nozzle, 

installation of the locking tubes, and welding of the instrumentation tube 

Only released skeletons could be selected. The 

The 

PI 4. 

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191 
The team observed bow- nd-twist gaging of a final fuel assembly; two stations 

were available to perform this activity, each by a different method. The wash 

and double rinse of the fuel assembly was done with an alkaline cleaner and 

followed by two water rinses. Each fuel assembly was checked for drag using a 

standard control rod assembly suspended from a force gage, and a final visual 

inspection was performed. All of the operations and inspections were 

performed in the sequence specified by a Fuel Assembly Routing Card. 

The team observed that activities affecting quality were being performed in 

accordance with written instructions, as required and that suitable tests and 

verifications had been satisfactorily completed for activities affecting 

qual i ty, as required. 

3.5.8 Fuel-Related Core Components 

The W family of fuel-related core components includes glass burnable 

absorbers, wet annul ar burnable absorbers (WABAs) , holddown assembl i es , thimble plugs, and rod cluster control assemblies (RCCAs). 

3.5.8.1 Burnable Absorbers and Thimble Plug Assemblies 

Burnable absorbers absorb neutrons in the reactor core during the early stages 

of a new load of more highly enriched fuel. 

designed for longer refueling intervals. 

refueling intervals, the burnable absorbers are necessary to maintain a 

controllable neutron flux ccnfiguration in the reactor core. 

used, the burnable absorber dissipates, maintaining a relatively constant 

neutron flux pattern. 

More highly enriched fuel is 

During the early portion of longer 

As the fuel is 

The burnable absorber and thimbl e pl ug assembly was capable of being assembled 

in various configurations, depending on the purpose. The major components of 

the various assembl ies were the spring guide/burnable poison welded assembly, 

the hold-down spring, the hold-down bar, and the various rod types and thimble 

plugs. Rod types included glass burnable absorber rods and WABA Rods. 

The tear observed the fabrication of the spring guide/burnabie pd I ;on assembly 

and the subsequent fabrication of the hold-down assembly, in which the spring 

and hold-down bar was added. These activities were performed in the sequence 

specified by a Hold Down Assembly Routing Card. Final inspection of this part 

was performed by PA personnel. 

The team examined the material specification defining the hold-down bar 

casting, NFD-31016, “Stainless Steel Castings,” Revision 13, dated December 6, 

1993. Paragraph 3.4, “Heat Treatment,” required, in part, that heat treatment 

parameters be reported per Section 4.5. 

that the manufacturer furnish a certified test report showing the results of 

all tests, inspections, retests, and reinspection. A review of the certified 

test reports showed that the heat treatment parameters had not been reported, 

even though CNFD COLA had accepted product. However, the CNFD COLA reported 

that Engineering Change Notice 27066, Revision 0, Change Number YYFF-41000, 

originated January 15, 1995, partly deleted paragraph 3.4. Tk? Engineering 

Change Notice was finalized March 6, 1995, and NFD-31016, Reviaion 14, was 

Paragraph 4.5.1 required, in part, 

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192 
issued the same day. Although the previous condition of accepting product not 

in compliance with the specification was considered a weakness, the team noted 

that the CNFD COLA corrective action was iinitiated before the team observed 

the condition. 

Assembly and inspection was performed according to the sequence specified on 

the Non-Fuel Bearing Routing Card. All manufacturing operations were 

performed by operating personnel according to writte3 manufacturing 

procedures. All inspection activities were performed by PA personnel 

according to written quality instructions. 

3.5. a. 2 WABAS 

UABA pellets were used as the burnable absorber. Assembly and inspection were 

performed according to the sequence specified on the Non-Fuel Bearing Routing 

Card. 

according to written manufacturing procedures. All inspection activities were 

performed by PA personnel according to written quality control instructions. 

All manufacturing operations were performed by operating personnel 

The team observed that (a) activities affecting quality were being performed 

in accordance with written instructions, as required; and (b) methods had been 

established to maintain identification of materials, parts, and components and 

a computerized record-keeping system used to trick them through each 

manufacturing and inspection operation, appeared suitable to prevent the use 

of incorrect or defective materials, parts, and components, as required; and 

(c) suitable tests and verifications had been satisfactorily completed for 

activities affecting quality, as required. 

3.5.8.3 RCCAs 

The rod cluster control assembly (RCCA) provides for the insertion of enough 

negative reactivity into the reactor core to shut down the reactor and 

maintain it in a safe-shutdown condition. The major components of the RCCA 

were the body, to which the control rod drive mechanism attaches; the vanes; 

and the fingers. 

assembly to be built. An absorber rod assembly is attached to each finger. 

The number of vanes and fingers depended on the type of fuel 

The team observed the fabrication of the spider assembly. These activities 

were performed in the sequence specified by a Spider ?ssembly Routing Card. 

All manufacturing operations were done by operating personnel according to 

written manufacturing procedures. All inspection activities were performed by 

PA personnel according to written quality control instructions. 

The team examined the process by which control rod assemblies were attached to 

the spider. Paragraph 3.6, "Chemical Composition," stated, in part, that the 

analysis may be made either chemically or spectrographically by a method of 

analysis approved by the purchaser. CNFD COLA did not document the agreement 

regarding the method of chemical analysis. CNFD COLA personnel imnediately 

contacted the manufacturer of the a1 loy and requested copies of the analytical 

methods used to perform the chemical analyses. 

approved the methods and provided them to ,the team for review before the close 

of the insptction. 

CNFD COLA reviewed and 

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193 
The team observed that (a) with one minor exception, activities affecting 

quality were being performed in accordance with written instructions, as 

required; (b) methods had been established to maintain identification of 

materials, parts, and components and a computerized record-keeping system used 

to track their status through each manufacturing and inspection operation 

appeared suitable to prevent the use of incorrect or defective materials, 

parts, and components, as required; and (c) suitable tests and verifications 

had been satisfactorily completed for activities affecting quality. 

3.5.9 Calibration 

The team reviewed the control and calibration of critical process 

instrumentation and measuring and test equipment (M&TE) . that CNFD COLA had a very large number of process plant instruments and M&TE 

in the calibration program. 

calibrate instruments for determining critical product parameters. 

inspectors reviewed a large number of calibration records and found all the 

instruments to be cal i brated within the scheduled frequencies. 

also reviewed traceability of the standards to the National Institute of 

Standards and Technology (NIST) and found no problems. 

population of items in the calibration system, the adherence to calibration 

schedules was considered a strength. 

The inspectors noted 

However, not all of these instruments measure or 

The 

The inspectors 

Given the large 

However, the team observed that the as-found calibration data was not required 

to be recorded. 

instrument was a weighing scale or the calibration was performed by an outside 

contractor. CNFD COLA stated that a conscience decision was made not to 

include the as-found data. The team noted that as-found data is useful in 

determining whether process instrumentation or M&TE exhibit excessive drift 

and whether an instrument was actually within the calibration tolerances. 

Generally, no as-found data was recorded unless the 

The team noted that, with the exception of the M&TE calibrated by the tool and 

gage group, no evaluations of out-of-tolerance M&TE were performed. For 

instance, 30-40% of the calibration measurements from 0.0000 to 300.0000 

millimeters (m, were out of tolerance for the coordinate measurement machine 

(VIEW 1220) which evaludted the critical measurements of all +).o +?p and 

bottom nozzles. 

of the high number of actual measurements that deviated from the nominal, the 

importance of precise critical nozzle dimensions, and the nozzle drawing 

dimensions, which are in 10 thousandths of an inch, not in mn. The team 

reviewed several top and bottom nozzle drawings and determined that the YIEW 

1220 out-of-tolerance measurements did not exceed the tolerances set by the 

nozzle drawings. 

coordinate measurement machine under a procedure requiring evaluations for 

out-of-tolerance as-found calibration results. The CNFD COLA actions taken 

during the inspection satisfied the team’s concern. 

The absence of a written evaluation was significant because 

As a result of this concern, CNFD COLA placed the VIEW 1220 

3.5.10 Analytical Services Laboratories 

The Analytical Services Laboratories (Metal 1 urgical and Chemical ) provided 

process control and product specification analysis and examinations required 

for the analysis and certification of fuel and fuel-related items. 

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194 
3.5.10.1 Hetall urgical Laboratory 

The CNFD COLA Metallurgical Laboratory processes production samples used to 

verify that no problems have deve'loped ' 

hardening of Inconel grids. Weldments from each rod line were sampled daily 

at the beginning and end of each shift. 

grids contained tensile specimens, which wer2 pulled to verify that the heat 

treatment had achieved the minimum yield strength requirements. 

thr? rod-welding process and the age 

Each age-hardening run of Inconel 

The team determined by microscopic examination that girth and seal weld 

penetration samples met the requirements specified. Rod 1 ine weld samples 

were corrosion-tested in an autoclave according to procedure QCI-108857, 

"Autoclave Operating Procedure For Aqueous Corrosion Testing At [specified "C 

and specified pounds per square inch, gauge (psig) pressure]," Revision 22, 

dated August 3, 1994. The team determined by visual examination that samples 

0532-9-1, 0532-9-2, 0536-3, 0531-7, 0533-8, 0534-2, 0535-1, and 0538 were 

acceptable by visual standards and in accordance with procedure QCI-108819, 

"Corrosion Evaluation and Disposition Practices," Revision 46, dated August 

12, 1994. The team noted that sample 0537 was not with the other samples. 

The review of Form 993, per QCI-108857, determined that sample 0537 had been 

dispositioned acceptable by the a lab technician. 

The team was advised by CNFD COLA that the senior lab engineer was holding 

sample 0537 for engineering review and promptly advised the lab technician 

that the disposition on Form 993 should have been identified as "Engineering 

Review. " 

for a circumferential gouge which might have been caused during rod 

manufacturing. QCDN 15527 documented that the gouge was caused when the 

sample was cut and was not related to the rod manufacturing process. The 

senior engineer emphasized that Form 993 was not used to release material and 

that Form 259 was used for that purpose. lhe team noted that Form 259 in 

QCI-108819 had not been completed for sample 0537. 

training meeting with all ?ab technicians on March 6, 1995, to stress the 

importance of accurately entering all documentation and following current 

procedures. 

Sample 0537 had passed the corrosion test but was being evaluated 

The senior engineer held a 

Prompt corrective action by CNFD COLA personnel was found to be a strength by 

the team, and the need for clarification in qCI-lO8819 relative to QCI-108857 

was found to be a weakness. 

demonstrated by lab personnel could be supported better by clarifying the 

documentation requirements. 

satisfied the team's concerns, 

The team observed that the technical competence 

The CNFO COLA actions taken during the inspection 

3.5.10.2 Chemical Laboratory 

The team observed operations in the chemical1 laboratory, inc'luding testing for 

hydrogen content in the UO, pellets, 

the calibration of the hydrogen analyzers, which measure the amount of 

hydrogen in the UO, pellets. The team had concerns with the apparently large 

calibration tolerances set by procedure, the methods used to check 

calibration, and the implementation of the calibration procedure. 

?he team identified concerns regarding 

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195 
I 

In reviewing calibration records, the team noted that the yearly one-point 

Calibration check of two hydrogen analyzers at 560 amperes found both tQ 

exceed the cal i brat i on tal erances . One of the hydrogen analyzers was reading 

low; however, the other read 660 amperes when checked with a current source of 

560 amperes. The team was informed that procedures do not require written 

evaluations for instruments of this type that exceeded calibration tolerances. 

The team questioned whether the out-of-tolerance analyzer could significantly 

affect its accuracy and, therefore, erroneously allow acceptance of the UO, 

pellets with hydrogen content above the CNFD COLA action limit specified in 

ppm. 

known standard and that, since all calibration checks were within the 

tolerance band of 21.6 micrograms, rto further action was necessary. According 

to CNFD COLA, the 560 ampere-hydrogen analyzer reading should correspond to a 

temperature of 1700 "C. The temperature is an important parameter since at 

1700 'C nearly all the hydrogen in the UO, pellets comes out of solution and 

can, therefore, be measured. However, since the hydrogen analyzer was reading 

high, the temperature was actually less, by about 100 amperes, corresponding 

to a lower temperature. 

temperature, would make it appear that the UO, pellet hydrogen content was 

lower than it actually was because less hydrogen would be driven from solution 

at the lower temperature. 

CNFD COLA stated that the hydrogen analyzers were checked daily with a 

A lower amperage reading, corresponding to a lower 

The team also had concerns with the hydrogen analyzer calibration tolerances 

set by procedure 1-03, "Determination of Hydrogen in Uranium Oxides, Ceramics, 

and Metals," Revision 16. The hydrogen is extracted from the sample by 

heating in an argon atmosphere, and the evolved gases are separated chemically 

and the amount of hydrogen is read out directly in micrograms. However, 

during daily calibration checks with a known NIST titanium standard, the 

procedure allowed a 236% tolerance or f1.6 micrograms from a standard of 

4.47 micrograms. The accuracy stated in the hydrogen analyzer vendor manual 

is fO.10 ppm or 3%, whichever is greater. A review of past calibration 

records indicated that some of the hydrogen analyzers had not detected the 

accuracies as stated in the vendor manual or in the titanium standard. The 

certificate of analysis for the titanium standards stated the amount of 

hydrogen in ppm as 43 (f3). Although the measurement of hydrogen is in 

micrograms, the tolerance set by the CNFD COLA procedure would correspmd to 

k15 ppm. The team noted that during calibration checks, the detected amount 

of hydrogen is sometimes off by 1.3 micrograms from the standard, or about 

13 ppm. 

amount of hydrogen detected should at least be within the k3 ppm tolerance 

stated in the titanium certificate of analysis. 

21.6 microgram tolerance was acceptable; however, there was no evaluation to 

on which to base the tolerance. 

Discussions with the hydrogen analyzer vendor indicated that the 

C#FD stated that the 

The team was also concerned with the method used for calibrating the hydrogen 

analyzer with a blank sample. 

calibrated with a blank tin flux sample that should have no detectable 

~y~rog~ntin flux seemed unusually high, in the range of 30-4596 of the hydrogen 

detected in a titanium sample. In addition, the calibration procedure in 

Section 6.0, "Instrument Calibration," of procedure 1-03 required that, in 

checking the tin flux sample, the digital voltmeter reading should register 

Each month the hydrogen analyzers are 

During calibration, however, the amount of hydrogen detected in the 

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196 
0.00 (f0.05) (no scale specified). This reading is expected since there 

should be no hydrogen in the sample. 

steps in Section 6.0 be repeated several times to achieve a 0.00 (f0.05) 

value. The procedure 

further stated that if the digital voltmeter value does not read 0.00 (f0.05), 

it would be necessary to adjust the blank potentiometer. The inspectors noted 

that on numerous monthly checks, the blank digital voltmeter readings were 

above the value of 0.05 and that there was no documented adjustment of the 

blank potentiometer controls, no written evaluation of the acceptability of 

this practice, and no statement explaining why the blank sample should be 

reading so high. 

voltmeter reading of the standard sample to the blank sample, although this 

practice is not mentioned or included in the procedure. 

Moreover, the procedure and calculations required that all test samples and 

test standards be tested on the 7.0 weight compensator gram scale. 

of the uranium dioxide pellets is generally five grams. 

determined that the weight compensator for three of the hydrogen analyzers 

were set not at 7.0 grams but at 2.0 grams. 

schematic diagram and front view diagram for the model RH-1E hydrogen 

analyzers and determined that the dial setting for the weight compensator 

corresponded to 2.0 grams. The CNFD COLA chemical lab personnel stated that 

the 2.0 gram setting was equivalent to the 7.0 gram setting and that internal 

circuitry had been modified to achieve this equivalency. The CNFD COLA 

personnel could not provide documentation of the internal circuity change or 

whether the dial setting was equivalent to the required 7.0 gram setting. 

The team considered the methods used to calibrate the hydrogen analyzers a 

weakness. The 1 arge difference in accuracy between the hydrogen analyzer 

manufacturer's manual and the CNFD COLA chemical 1 aboratory was significant. 

In addition, documentation of several steps in the calibration procedure were 

not clear. Because the team's concerns regarding these calibration practices 

were not resolved during the course cf the inspection, these concerns are 

unresolved. 

samples (ranging upward from 0.20-0.30 ppm) the apparent inaccuracy in the 

hydrogen analyzer calibration is not considered a safety concern because; a 

30% error would not exceed the CNFD COLA acceptance criteria lower 1:rllit of 

hydrogen ppm for UO, pellets. 

The team identified CNFD COLA evaluation of the weaknesses identified in the 

hydrogen analyzer calibration practices as an open item and requested that NRC 

be notified when CNFD COLA has completed its analysis of the calibration 

practices. (Open Item 95-01-01) 

The procedure further required that 

The team noted that these steps were not documented. 

The team found that lab technicians added the digital 

The mass 

The inspectors 

The inspectors reviewed the 

However, because of the detected hydrogen content in the UO 

3.5.11 Concl usi ons 

The team conducted a performance-based audit of CNFD COLA to provide a basis 

for confidence that CNFD products will perform their intended safety 

functions. 

staffing, training, and qualification of the operators, technicians, and PA 

staff. 

the CNFD CnlA staff, processes, and products in each one of these areas was 

To reach that conclusion, the team evaluated the organization, 

At the end of the audit, the team determined that the performance of 

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197 
adequate. 

COLA and that good QA practices were routinely performed. 

observation on the following: 

The team observed that a strong quality culture existed within CNFD 

The team based this 

(a) The training of QC inspectors was very thorough. QC inspectors are 

retested on a regular basis to maintain their certifications. 

(b) All instruments and M&TE used to measure directly or indirectly a 

critical parameter were calibrated at appropriate intervals. 

(c) The heat treating of Inconel 718 and the annealing of Zr4 is 

controlled automatically by electronic systems that ensure thc required 

properties of critical parameters with low possibility of human error. 

100% RT and UT of fuel rod girth and seal welds were performed 

automatically by systems that minimize human error and provide exceptional 

control of weld quality. The Level 111 and Level 11s exhibited expertise in 

NDE in their use of these systems for controlling fuel rod weld quality. 

(d) 

(e) The RAMS system for material control tracks individual fuel rods 

using bar codes for identification and status. 

(f) The system of evaluating conditions adverse to quality and 

corrective actions by the Corrective Action Committee was excellently 

implemented. 

3.6 CNFD Western Zirconium Plant 

The CNFD Western Zirconium Plant (WZ) in Ogden, Utah, established in 1978, 

transforms zircon sand into zirconium (Zr) hafnium, and zircaloy. Zircaloy 

was fabricated into tubular extrusions, plate, strip, sheet, and bar product 

forms. CNFD WZ produces Zr2 for use in boiling-water reactors (BWRs); Zr4 for 

use in PWRs; and ZIRLO", a &developed advanced zircaloy that contains niobium 

for additional corrosion resistance at high temperature. 

3.6.1 Product Assurance 

The team conducted this inspection, in part, through interfaces with personnel 

performing specific Zr alloy fabrication operations and the PA organization. 

The PA organization consists of the following groups: (a) Quality 

Engineering, (b) Quality Inspecticn, (c) Laboratory Services, (d) Audit 

Services, and (e) Records Control. 

The team determined that PA personnel were functioning as expected; activities 

were performed by trained people according to approved written procedures. 

The team concluded that the PA organization works well and meets QA 

requirements. 

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198 
3.6.2 Customer Requi rements 

The team evaluated the procurement interfaces between CNFD WZ and its sister 

organizations, CNFD SMP and CNFD COLA, to determine how customer requirements 

were passed from CNFD SMP and CNFD COLA to CNFD WZ. Customer requirements for 

various Zr alloys and product forms were expressed in terms of standard CNFD 

material and product specifications. The team reviewed several procurement 

documentation packages and determined that the proper regulatory requirements 

had been invoked. 

established to assure that applicable regulatory and design basis requirements 

were suitably included or referenced in the procurement documents. 

The team also determined that adequate measures were 

3.6.3 Zircaloy Fabrication 

The team evaluated the CNFD WZ zircaloy fabrication processes performed from 

the vacuum arc melting of electrodes through the production of finished tube 

reduced extrusion (TREX) for fuel clad tubing, bar products for fuel rod end 

plugs, and plate, sheet, and strip products for grid spacers and other fuelre1 ated products. 

For its evaluation of the Zr alloy fabrication processes described below, the 

team reviewed the applicable process procedures, work instructions, and 

fabrication travelers, which followed process cut1 3ne descriptions and 

specified production requirements. 

alloy fabrication processes and in addition to the specific strengths 

described below, the team made the following determinations: 

(a) Zr a1 1 oy fabrication operations affecting qual i ty were we1 1 

documented in instructions, procedures, and drawings appropriate to the 

production of Zr alloy products; that the Zr alloy fabrication operations were 

accomplished in accordance with those instructions, procedures, and drawings; 

and that those instructions, procedures, and drawings included appropriate 

acceptance criteria for determining that operations important to quality had 

been performed satisfactorily. 

On the basis of its evaluation of the Zr 

(b) Measures had been established to assure that induction heating/beta quenching processes were control led and accomplished by qual ified persannel 

using qililified procedures within the 1 imits established by documented 

qualification projects. 

(c) A program for verifying and inspecting activities affecting quality had been established and was being executed to verify conformance with 

documented instructions, procedures, and drawings. 

3.6.3.1 Zirconium Sponge and Recycle Uaterial 

CNFD WZ produces pure Zr sponge through the chemical reduction and extraction 

of zircon sand, the raw material feed to the Zr alloy fabrication process. 

The sponge was crushed to facilitate inspection and compaction into briquettes 

for the construction of production melting electrodes. The team observed 

removing unacceptable material from crushed sponge. The acceptable sponge was 

loaded into drums through a process that ensured homogeneity. PA inspectors 

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199 
inspected a sample of the sponge material to be pressed into briquettes and 

joined together by the electron beam welding (EBW) process to produce a 

melting electrode. This electrode was vacuum arc melted to produce an 

evaluation ingot, from which conical samples were milled and chemical analysis 

was performed. 

re1 eased. 

On the basis of this analysis, the sponge lot of drums were 

Recycle Zr alloy material (forge shears, cut off ends, edge trim, and log 

butts) that occurred throughout the Zr alloy fabrication process was c’leaned 

and compressed into briquettes and subsequently assembled, through EBW, into a 

recycle melting electrode. The recycle electrode was vacuum arc melted to 

produce a recycle ingot, from which conical samples were milled and chemical 

analysis was performed. The released recycle ingot was forged into a slab, 

called a spar, used with briquettes of Zr sponge to assembly production 

me1 ti ng el ect rodes . 

3.6.3.2 Vacuum Arc Helting 

The melting electrode material used to produce a conditioned ingot was 

constructed by blending the drums of released sponge with calculated and 

weighed amounts of alloying elements to produce the feed to the compaction 

press. 

calculate the quantity of alloying elements (tin, iron chromium, nickel , and 

silicon) to be added to Zr sponge and a recycled spar of previously analyzed 

chemical composition. The team identified the work of this technician as a 

strength because the calculations and weighing instructions resulted in the 

chemical composition of CNFD WZ products to meet the requirements of customer 

specifications. 

The team reviewed the technician’s use of a computer program to 

Because an off-analysis ingot had been produced when a melting operator 

inadvertently did not add iron to the compaction press feed material, CNFD WZ 

initiated a requirement for a second operator to overview the weighing of 

alloy material by the first operator. 

procedure was followed during the weighing and insertion of alloying elements 

in the blender and that both operators signed the alloy weighing records for 

ingot U03452P. 

The team verified that this overview 

To ensure that impurities such as tungsten, cobalt, and iron were not trapped 

in recycled naterial, spars were RT examined. 

into semicircular briquettes and assembled on either side of a spar and then 

welded (EBW) to construct the melting electrode for the first melt. 

second melt consisted of 2 ingots from the first melt, and the third melt was 

a final vacuum arc melt of the one ingot from the second melt. The 

conditioned ingot, produced by the third vacuum arc melt, was conditioned by 

machining the OD to eliminate visual porosity and UT examined as required by 

contract. 

The feed material was compacted 

The 

Chemical analysis samples were taken at specified locations along the ingot 

length. The results were reported individually for each location; no 

averaging was performed. Chemical analysis results reported out of limit at 

any of the five or six sample locations rejected the ingot. Machining the OD 

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200 
or cropping the end in the vicinity of the out of limit sample was performed 

to eliminate surface effects and additional analysis was performed. This 

action usually brought the chemical composition within acceptable limits. 

Ingots identified as acceptable were assigned to customer requirements. 

3.6.3.3 Forging 

Cast conditioned ingots were forged into 'logs or slabs for further processing 

to final products. Ingots for TREXs, as well as bar and wire, were forged to 

a specified OD and cut into bilJets of a specified length for extrusion and 

pilgering or swaging and drawing. Ingots for plate, sheet and strip products 

was forged into a slabs for rolling. 

3.6.3.4 Extrusion and Pilgering 

After cutting to length, billets were beta quenched using an induction heating 

furnace. Billet temperature was monitored using a two-color infrared 

pyrometer and thermocouples contacting each end of the billet. Proper beta 

quenching is essential to satisfactory corrosion resistance of fuel clad 

tubing. To assure proper beta quenching, qual ification projects were 

performed for each combination of induction heating furnace and billet size 

heated in the furnace. 

relationships among material properties, processing parameters, and equipment 

variables. 

operating variables were selected. 

A-time is a measure of the accumulated time at elevated temperatures during 

alpha annealing and stress relieving. Beta quenching sets the A-time clock to 

zero for subsequent heat treatment time. 

demonstrated optimum corrosion resistance for Zr2, Zr4 and ZIRLO". CNFD UZ 

personnel stated that they used the A-time methodology to govern heat 

treatment operations. 

Qualification projects established the working 

A completed qualification project formed the basis upon which 

referenced by CNFD WZ staff established the concept of A-time. 

Different A-time ranges have 

After beta quenching, a corrosion resistance sample was taken for laboratory 

testing. 

lot would not be released from final inspection without passing the corrosion 

resistance test. The billets were rough machined and a centerline hole bored 

Fabrication operations on the billets in a lot continued, but the 

* Garzarolli, F., Stehle, H., Steinberg, E., and Ueidinger, H., "Progress 

in the Knowledge of Nodular Corrosion," pp 417-430, (R.B. Adamson and L.F.P. 

VanSwam, eds.), Zirconium in the Nuclear Industry: Seventh International 

Syrrposiua, ASTN STP 939, American Society for Testing and Materials, 1987. 

Garzarolli, F., Steinberg, E., and Weidinger, H. G., "Microstructure 

and Corrosion Studies for Optimized PUR and BUR Zircaloy Cladding," pp 202- 

212, (L.F.P. VanSwam and C.M. Euken, eds.), Zirconiun in the Nudear Industry: 

Eighth International Symposium, ASTin STP 1023, American Society for Testing 

and Hateri a1 s , 1989. 

- 59 - 

20 1 
in preparation for extruding. The billets were extruded to a specified OD and 

10. 

and one pilgering pass was performed to produce the TREX. 

Surfaces of the extrusions were conditioned in preparation for pilgering 

3.6.3.5 Rolling 

Forged slabs were hot rolled to plate product, annealed, and UT inspected. 

For sheet and strip material, reductions were obtained by cold rolling. 

the final cold rolling process, sheet supplied as coil product was slit intostrip and final vacuum annealed. 

After 

3.6.3.6 Swaging and Drawing 

A conditioned ingot was forged into a log using similar procedures to those 

observed for forged slabs. 

Bar products were manufactured by successive swaging reductions followed by 

either a salt bath anneal or a vacuum anneal. 

The logs were cut into billets and beta quenched. 

3.6.4 Calibration 

The team evaluated the CNFD WZ calibration laboratories for mechanical, 

instrumentation, and contsoi devices. CNFD WZ had adequately specified 

actions to be taken regarding product accepted by calibrated equipment 

subsequently determined to be out of limit. 

mechanical, instrumentation, and control devices were maintained in a computer 

based system that tracked current cal ibration status and provided advanced 

notification of required calibration activity. 

Calibration records for 

With one exception, all measuring equipment examined was marked with current 

calibration stickers. The team observed three stage micrometers in the 

metallography laboratory that had not been currently calibrated. This 

situation occurred as the result of a failure to include stage micrometers in 

the standard calibration system. As a result of the team observation, CNFD WZ 

proposed to: 

(a) Revise "he laboratory work instructions to verify tt~di -:age 

micrometers have been calibrated and are in good working condition. 

instructions on the required hand1 ing of stage micrometers to prevent damage 

and deterioration. 

Add 

(6) Calibrate the existing stage micrometers by a method providing traceability to NIST standards and maintain calibration certifications in the 

cal i brat ion 1 aboratory . 

(c) Modify the purchasing program to require all stage micrometers 

purchased to be delivered to the calibration laboratory prior to being 

distributed to the end user. Revise calibration instructions to assure that 

new stage micrometers are calibrated by a method providing traceability to 

NIST standards. 

- 60 - 

202 
(d) Assure that the calibration service providing NIST traceability is 

an ESBU qual if i ed suppl i er . 

The CNFD ac-ions taken during the inspet tion satisfied the team's concerns. 

3.6.5 Laboratory Services 

The team reviewed the metallographic examination of 20 samples of extruded Zr4 

bar from ingot 403432P. 

at 50x magnification to verify that the tail ends of the first extrusions were 

free from carbon defects. 

This visual examination traversed mounted specimens 

The chemical analyses of samples obtained from the evaluation ingot, the 

recycle ingot, and the conditioned ingot were performed in the North 

Laboratory which was located for easy access to the vacuum arc furnaces. 

team observed a laboratory technician prepare samples, cal ibrate the ion 

coupled plasma (ICP) spectrometer using known standards, and analyze 

evaluation ingot sample 95057. 

information management system (LIMS) compared millivolt (mv) readings with a 

calibration curve of mv and percent of each element in a standard traceable to 

the NIST. 

analyses were performed. The skill of the personnel and the quality of the 

analytical equipment and computer software used to perform chemical analyses 

were observed as strengths by the team. 

The 

Computer software on the 1 aboratory 

Chemistry reports were automatically printed by LIMS as the 

3.6.6 Training 

The team evaluated the 'ndoctrination and training of certain personnel to 

determine that the training provided was appropriate to the activities 

performed and that, where appropriate, refresher training was provided. The 

team observed that in all cases appropriate and sufficient training was 

provided on the schedule required by the training system. CNFD WZ had 

recently implemented a computer based training record system which made 

retrieval of training information easier and also provided timely reminders of 

required training activity. 

The team determined that a program providing for the indoctrination and 

training of personnel performing activities affectin, quality had been 

established and maintained. The team also determined that the training 

program, including the computer based trainiing record system, as a strength of 

the CNFD WZ operation. 

3.6.7 Concl usi ons 

The team conducted a performance-based audit of CNFD WZ to provide it basis for 

confidence that CNFD products will provide their intended safety functions. 

In order to reach that conclusion, the team evaluated the organization, 

staffing, training, and qualification of the uperators, technicians, and PA 

staff. At the end of the inspection, the team determined that the performance 

- 61 - 

203 
of the CNFD WZ staff, processes, and products in each one of these areas was 

adequate. 

WZ and that good QA practices were routinely performed. The team based this 

observation on the following items: 

The team observed that a strong quality culture existed within CNFD 

(a) The personnel skill of a technician’s use of a computer program to 

calculate the quantity of alloying elements added to Zr sponge because the 

calculations and weighing instructions resulted in the chemical composition of 

CNFD WZ products to meet the requirements of customer specifications. 

(b) The skilled personnel and the laboratory equipment used to perform 

the chemical analyses that verify the chemical composition of ingots that 

result from the calculations performed by the above technician. 

(c) CNFB WZ implementation of the just-in-time fabrication methodology and the computer based documentation and tracking systems (e.9. , LIMS). 

(d) CNFD WZ identification, recognition by staff, and control of 

critical parameters such as chemistry, metallurgical structure, and freedom 

from defects in Zr alloy products produced by CNFD WZ. 

3.7 10 CFR Part 21 

During this inspection of CNFD, the team evaluated the ESBU procedure, the 

CNFD procedure, and CNFO plant specific procedures that address the 

requirements of 19 CFR Part 21. 

and CNFD procedures met the requirements of 10 CFR Part 21, the evaluation 

identified minor weaknesses in the conformance of the plant specific 

procedures with the CNFD procedures such that, when the CNFD plant procedures 

are taken together as written, they could result in the failure to evaluate 

deviations. 

While the evaluation determined that the ESBU 

CNFD responded to this concern by drafting a comon procedure that addressed 

the identified weaknesses. 

integrated into the plant specific administrative procedures iw ..it0 the EP 

manual. 

concerns. 

According to CNFD, the cmon procedure will be 

The CNFD actions taken during the inspection satisfied the tearn’s 

- 62 - 

204 
APPENDIX A 

PERSONS CONTACTED 

The U.S. Nuclear Regulatory Commission staff participating in the inspections 

of Westinghouse Electric Corporation, Commercial Nuclear Fuel Division (CNFD) 

facil i ties at the Westinghouse Energy Center, CNFD Specialty Metals P1 ant 

(SMP), CNFO Columbia Plant (COLA), and the CNFD Wester Zirconium Plant (WZ) 

and the Westinghouse Electric Corporation personnel contacted during these 

inspections are listed below and designated as follows: 

indicates that person attended the entranc:e meetings; a dagger (t) indicates 

that person attended the interim exit meetings; and a double dagger (*) 

indicates that person attended the interim exit meetings, or the exit meeting, 

via teleconference or videoconference. 

a bullet (0) 

CNFDlWes tinahouse Enerw Center - February 5-10. 1995 

Westi nghouse El ectri c Corporation: 

t Allison, D.K. 

0 t Ament, G.G. 

* Bartman, T.A. 

Bell, R.M. 

t Camden, T.M. 

t Casadei, A.L. 

t DeWitt, M.M. 

$ Fici, J.A. 

t Hauser, C.S. * Hinson, H.H. 

t Holbrook, D. 

Hoskins, K.C. 

Iannucci, J.V. 

t Johansen, B.J. 

Kapil, S. 

$ Keelen, E.E. 

$ Kirby, W.E. 

Koga, R.H. 

t Komosinski, J. 

t Livingston, L.A. 0 McKinley, D.J. 

t Miller, R.S. 

t Petrarca, D.J. 

t Ray, S. 

$ Rice, G.F. 

Engineer, Development Programs, CNFD 

Sr. Engineer, Nuclear Safety Analysis (NSA), Nuclear 

Techno1 ogy Di vi s i on (NTD; Manager, Product/Process Development & Design 

(P/PD&D) , CNFD COLA 

Fuel Project Engineer, Operating Plant Business Area 

(OPBA) , ESBU 

Manager, Technology Product Services (TPS), Core 

Engineering, CNFII 

Manager, Core Engineering, CNFD 

Manager, Quality Policy Deployment (QPD), Quality and 

Strategic Management (QaSM), ESBU 

Plant Manager, CNFD COLA 

Manager, Transient Analysis, NSA, NTD 

Manager, Customer Support & Product Records (CS&PR), 

Product Assurance (PA), CNFD COLA 

Manager, U.S. Sales, OPBA, ESBU 

Fuel Project Engineer, OPBA, ESBU 

Engineer, NSA, NTD 

Manager, Core Design D, Core Engineering, CNFD 

Manager, Core Design A, Core Engineering, CNFD 

Manager, PA, CNFD COLA 

Manager, Design Specification & Drafting (DSaD), 

P/PD&D, CNFD COLA 

General Manager, CNFD 

Project Manager, OIPBA, ESBU 

Project Jobs Manager, OPBA, ESBU 

Software Engineer, Software Product Engineering (SPE) , Core Engineering, CNFD 

Manager, Fuel Performance Technology, P/PD&D, CNFD 

Sr. Project Engineer, OPBA, ESBU 

Manager, Fuel Licensing Integration (FLI), Core 

Engineering, CNFD 

Qual i ty Engineer, Qual i ty Systems Assessment (QSA) , Q&SM, CNFD COLA 

20 5 
Savage, C.R. 

t Scarfutti , J.P. 

t Shefcheck, J. 

t Tylman, L.3. 

Weber, H.J. 

Werner, H.B. 

APPENDIX A Continued 

Manager, Core Design 6, Core Engineering, CNFD 

Project Sales Manager, OPBA, ESBU 

Manager, Fuel Analysis, Core Engineering, CNFD 

Manager, RCSE, NSD 

Fuel Project Engineer, OPBA, ESBU 

Principal Engineer, PA, CNFD SMP 

U.S. Nuclear Regulatory Connrission: 

t Carew, Dr. J.F. 

t Cwalina, G.C. 

t Grow, R.L. 

t Kendrick, E.D. 

t Lacy, P.S. 

t Matthews, S.M. 

t Neogy, Dr. P. 

Neutronics Specialist, Brookhaven National Laboratory 

Chief , VIS/TSIB/OOTS 

Neutronics Specialist, Pardmeter, Inc. 

Reactor Engineer, SRXB/DSSA 

Neutronics Specialist, Parameter, Inc. 

Qual i ty Assurance Special i st, VIS/TSIB/DOTS 

Neutronics Specialist, Brookhaven National Laboratory 

CNFD SDecialtv Hetals Plant - February 6-10. 1995 

Westinghouse Electric Corporation: 

% Anderson, S.E. * Baird, J.E. * Bates, J.F. 

Brunner, D.R. 

t Butina, M.A. 

t Carrera, F.J. 

t Conroy, K.L. * Cook, J.F. 

DeWitt, M.M. 

t Ewing, J.H. 

% Fici, J.A. 

t Gideon, B.W. 

t Goblinger, 6.0. 

$ Gray, D.G. 

t Hahn, V.D. 

t Haskins, T.E. 

* Hinson, H.H. 

t Jacobsen, W.A. 0 t Jones, B.R. 

t Jorstad, S.C. 

t Judkins, C,L. 

0 Kaiser, R.S. 

t Katsuleris, G. 

t Keefe, R.W. * Keelen, €.E. 

t Kesterson, R.L. 

t Kirby, W.E. 

t Kutchenriter, K.W. 

Manager, Quality Engineering, PA, CNFD WZ 

Auditor, Audit Services, PA, CNFD WZ 

Manager, PA, CNFD WZ 

Controller, Administrative Services, CNFD SMP 

Manager, Production, CNFD SMP 

Process Engineer, Production Services (PS), CNFD SMP 

Intern, Human Resources 

Supervising Engineer, Engineering, CNFD WZ 

Manager, QPD, Q&SM, ESBU 

Manager, PA, CNFD SMP 

Plant Manager, CNFO COLA 

Manager, Process & Equipment Development, PS, 

Finishing Supervisor, Production, CNFD SMP 

Quality Control Engineer, PA, CNFD 'WZ 

Supervisor, Pickle & Anneal, PS, CNFO SMP 

Manager, Maintenance and Equipment Re1 iabil ity, 

Manager, CS&PR, PA, CNFD COLA 

Process Engineer, Production Services, CNFD SMP 

Manager, PA Engineering, CNFD SMP 

Communications, Human Resources, CNFD SMP 

Audit Coordinator, PA, CNFD WZ 

Manager, PS, CNFD SMP 

Manager, Tube Forming Services, PS, CNFD SMP 

Lead Human Resources Representative, CNFD SMP 

Manager, PA, CNFD COLA 

Engineer, Fuel Performance Technology, CNFD SHP 

Manager, DS&D, PJPDID, CNFD COCA 

Sr. Engineer, Tube Forming Services, PS, CNFD SMP 

CNFD SMP 

PS, CNFD SMP 

- A-2 - 

206 
APPENDIX A Continued 

t Leiphart, C.R. 

t Leysock, G.E. 

t Mitchell, C.R. 

Monaco, C I. 

$ Morris, M.W. 

t Narayan, 3.8. 

t Pielert, M. 

t Pokrzywinski, R.A. * Rice, G.F. 

t Sanders, T.M. 

t Schoenberger, 6.3. 

t Sekera, K.R. 

t Shirlej, Y.R. 

t Skuplen, C.L. 

t Trosell, K.N. 

t Troxell, M.A. 0 t Weisser, R.J. 

t Werner, H.B. 0 t Whitehead, W.C. 

Supervisor, Inspection, CNFD SMP 

Manager, Tube Finishing Services, CNFD SMP 

Manager, Human Resources, CNFD SMP 

Sr. Engineer, Custcmer Support, PA, CNFD SMP 

Quality Control Engineer, PA, CNFD WZ 

Fellow Engineer, Customer Support, PA, CNFD SMP 

Supervisor, Pilger, PS, CNFD SMP 

Manager, Storeroom, Plant Services, CNFD SMP 

Quality Engineer, QSA, Q&SM, ESBU, CNFD COLA 

Manager, Plant Services, CNFD SMP 

Engineer, Fuel Performance Technology, CNFD SMP 

Manager, Computer Apgl ications, CNFD SMP 

Supervisor, Pickle & Anneal, PS, CNFD SMP 

Manager, Planning & Control, Administrative Services, 

Manager, Materials, Administrative Services, CNFD SMP 

Supervisor, Inspection, PA, CNFD SMP 

Manager, PA Operations, CNFD SMP 

Principal Engineer, PA Engineering, CNFD SMP 

Plant Manager, CNFID SMP 

CNFD SMP 

U.S. Nuclear Regulatory Comnission: 

t Brewer, D.H. 

t Cilimberg, R.L. 

t Cwalina, G.C. 

t Matthews, S.M. 

Metallurgical Engineer, VIS/TSIB/DOTS 

Metallurgical and Fuel Specialist, Parameter, Inc. 

Chief, VIS/TSIB/DOTS 

Quality Assurance Specialist, VIS/TSIB/DOTS 

CNFD Columbia Plant - February 27-March 10. 1995 

Westinghouse Electric Corporatfon: 

t Allen, R.J. 

$ Allison, D.K. 

t Alstadt, C.D. 

t Bartman, T.A. 

Batson, D.M. 

0 t Berry, J.J. 

t Bush, J.R. 

t Davis, 6. 

DeWitt, M.M. 

0 t Fici, J.A. 

t Flake, R.M. 

t Goldbach, 0.2. 

0 t Goodwin, W.L. 

t Hinson, H. 

Manager, Backshift Operations, Manufacturing, 

Engi neer, Devel opment Programs, CNFD 

Manager, IFBA & Rod Manufacturing (IFBA/RM), 

Manufacturing, CNFD COLA 

Manager, P/PD&D, CNFD COLA 

Manager, P1 ant Systems Engineering, Technical Services 

(TS), CNFD COLA 

Manager, PA Engineering, PA, CNFD COLA 

Manager, Manufacturing, CNFD COLA 

Controller, CNFD COLA 

Manager, QPD, QUH, ESBU 

Plant Hanager, CNFU COLA 

Team Manager, Uranium Recycle & Recovery Services 81 

Pellet Manufacturing (URRS/PPI), Manufacturing, 

CNFD COLA 

Managzr, Process Engineering, TS, CNFD COLA 

Manager, Regu1ator.y Affairs, CNFD COLA 

Manager, CS&PR, PA, CNFD COLA 

CNFD COLA 

20 7 
APPENDIX A Continued 

$ Holbrook, 0. 0 t Keelen, E.E. 

Kirby, Y.E. 

Knott, R.P. 

t Lowder, G.T. 

t Marshall, T.A. 

Matthews, D.E. 

t McDonald, S.G. 

Menke, H.F. 

0 t Miller, J.C. 

O'Cain, M.B. 

t Perkins, C.J. 

* Petrarca, D.J. 

Pollard, R.K. 

t Precht, D.J. 0 t Pregnall, R.A. 

t Rawlings, W.J. 

t Rice, G.F. 

Roberts, E. 

Z Sabol, G. 0 Sieradzki, G.B. 

Sloan, C.K. 

t Smith, J.Q. 

t Trevett, D.R. 

t Ward, U.W. 

Williams, R.A. 

t Workman, G.D. 

Manager, U.S. Sales, OPBA, ESBU 

Manager, PA, CNFD COLA 

Manager, DS&D, P/PD&D, CNFD COLA 

Manager, P/PD&D, CNFD COLA 

Manager, Maintenance, Manufacturing, CNFD COLA 

Manager, Conversion Services, Manufacturing, CNFD COLA 

Manager, Inspection, PA, CNFD COLA 

Manager, TS, CNFD COLA 

Manager, Product Design, P/PD&D, CNFD COLA 

Sr. Engineer, CS&PR, PA, CNFD COLA 

Engineer, Program Manger, P/PD&D, CNFr) COLA 

Manager , Mec ban i ca 1 Manu f ac t uri ng , Manuf act uri ng , 

Sr. Project Engineer, OPBA, ESBU 

Manager, Manufacturing & Industrial Engineering 

Manager, Materials Planning & Control , CNFD COLA 

Principal Engineer, PA Engineering, PA, CNFD COLA 

Sr. Project Engineer, CS&PR, PA, CNFD COLA 

Qual i ty Engineer, QSA, Q&SM, ESBU, CNFD COLA 

Manager, Materials & Mechanical Process Development, 

Manager, Devel opment Programs , CNFD 

Manager, T/H Testing Analysis, P/PD&D, CNFD COLA 

Design Engineer, DS&Q, P/PD&D, CNFD COLA 

Sr. Manufacturing Engineer, MLIE, TS, CNFD COLA 

Manager, Component Services, CNFD COLA 

Manager, URRS/PM, Manufacturing, CNFD COLA 

Advisory Engineer, Regulatory Affairs, CNFD COLA 

Manager, Analytical Services Laboratories, PA, 

CNFD COLA 

(ME), TS, CNFD COLA 

P/PD&D, CNFD COLA 

CNFD COLA 

U.S. Nuclear Regulatory Comnission: 

t Brevar, D.H. 

t Brite, D.W. 

0 Carew, Qi-. J.F. 

t Cilimberg, R.L. 

t Galls, R.M. 

Kendrick, E.D. 

t Matthews, S.M. 

t Hendez, R. 

Metallurgical Engineer, VIS/TSIB/DOTS 

Conversi on/Pel 1 et i ng Speci a1 i st, Bat tell e Paci f i c 

Neutronics Specialist, Brookhaven National Laboratory 

Metallurgical and Fuel Specialist, Pardmeter, Inc. 

Chief, TSIB/DOTS 

Reactor Engineer, SRXB/OSSA 

Quality Assurance Specialist, VIS/TSIB/DOTS 

Electrical / I&C Speci a1 i st, VIS/TSIB/DOTS 

Northwest Laboratory 

CNFD Western Zirconium Plant - Uarch 20-24. 1995 

Westinghouse Electric Corporation: 

t Anderson, S.E. Manager, Quality Engineering, PA, CNFD W? 0 t Baird, J.E. Auditor, PA, CNFD WZ 

t Bates, J.F. Manager, PA, CNFD 'CJZ 

- A-4 - 

208 
APPENDXX A Continued 

t Clawson, R.L. 

t Cook, T.F. 

t Cordingly, D.C. 

$ DeWitt, M.M. 

Doughterty, J.P. 

0 Evans, S.C. 

Galm, R.D. 

Gray, D.G. 

Griffin, L.E. 

Inniss, R.W. 0 t Judkins, C.L. 

Kahler. D.W. 

0 Monson, 6.F. 

Morris, M.V. 

t O’Dell, K.G. 0 t Propst, R.L. 

t Seppich, T.B. 

t White, K.R. 

Trainer, CNFD WZ 

Acting Engineering Manager, Engineering, CNFD WZ 

Manager, Maintenance, CNFD WZ 

Manager, QPD, Q&SH, ESBU 

Sr . Process Engi neer, Engi mer i ng , CNFD WZ 

Sr. Process Engineer, Engineering, CNFD WZ 

Controller, CNFD WZ 

Sr. QC Engineer, PA Engineering, CNFD WZ 

Sr. QC Engineer, PA Engineering, CNFD WZ 

Manager, Human Resources, CNFD WZ 

Audit Coordinator, PA, CNFD WZ 

Act i ng Mater i a1 s Manager, Product i on P1 ann i ng & 

Sr. Process Engineer, Engineering, CNFD WZ 

Sr. QC Engineer, PA Engineering, CNFD WZ 

Trainer, CNFD WZ 

Acting Plant Manager, Production Manager, CNFD WZ 

Manager, Quality Inspection, PA, CNFD WZ 

Supervisor, Laboratory Services, PA, CNFD WZ 

Control Manager, CNFD WZ 

U.S. Nuclear Regulatory Comnission: 

t Brewer, D.H. Metallurgical Engineer, VIS/TSIB/DOTS 

Cilimberg, R.L. Metallurgical/Fuel Specialist, Parameter, Inc. 

t Matthews, S.M. Quality Assurance Specialist, VIS/TSIB/OOTS 

Exit Neetinq. CNFDIWestinqhouse Enercly Center - ADril 13. 1995 

Westinghouse Electric Corporation: 

Allison, D.K. 

Bartman , T. A. 

3 Bates, J.F. + Baird, J.E. 

3 Bush, J.R. 

3 Butina, M.A. 

Casadei, A.L. 

Camden, T.M. 

DeWitt, M.M. 

3 Ewing, J.H. 

3 Fici, J.A. 

3 Gerwels, R.E. 

3 Judkins, C.L. 

3 Kahler, D.W. 

Johansen, B. J . 

3 Kaiser, R.S. 

3 Keelen, E.E. 

Engineer, Development Programs, CNFO 

Manager, P/PD&D, CNFD COLA 

Manager, PA, CNFD WZ 

Auditor, PA, CNFD WZ 

Manager, Manufacturing, CNFD COLA 

Manager, Production, CNFD SMP 

Manager, Core Engi rreeri ng , CNFD 

Manager, TPS, Core Engineering, CNFD 

Manager, QPD, Q&SM,, ESBU 

Manager, PA, CNFD SMP 

Plant Manager, CNFI) COLA 

Plant Manager, CNFD WZ 

Manager, Core Design D, Core Engineering, CNFD 

Audit Coordinator, PA, CNFD WZ 

Manager, Production Planning & Control, Materials, 

Manager, PS, CNFD SMP 

Manager, PA, CNFO COLA 

CNFD WZ 

- A-5 - 

209 
Koga, R.H. * Mitchell, C.R. * Propst, R.L. * Rice, G.F. * Sanders, T.M. * Seppich, T.B. 

Shefc hec k, 3. * Trosell, K.N. 

Utlak, R. * Uhite, K.R. 

Whitehead, W.C. 

APPENDIX A Continued 

General Manager, CNFD 

Manager, Human Resources, CNFD SMP 

Manager, Production, CNFD MZ 

Quality Engineer, QSA, Q&SH, ESBU, CNFD COLA 

Manager, PS, CNFD SMP 

Manager, Quality Inspection, PA, CNFD WZ 

Manager, Fuel Analysis, Core Engineering, CNFO 

Manager, Materials, Administrative Services, 

Sr. Engineer, Administration, CNFD 

Manager, Laboratory Services, PA, CNFD WZ 

Plant Manager, CNFD SMP 

CNFD SMP 

U.S. Nuclear Regulatory Coarnission: 

Brewer, D.H. Metallurgical Engineer, VIS/TSIB/DOTS 

Gallo, R.M. Chief, TSIB/DOTS 

Matthews, S.M. Quality Assurance Specialist, VIS/TSIB/DOTS 

- A-6 - 

21 0 
APPENDIX B 

GLOSSARY OF KRONYHS 

ADU 

ALPS 

ANC 

ASME 

ASNT 

Bef f 

BORDER 

BOM/KS 

BUR 

C 

CAOC 

CATD 

CEC 

COLA 

COLR 

CRDL 

CSR 

DEVL 

DFBN 

DNBR 

EBW 

EFPT 

EP 

ESBU 

F 

F 

FbC 

FUDDL 

GDC 

gPm 

GRF 

GWD/MTU 

H 

HFP 

HL&P 

Hf 

Hz 

HZP 

ICP 

ID 

I FBA 

I FM 

kg 

LIMS 

Amnonium Diuranate Conversion Process 

Advanced Loading Pattern Search 

Advanced Nodal Code 

The American Society of Mechanical Engineers 

American Society for Nondestructive Testing 

Bet a-ef fec t i ve 

Boron Design Requ i remen t s 

Bill of Material/Key Sheet 

Boiling-water Reactor 

Carbon 

Constant Axial Offset Control 

Contract and Technical Data 

Comnonweal th Edi son Company 

Cal i fornium 

Commercial Nuclear Fuel Division 

CNFD Columbia Plant 

Core Operating Limits Report 

Current Design List 

Contractile Strain Ratio 

Design Eva1 uat i on Veri f i cat i on Li st 

Debris-Filter Bottom Nozzle 

Departure from Nucleate Boiling Ratio 

Electron Beam Welding 

Error Free Performance Team 

Engineering Procedure 

Energy Systems Business Unit 

Fl uori de 

Hydride Fraction 

Radial Texture Parameter 

Fuel Parameters Check1 ist 

Fuel Design Data List 

General Design Criteria 

Gallons Per Minute 

Grid-To-Rod Fretting 

Gigawatt-Days Per Metric Tonne of Initial Uranium Metal 

Hydrogen 

Hot Full -Power 

Houston Lighting and Power Company 

Haf n i um 

Hertz 

Hot Zero-Power 

Ion Coupled Plasma 

Inside Diameter 

Integral Fuel Burnable Absorber 

Intermediate Flow Mixer Grids 

Ki 1 ogram 

Laboratory Informati on Management System 

- B-1 - 
APPENDIX B Continued 

LOCA 

LPD 

mv 

METCOM 

MTC 

N 

N 

NDR 

NIST 

NSA 

NSD 

NTD 

OD 

OPBU 

PA 

PCMI 

PDMS 

PO 

P/PD&D 

PPm 

PSE&GC 

PWR 

QA 

QCDN 

RAOC 

RAMS 

RCCA 

REA 

RIQ 

R/ PR 

RSAC 

RSE 

RSER 

RSLC 

RTN 

SAS 

SCE&G 

SER 

SMP 

STP 

T/H 

TIR 

TS IB 

TREX 

NBE 

UF6 

uo2 UT 

VIS 

w 

Loss-of-Cool ant Accident 

Low-pressure Drop 

Mil 1 i vol t 

Met hods Comnun i cat i on 

Moderator Temperature Coefficient 

Nitrogen 

Sod i um 

Nondestructive Examination 

Nuclear Design Report 

National Institute of Standards and Technology 

Nuclear Safety Analysis 

Nuclear Services Division 

Nuclear Technology Division 

Outside Diameter 

Operating Plant Business Unit 

Product Assurance 

Pellet/Cladding Mechanical Interaction 

Project and Design Milestone Schedule System 

Purchase Order 

Product/Process Development & Design 

Parts-per-million 

Public Service Electric & Gas Company 

Pressurized-water Reactor 

Quality Assurance 

Qual i ty Control Devi at i on or Not i f i cat i on 

Relaxation of Constant Axial Offset Control 

Rod Accountability Monitoring System 

Rod C1 uster Control Assembly 

Rod Ejection Accident 

Reload Initialization Questionnaire 

Request/Probl em Report 

Reload Safety Analysis Checkl ist 

Relcad Safety Evaluation Report 

Relcad Schedule and Energy Requirements 

Re1 oad Safety and Licensing Checkl i st 

Removable Top Nozzle 

Safety Analysis Standards 

South Carolina Electric and Gas Company 

Safety Evaluation Report 

CNFD Speci a1 ty Metal s P1 ant 

South Texas Project 

Thermal -Hydraul i cs 

Total Indicator Reading 

Special Inspection Branch 

Tube Reduced Extrusion 

Uranium Hexafluoride 

Uranium Dioxide 

U1 trasonic Testing 

Vendor Inspection Section 

Westinghouse Electric Corporation 

- 5-2 - 

21 2 
APPENDIX B Continued 

UABA 

WEPC Wisconsin Electric Power Company 

w/o Weight Percent 

wz CNFD Western Zirconium Plant 

Zr 

ZrB, Zirconium Diboride 

Xe Xenon 

Wet Annul ar Burnable Absorber 

Z i rcon i um 
Selected Generic Correspondence on the Adequacy of 

Vendor Audits and the Quality of Vendor Products 

Ident i f i er 

Information Notice 95-32 

Information Notice 95-34 

Title 

Thermo-Lag 330-1 F1 ame Spread Test Results 

Air Acturator and supply Sir Regulator Problems 

in Copes-Vu1 can Pressurizer Power-Operated 

Re1 i ef Valves 

214 
URC FORM 335 

2-59) 

JRCM 1102 

US. NUCLEAR REGULATORY COMMISSION 1 REPORT NUMBER 

/Adend bv NRC. Add Vol., Supp., Rev., and Addendum Numbers, if any.) 

120f.3202 ’ BIL BLIOGRAPHIC naTa I-.- fiMFFT -..--. 

(See instructions on the reverse) NUREG-0040 

Vol. 19, No. 3 

).TITLE AND SUETtTtE 

Licensee Contractor and Vendor Inspection 

Status Report 

Quarterly Report 

July - September 1995 

3. DATE REPORT PUBLISHED 

MONTH YEAR 

er 1995 

4. FIN OR GRANT NUMBER 

I. AUTHORW 6. TYPE OF REPORT 

Quarter 1 y 

7. PERIOD COVERED (lndusive Dates/ 

July - 

September 1995 

i. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Officeor Region, U.S. Nuclear Regulatory Commission, andmai1ingaddress;ifconrractor. provide 

name and mailing address1 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

U.S. Nuclear Regulatory Commission 

Washington, D. C. 20555-0001 

). SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, type “’Same as above’? if contractor, provide NRC Division, Office or Region, U.S. Nuclear Regulatory Commission, 

and mailing address.) 

Same as above 

0. SUPPLEMENTARY NOTES 

1. ABSTRACT 1200 words or Iessl 

This periodical covers the results of inspections performed by the NRC’s Special 

Inspection Branch, Vendor Inspection Section, that have been distributed to the 

inspected organizations during the period from July through September 1995. 

2. KEY WORDS/DESCR!PTORS (List words orphrases that wilIassist researchers in Iocating therepart,) 

Vendor Inspection 

IRC FORM 335 (2.89) 

13. AVAILABILITY STATEMENT 

Unlimited 

14. SECURITY CLASSIFICATION 

(This Pwel 

Unclassified 

Unclassified 

(This Report1 

15. NUMBER OF PAGES 

16. PRICE