NUREG-0040 

Vol. 21, No. 2 

Licensee Contractor 

and Vendor Inspection 

Status Report 

Quarterly Report 

April = June 1997 

8 &#p ,-- 

U.S. Nuclear Regulatory Commission 

Office of Nuclear Reactor Regulation 
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document. 
NUREG-0040 

VOl. 21, No. 2 

Licensee Contractor 

and Vendor Inspection 

Status Report 

Quarterly Report 

April - June 1997 

Manuscript Completed: October 1997 

Date Published: November 1997 

Division of Reactor Controls and Human Factors 

Ofice of Nuclear Reactor Regulation 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 
NUREG0040, Vol. 21, No. 2 has been 

reproduced from the best available copy. 
ABSTRACT 

This periodical covers the results of inspections performed between April 1997 and June 1997 

by the NRC's Special Inspection Branch, Vendor Inspection Section, that have been distributed 

to the inspected organizations. 

... 

Ill 

CONTENTS 

PAGE 

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

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

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

AB6 Service incorporated 

Regional Service Center 

Fenton, MO 

C&D Charter Power Systems, Inc. 

Conshohocken, PA 

Dubose National Energy Services, Inc. 

Clinton, NC 

General Electric Nuclear Energy 

Wilmington, NC 

SynTech Products Corporation 

Toledo, Ohio 

Nuclear Advanced Technology Division 

Westinghouse Electric Corporation 

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

(99901305/97-01) ....................... 7 

(99900861/97-01) ....................... 21 

(99900003/97-01) ....................... 42 

(99901313/97-01) ....................... 58 

(99900404/97-01) ....................... 74 

Select Generic Correspondence on the Adequacy of Vendor ...................................... 88 

Audits and the Quality of Vendor Products 

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. The Federal government and nuclear industry 

have established a system 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 Federal Regulations). 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 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 NRC reviews and inspects nuclear steam system suppliers (NSSSs), architect engineering 

(AE) firms, suppliers of products and services, independent testing laboratories performing 

equipment qualification tests, and holders of NRC construction permits and operating licenses 

in vendor-related areas. These inspections are done to ensure that the root causes of reported 

vendor-related problems are determined and appropriate corrective actions are developed. The 

inspections also review vendors to verify conformance with applicable NRC and industry quality 

requirements, to verify oversight of their vendors, and coordination between licensees and 

vendors. 

The NRC does inspections to verify the quality and suitability of vendor products, licenseevendor interface, environmental qualification of equipment, and review of equipment problems 

found during operation and their corrective action. When nonconformances with NRC 

requirements and regulations are found, the inspected organization is required to take 

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

When generic implications are found, NRC ensures that affected licensees are informed 

through vendor reporting or by NRC generic correspondence such as information notices and 

bulletins. 
This quarterly report contains copies of all vendor inspection reports issued during the calendar 

quarter for which it is published. Each vendor inspection report lists the nuclear facilities 

inspected. This information will also alert affected regional offices to any significant problem 

areas that may require special attention. This report also lists selected bulletins, generic letters, 

and information notices, and include copies of other pertinent correspondence involving vendor 

issues. 

viii 
INSPECTION REPORTS 
UNITED STATES 

NUCLEAR REGULATOSV COMMISSION 

WASHINGTON, D.C. -1 

Mr. D. P. Polizzi 

Manager, Regional Service 

ABB Service Inc. 

Regional Service Center 

400 Biltmore, Suite 520 

Fenton, MO 63026 

June 4, 1997 

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

Dear Mr. Polizzi: 

This letter refers to the inspection conducted by Mr. K. R. Naidu and Mr. V. 

L. Beaston of this office on May 15, 1997. The inspection included a review 

of activities conducted at your facility in Fenton, Missouri. 

report presents the results of that inspection. 

The enclosed 

The inspectors evaluated the program that ABB Service Inc. (ABB) established 

to implement the provisions of Part 21 of Title 10 of the Code of Federal 

Requlations (10 CFR Part 21) for reporting defects and noncompliances, and 

Appendix B to 10 CFR Part 50. The inspectors reviewed the documentation on 

selected metal-clad K-line circuit breakers that Illinois Power Company's 

Clinton Power Station sent to your Fenton shop for refurbishing. 

areas, the inspection consisted of an examination of procedures and 

representative records, interviews with personnel, and observations by the 

inspectors. The inspectors identified one unresolved item relative to the 

availability of breaker specific drawings to verify the correct contact 

configuration and propose to obtain additional information on this item during 

a future inspection at your Houston Service Center. 

Within these 

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 

Speci a1 Inspection Branch 

Division of Inspection and Support Programs 

Docket No.: 99901321 

Enclosure: Inspection Report 99901321/97-01 

-2- 
INSPECTION REPORT 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF INSPECTION AND SUPPORT PROGRAMS 

ORGAN I ZAT I ON : 

REPORT NO.: 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY: 

INSPECTION CONDUCTED: 

INSPECTORS: 

APPROVED BY: 

ABB Service Inc. 

Fenton, Missouri 

99901321/97-01 

Mr. D. P. Polizzi, Manager, Regional Service 

314 343 0232 

ABB Service Inc. services switchgear 

manufactured by I.T.E. 

May 15, 1997 

Kamalakar R. Naidu, NRR 

Virgil L. Beaston, NRR 

Gregory C. Cwalina, Chief 

Vendor Inspection Section 

Special Inspection Branch 

Division of Inspection and Support Programs 

Off i ce of Nucl ear Reactor Regul at i on 

-3- 
1 i WECT ION SUMMARY 

ABB Service Inc. , located in Fenton, Missouri , performs switchgear overhaul ing 

services for commercial and nuclear customers. Recently it performed services 

on 480 volt metal-clad circuit breakers installed at Illinois Power Company's 

(IPC) C1 inton Power Station (Clinton). 

designed and manufactured by I.T?E Imperial Company (ITE). The ownership 

subsequently changed and the company became known as 1.T.E.- Gould, GouldBrown Boveri, Brown Boveri Electric and finally ASEA Brown Boveri (ABB). ABB 

manufactures low-voltage metal-clad circuit breakers, and medium-vol tage 

circuit breakers at Florence, South Carolina. ABB at Sanford, Florida, 

assembles complete switchgear installations. 

centers at several locations in the U.S. The service center at Fenton 

currently services the circuit breakers installed in Clinton. 

decided to service its nuclear customers from four locations, Columbia, 

Maryland; Cleveland, Ohio; Charlotte, North Carolina; and Houston, Texas. In 

the future, breakers which need service will be shipped to ABB Services, 

Houston , Texas. 

The circuit breakers were originally 

ABB has established service 

Recently, ABB 

The inspectors reviewed the documentation on selected metal-clad K-line 

circuit breakers that IPC's Clinton sent for refurbishment. 

The inspection bases were: 

e 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 Federia 

Requlations (10 CFR Part 50) 

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

During this inspection, the NRC found no instance where ABB failed to meet NliC 

requirements. However, the inspectors identified one unresolved item 

regarding the availability of breaker specific drawings (Section 3.1). 

2 STATUS OF PREVIOUS 

This is the first inspect 

3 INSPECTION DETAILS 

NSPECTION FIND 

on at ABB Serv 

3.1 Breaker Refurbishment 

a. ScoDe 

On May 15, 1997, the inspect rs reviewed 

NGS 

ce Inc., Fenton, Missouri. 

g the 

refurbishment performed on C1 inton breakers and reviewed the refurbishment 

the records documenti 

activities with the personnel who performed the work at the Fenton shop. 

The inspectors also reviewed the purchase order (PO) that IPC issued to ABB 

for the refurbishing services including selected revisions to the PO to 

determine if the work preformed on selected Clinton breakers met the quality 

requirements. 

1 

-4- 
b. Observations and Findinas 

IPC issued PO 554794 to AB6 Sanford, Florida, to reworkpefurbish a circuit 

breaker with serial number 51752A-9-02616 with SS-13 type solid state trip 

device in compl i ance with ABB Procedure MS 3.1.1. 9-2D, "Maintenance and 

Survei 11 ance, Low Voltage Swi tchgear Equipment. 'I In subsequent revisions to. 

the PO, Clinton added additional breakers to be refurbished. 

IPC attached to the PO a copy of its "Supplier Quality Assurance Program 

Requirements," QAP-407-02F02, Revision 2, which invoked 10 CFR Part 50 

Appendix 6, and 10 CFR Part 21. IPC required AB6 to obtain its approval for 

all nonconformance reports initiated by ABB during the refurbishment where ABB 

would recommend "use as is" or "repair." 

Certificate of Rework or Repair certifying that the item was reworked using 

original equipment parts by a qualified technician, and that the item will 

function in accordance with the original design specification. 

IPC required ABB to issue a 

Revision 02 to the PO, dated June 23, 1995, required the replacement of all EQ 

sensitive (organic) items with less than 40-year qualified life. As a 

minimum, Agastat time delay relays, Gould auxiliary relays (J13PA4312) , and 

control and power Shawmut type fuses were to be replaced. 

Revision 09 to the PO, dated November 29, 1995, required ABB to test the 

breakers for "as-found" conditions. 

lubricate the breakers, and test the trip devices to verify the time/current 

curve. 

ABB was also to rework/refurbish and 

Revision 12 to the PO, dated March 12, 1996, added the refurbishment of 

breaker Serial Number (S/N) 51752E-9-02616. ABB was to (1) reconfigure the 

breaker and update documentation as appropriate, (2) replace the electrically 

operated mechanism with a manually operated mechanism, (3) replace the lower 

molding and CT assembly with 600-A variety, (4) change the solid state trip 

device to 600-A (SS 14), (5) add an alarm switch, and other minor accessories 

as required to conform with Drawing 708 663, Revision 0. 

The records on work performed on breakers with S/Ns 51752AB-109-201054, 

51752B-9-11456, 51752E-9-02616 , 51752B-9-01456 , 51752F-11-02616 , and 517525-9- 

01616, indicate that ABB personnel had operated the C1 inton breakers, 

disconnected the breaker mechanisms, and inspected the main roller and carrier 

and other major pins and moving components in the breakers. Records indicated 

that the operating mechanisms were dirty and had to be disassembled, cleaned, 

lubricated and reassembled. On receipt of the breakers from ABBY Clinton 

personnel performed receipt inspections and documented adverse findings 

(discussed below) in Condition Reports (CRs). The inspectors discussed the 

adverse findings in these CRs with the ABB personnel and reviewed the 

refurbishment records. 

During receipt inspections, Clinton identified a deficiency related to 

incorrect wiring of leads 11 and 12 on the alarm switch in four breakers. 

Clinton corrected this error by switching the leads, and informed ABB during a 

meeting at the Clinton plant. The inspectors interviewed the ABB personnel 

2 

-5- 
who worked on the breakers, and determined that the technician who kired the 

switch had made an error. 

correct wiring of the alarm switch because Clinton did not send them the 

drawings for the individual breakers. At the time of the inspection, ABB 

personnel stated that this condition had not been documented in a nonconformance report because the technician who made the error had been away 

working on a job in Maine, and returned to the shop only the previous week. 

ABB personnel subsequently documented this in a nonconformance report. 

ABB personnel stated that they could not verify the 

Clinton also identified a problem related to the configuration of the 8-point 

auxiliary switch which is mounted on the 'IC" phase side below the Power Shield 

unit. The switch is connected through a right-hand link assembly to the jack 

shaft of the breaker. The contacts in the auxiliary switch change state when 

the breaker operates. Clinton personnel had documented that, when the 

breakers were receipt inspected, the contact configuration did not meet the 

breaker drawing requirement. ABB personnel stated that different breakers 

have di fferent contact configurations. A 90-degree rotation of the operating 

lever will change the contact configuration. ABB personnel informed the 

inspectors that Clinton did not provide them with the drawings for the 

individual breakers. Therfore, ABB could not verify the correct contact 

configuration. 

perform safety-re1 ated refurbishments. 

Further, ABB did iiot identify the need for the drawings to 

c. Conclusion 

Initially, Clinton did not provide individual breaker drawings to ABBY and ABB 

did not request them. 

inadequate Licensee - Vendor interface. Adequate technical information 

exchange (e .g. , circuit breaker drawings) between personnel at C1 i nton and ABB 

could have reduced the number of problems that were identified in the 

condition reports. Since the service center at Fenton will no longer 

refurbish safety-related breakers, this concern is being identified as an 

unresolved item and will be followed up at a future inspection of the ABB 

Service Center at Houston, where ABB proposes to refurbish the Clinton 

breakers in future. 

The inspectors considered this to be an indication of 

(Unresolved i tem 99901321/97-01-01) 

4. PERSONS CONTACTED 

ABB Service ComDany 

D. P. Polizzi, Manager, Regional Service 

R. C1 ostermann , Service Engineer 

J. 0. Webb, Director of Qual i ty 

ITEMS OPENED, CLOSED, AND DISCUSSED 

ODened 

99901321/97-01-01 URI unavailability of breaker drawings 

3 

-6- 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-0001 

May 16, 1397 

Dr. Lesl ie S. Holden 

Vice President Techno1 ogy 

C&D Charter Power Systems, Inc. 

Washington & Cherry Streets 

Conshohocken, PA 19428 

SUBJECT: NRC INSPECTION REPORT 99901305/97-01 

Dear Dr. Holden: 

On January 9, 1997, thy U.S. Nuclear Regulatory Commission (NRC) completed an 

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

Leola, Pennsylvania, regarding cy1 indrical lead-acid battery cells, "round 

cells," that C&D manufactures and supplies to nuclear power plant facilities 

as a contractual manufactyping representative for Lucent Technologies, 

Incorporated, (Lucent). The round cells are designed, qualified and 

contractually controlled by Lucent, and are used 'r! Class 1E electrical 

safety-related applications at five nuclear power plant facilities. 

enclosed report presents the results of the inspection. 

The 

The NRC team determined that C&D has manufactured and controlled the nuclear 

power plant customer orders under its commercial manufacturing controls, in 

association with Lucent's specifications. 

an International Organization for Standardization (IS0)-9001, "Quality 

Systems-Model for Qual i ty Assurance in Design/Devel opment , Production, 

Installation and Servicing" quality assurance program at its Leola facility 

through its corporate office in Blue Bell, Pennsylvania. Although the team 

assessed portions of your quality assurance program implementation, your 

activities were not assessed to Part 50 of Title 10 of the Code of Federal 

Reaulations (10 CFR Part 50), Appendix B quality assurance requirements; 

therefore, no response to this letter is required. 

C&D has established and implemented 

However, the team determined that Lucent contractually accepted 10 CFR Part 

50, Appendix B quality requirements from the Duke Power Company but did not 

pass those requirements on to C&D. 

portions of C&D's quality activities to Appendix B quality requirements to 

assess Lucent's compliance with regulatory requirements. 

review are discussed separately in NRC's inspection report of Lucent 

Technologies, Report 99901309/96-01. 

Therefore, the team reviewed certain 

The results of that 

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

-7- 
Dr. L. S. Holden -2- 

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

letter and enclosures will be placed in the NRC's Pub1 ic Document Room (PDR) . Should you have any questions concerning the issues discussed in this letter, 

we will be pleased to discuss them with you. 

Sincerely, 

n 

Robert M. Gallo, Chief 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Docket No. : 99901305 

Enclosure: Inspect ion Report 99901305/97-01 

cc: Mr. T.J. Kinden, Director 

Quality Assurance 

C&D Charter Power Systems, Inc. 

Washington & Cherry Streets 

Conshohocken, PA 19428 

Mr. Greg Stoermer, Plant Manager 

C&D PowerCom 

82 East Main Street 

Leo1 a, PA. 17540- 1940 

-8- 
Report No. : 

Organization: 

Contact Address: 

Faci 1 i ty 

Inspected: 

Nuclear Industry 

Activity at 

Facility: 

Inspection Date: 

Inspector: : 

Approved by: 

Enclosure 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REI\CTOR REGULATION 

99901305/97-01 

C&P' Charter Power Systems, Incorporated 

1400 Union Meeting Road 

B1 ue Bel 1, Pennsylvania 19422-0858 

Terry J. Kinden, Director of Quality Assurance 

Washington & Cherry Streets 

Conshohocken, PA 19428 

(610) 825-2150 ext. 245 

C&D Powercorn' 

82 East Main Street 

Leo1 a, Pennsyl vani a 

Manufacturers of Lucent Technologies cylindrical 

lead-acid battery cells, Lineaged 2000 Round Cell 

batteries (round cells), that are supplied to nuclear 

power plant facilities for Class 1E electrical 

safety-related applications. 

January 6-9, 1997 

Joseph J. Petrosino, Team Leader 

Saba N. Saba, EELB:DE:NRR 

Stephen D. Alexander, V1S:PSIB:DISP 

Gregory C. Cwalina, Chief 

Vendor Inspection Section 

Speci a1 Inspection Branch 

Division of Inspection and Support Programs 

Office of Nucl ear Reactor Regul at i on 

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

A division of C&D Charter Power Systems, Inc. 

Registered trademark of Lucent technologies, Incorporated (formerly 

known as American Telephone and Telegrhph Company [AT&T]). 

-9- 
1 INSPECTION SUMMARY 

During this inspection, the NRC inspectors reviewed the implementation of 

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

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

round cells, reviewed pertinent design, engineering and manufacturing 

documents, reviewed the Lucent Technologies, Incorporated (Lucent) 

manufacturer requirements regarding Lucent's control of C&D's Leola 

manufacturing activities, and compared thc Lucent requirements kith actual 

manufacturing procedures and practices. 

nuclear pl ant round cell batteries experienced premature 1 oss of capacity. 

The team was aware that several 

The inspection bases were: 

Appendix 8, "Quality Assurance Criteria for Nuclear Power Plants and 

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

Requlations (10 CFR Part 50) 

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

The American National Standard InstitctejThe American Society of 

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

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

endorsed by U. S. Nuclear Regulatory Commission Regulatory Guide 

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

Construction)," Revision 2, February 1979 (RegGuide 1.28) 

Although C&D's quality program for its manufacture of round cells was found to 

be generally well established and implemented, the team found some weaknesses 

that are discussed in this report. Those weaknesses are also discussed in NRC 

Inspection Report 99901309/96-01, Lucent Techno1 ogies-Dal 1 as, Texaz. 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

This was tie first NRC inspection of C&D's facilities at Leola, Pennsylvania. 

3 INSPECTION FINDINGS AND OTHER COMMENTS 

3.1 Qualitv Assurance Proqram and Reauirements 

a. ScoDe 

The inspection team (team) reviewed the establishment and implementation 

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

the C&D Quality System Program Manual (QSPM), Issue P, dated June 1995. 

b. Observations and findinas 

The QSPM stated that its purpose was to establish the basic operating 

pol icies and procedures to be employed by C&D, and to meet appl icable 

requirements of International Organization for Standardization 

(IS0)-9001, "Quality Systems-Model for Quality Assurance in 

2 

-10- 
C. 

3.2 

a, 

b. 

Des ign/Devel opment , Production, Ins t a1 1 at i on and Serv i c i ng , 'I 

ANSI ENl45.2-1977 and other imposed specifications. The team 

noted that C&D's Leola facility had been certified to ISO-9001. 

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

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

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

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

QSPM, departmental standard operating procedures, work instructions, 

drawings, bills of materials, material specifications, process 

specifications, and quality and test procedures. The team reviewed 

applicable licensee and Lucent procurement documents and did not find 

any instances where AT&T/Lucent had imposed any unique nuclear quality 

assurance requirements upon C&D. 

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

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

Headquarters-Blue Bell, PA-Control and maintenance of design and 

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

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

chargers; (4) Conshohocken, PA-seismically q!Jalified, Class 1E station 

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

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

products ; and (8) Ratel co , Incorporated , Seat t 1 e, WA-el ect ron i cs , including battery charger assemblies used in safety-related appl ications 

of C&D battery chargers. 

Concl usions 

The team determined that C&D-Leola manufactured the round cells 

according to its commercial QA program. No nuclear qualification, 

manufacturing or testing requirements were imposed onto C&D by Lucent 

for the manufacture of the round cells. 

10 CFR Part 21 Prosram 

ScoDe 

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

requirements of 10 CFR Part 21 by reviewing Standard Policy and 

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

accordance with Federal Regulation 10 CFR 21 (US Nuclear Regulatory 

Commission)," dated March 18, 1997. 

Observations and findinqs 

Although C&D is not responsible for the requirements of 10 CFR Part 21 

at its Leola facility, the team determined that C&D had implemented the 

provisions of Procedure A-14-4 into its Leola facility round cell 

3 

-1 1- 
C. 

3.3 

a. 

b. 

operation for nuclear power plant orders. Additionally, the team noted 

that documents required by Part 21 had been posted at various locations 

at the Leola facility. 

During a previous inspection, in September 1996, at C&D’s Attica, 

Indiana, facility, the team performed a review of Procedure A-14-4. The 

review determined that some weaknesses existed in the procedure and C&D 

committed to modify the procedure to address the NRC concerns within 90 

days of the date of Inspection Report 99901304/96-01, (March 17, 1997). 

Procedure A-14-4 had not yet been revised and provided to the C&D-Leola 

facility. 

Concl usi ons 

The team did not identify any concerns in this area. 

ImDlementation and Control Of Quality Prosram 

ScoDe 

During the inspection at the C&D-Lrola facility, the team observed 

manufacturing activities and reviewed associated documents to assess the 

adequacy of the manufacture of the round cells. 

Observations and Findinqs 

Process Controls The team observed numerous manufacturing process 

control activities at C&D’s facility and reviewed associated documents. 

The team noted that many of the C&D manufacturing work stations had key 

sections of the applicable process control work instructions posted for 

use of the area personnel. The team also noted that a Lucent required 

resident inspector is stationed at the C&D facility. 

inspector is typically on the premises several days a week. 

resident inspector witnesses testing activities of round cell battery 

strings, and performs in-process manufacturing verifications and random 

inspection of incoming material. No anomal ies were noted by the team in 

the manufacturing process control activities that were observed. 

The resident 

The 

Charqinq & Conditioninq The team performed a comparison of the C&D 

procedures, practices and documentation used during the process of 

initial charging and conditioning of round cell strings with the 

charging and conditioning requirements (design bases) of Lucent’s design 

and manufacturing specification KS-20472 and associated documents. 

team reviewed the charging records for string 345-96 for McGuire Station 

and strings 223-92 and 224-92 for Oyster Creek. 

The 

C&D conducts all of its manufacturing activities (including charging and 

conditioning) in the same manner, in accordance with its documented ISO9001 quality system in conjunction with Lucent’s design and 

specification documents. However, the team identified several instances 

4 

-12- 
in which C&D's procedures, practices and documentation for the charging 

and conditioning process departed from the requirements of the Lucent 

design specification. The following are examples of these departures: 

KS-20472, Section 5.25, Step (c), prescribed the initial formation 

charge for the string. The step included a cautionary note that 

stated that the cell tempe;ature should not be allowed to exceed a 

certain value. 

monitored, Liiey were not recorded, nor was there a provision in the 

C&D record forms to do so. 

Although C&D stated that the teniperatures are 

KS-20472, Section 5.25, Step (d), prescribed the first conditioning 

discharge following the initial formation charge, per Table I1 for 

List 1s round cells, to Sr : certain amperage for a certain 

duration. This discharge removed a specific amount of ampere (ah) 

from each cell of the string. 

individual cell should be allowed to go below [specified] volts." 

KS-20472, Section 5.25, Step (e) required that, following the 

Discharge, the string be recharged to a specific percentage of the 

ampere-hours removed during the discharge for a minimum time 

duration. This charge then should have restored a minimum and a 

maxi mum Val ue . 

The step included the warning "NO 

However, tne overall charging record sheet, cal led "Procedure 

Sheet 1," was pre-printed with the annotation, "[specified number] 

amps for [certain number of] hours" for this charge. This charge 

would put in more ampere-hours than allowed by the specification. 

Although C&D produced a 1986 memorandum from Lucent (then AT&T/Bell 

Laboratory) authorizing this departure from KS-20472, the 

specification had not been updated, in over four years, to reflect 

this modification to the prescribed charging and conditioning (C&C) 

process. The charge records reflected that the string was, in fact, 

recharged per Procedure Sheet 1 instead of KS-20472. 

KS-20472, Section 5.25, Step (f), prescribed a "First Taper 

Discharge," per Table 111 for List 1s round cells, to be for a 

Specific periods at certain current values to remove a specific 

number of ampere-hours. The first post-taper recharge per Step (9) 

was supposed to ensure that a certain percentage of the ampere-hours 

were removed in the previous discharge. However, Procedure Sheet 1 

was pre-printed with the annotation, "[specified value] amps for 

[certain number of] hours" followed by "[specified value of] amps 

for [certain number of] hours," more than allowed by the 

specification. Although C&D produced another 1986 memorandum from 

Lucent (then AT&T/Bell Laboratory) authorizing this departure from 

KS-20472, the specification did not reflect this four-year old 

design basis change to the prescribed charging and conditioning 

(C&C) process. The charge records reflected that the string was 

recharged per Procedure Sheet 1 instead of KS-20472. 

5 

-13- 
KS-20472, Section 5.25, Steps (h) and (j), prescribed the required 

sequential "[specified value]-Amp Discharges." The steps stated 

"Discharge at the Table IV rate until the first cell reaches 

[specified value] volts." However, C&D 01-20-0900.2, "Charging" 

(Section 6.3.16, "Water Discharges") , did not provide for normal 

discharge termination upon reaching a prescribed voltage (on the 

first cell to reach it). Instead, Step 6.3.16.3 called for 

termination "at the,normal discharge end time." The note above step 

6.3.16.3 stated: "If any cell has a voltage of [certain number of] 

volts or less, terminate the discharge and notify your supervisor." 

Therefore, the C&D procedure called for discharging until an end 

time that is not defined as the termination point for this 

particular discharge in KS-20472. Also, the C&D procedure 

prohibited allowing any cell to go below a certain voltage, but did 

not require discharging until reaching a specified voltage on the 

first cell that reaches this voltage as the language of the Lucent 

specification states. 

In view of the explicit language in Step (h) of the specification 

(as distinguished from language explicitly prohibiting discharging 

below a specified voltage elsewh?:.? in the specification), the 

language in the C&D 01 was contrarL, +c) the requirements of 

specification KS-20472. As a result, the charge records showed that 

both specified discharges went only to a certain voltage on the 

lowest cells. Although the practice of discharging until the first 

cell reached the specified voltage (instead of the prescribed 

voltage) had been authorized in 1986 on a limited, experimental 

basis, C&D could not provide documentation authorizing this 

departure from KS-20472 for general production use. This resulted 

in the string not being discharged in this step to the extent 

required by KS-20472. 

KS-20472, Section 5.25, Step (i), prescribed the recharges for the 

two sequent i a1 I' [ speci f i ed val ue] -Amp Di scharges" of Steps (h) and 

(j) respectively. The recharges rJere supposed to be a specified 

percentage of the ampere-hours removed in the previous discharge. 

Since both discharges continued for several hours, they were 

discharged to a certain value. Therefore, according to the KS-20472 

formula, a specified minimum of the ampere-hours should have been 

recharged. However, Procedure Sheet 1 was pre-printed with the 

annotation, "[specified value of] amps for [certain number of] 

hours," less than required by the KS-20472 formula using the 

abbreviated discharges. 

KS-20472 because the recharge amount, even if calculated and 

performed in accordance with the KS-20472 formula, would have been 

based on a discharge to a nominal voltage, less than the amount of 

discharge specified by KS-20472 in Steps (h) and (j). 

This was also much less than intended by 

KS-20472, Section 5.25, Step (k), prescribed a "Second Taper," or 

"[specified value]-amp Taper" discharge per Table V for List 1s 

round cells, to be conducted in several steps at specific values. 

Step (1) required discharge to "[a specified value of] volts per 

6 

-14- 
cell average," or for a specified ampere-hour value with the caution 

that no cell should be allowed to go below a certain voltage. 

However the discharge record (Form RS533) showed. that no cell went 

below the minimum allowed voltage during the phase and none reached 

the specified voltage during any part of the second taper discharge. 

A1 though the minimum total ampere-hours were discharged, no cell 

achieved the lowest voltage specified by KS-20472. 

The final capacity testing specified by KS-20472 following chzrging 

and conditioning was modified for Strings 223-92 and 224-92 by a 

procedure required by Lucent's customer, GPU Nuclear, 1 icensee for 

Oyster Creek. GPU Nuclear had contracted National Technical Systems 

(NTS), Inc., of Acton, Massachusetts, to perform the dedication of 

the round cell strings for Oyster Creek purchased from Lucent as 

commercial grade items. NTS Procedure 60145-93N, Revision 2, was 

used as the basis for dedication of the strings. 

procedure modified the standard capacity test discharge rate 

following the specified float period prescribed in KS-20472. 

Instead, the NTS procedure called for a high-current discharge, 

recharge, and a several hour discharge, intended to better simulate 

design basis service conditions at Oyster Creek. The test record 

contained a notice of anomaly indicating that only 95-percent 

capacity had been demonstrated by the discharge tests, but that this 

was acceptable. The explanation of the reduced capacity was the 

interruption of the standard "conditioning" sequence to perform the 

NTS dedication tests. However, this language did not accurately 

characterize the circumstances in that the conditioning phase of the 

KS-20472 process, including the final several-day float charge, was, 

in fact, completed. Only the final capacity test discharge, not 

considered part of conaitioning, was omitted in favor of the NTS 

mu1 ti pl e hour di scharges . 

In reviewing the test and cell selection records for strings 223-92 

and 224-92, the inspector noted that one test for 223-92 was 

terminated several minutes early. A notice of anomaly (NOA) also 

noted the early stop, but had the wrong time. In addition to the 

records having a number of other anomalies, such as not all forms 

being fully or correctly filled out and some erroneous operation 

ending times recorded. For example, one recharge for a specific 

discharge was listed as staring at a certain value on 11/21/92, but 

ending at the same value on 11/22/92, a longer period than specified 

on Procedure Sheet 1; yet, the second discharge was recorded as 

starting earlier on 11/22, and the inspector believed this entry to 

likely be the correct one. The charge end time entry of the first 

entry on 11/22 was apparently in error, but there was no indication 

that the record had been reviewed and the error identified by C&D 

QA, engineering or management or representatives of Lucent or GPU 

Nuclear. 

The NTS dedication 
The inspector also noted that the cell in position 43 in string 223- 

92 (cell serial no. 78471) finished the test with a low-out-ofspecification voltage, yet this cell was selected for shipment by 

GPU Nuclear according to the selection record. Conversely, although 

the cell in position 64 had a low voltage, but met the acceptance 

criterion, it was rejected. The record indicated no rationale for 

thz apparently inappropriate cell selection, nor could C&D offer any 

expl anat i on. 

The inspector also reviewed the charging and conditioning records 

for round cell string 345-96, produced for Lucent’s customer, Duke 

Power for Class 1E service at McGuire Station. The same process was 

used for this string as for the Oyster Creek strings discussed above 

except that the standard final capacity test discharge was 

conducted. Therefore, this string underwent charging and 

conditioning with the same deviations from KS-20472 as occurred with 

the Oyster Creek strings, only some of which, as above, had ever 

been authorized by Lucent memorandum, and still without formal 

revision to KS-20472 (Issue 8 still in effect). 

Chemical Analysis of Epoxy Mixture The team reviewed the applicable 

section of KS-20472 regarding preparation and mixture verification of 

epoxy used to seal the round cell jars around the terminal posts. Also 

reviewed were the Lucent epoxy titration procedure referenced by 

KS-20472 that was prescribed on AT&T Drawing L193472, Issue 5, and the 

C&D epoxy titration procedure, BT-4, Revision 0, dated March 7, 1995. 

The procedures specify the quantitative analysis by titration of percent 

resin in hardener (done on each new mixed batch) and of the equivalent 

weight of hardener (done on each lot of hardener and each shift), the 

results of which are used in the percent-resin-in-hardener analysis 

cal cul at i on. 

Following this review, the team observed a routine set of epoxy 

titration analyses, performed by C&D’s usual analyst in the C&D 

laboratory. The analyses observed were the determination of percent 

hardener in resin, and the determination of the equivalent weight of 

hardener. The team identified two instances in which C&D epoxy 

titration procedure BT-4 deviated from AT&T Drawing L193472, 

spec i f i call y : 

L193472 called for titrating into the sample in a 125-1111 Erlenmeyer 

flask; whereas BT-4 prescribed a 250-1111 flask, and 

L193472 required adding a specific number of drops of Brom Creosol 

green indicator; whereas, BT-4 specified a range of drops that could 

be added. 

The team also determined that the size of the flask can affect the 

efficiency of mixing of the sample (by means of a magnetic stirrer) with 

the indicator, and more importantly, with the drops of titrant (sulfuric: 

acid). The amount of indicator added can affect the visibility of the 

endpoint. Other deficiencies in C&D procedure BT-4 noted were: 

8 

-16- 
The C&D procedure i nappropri ately defined the titration endpoint as 

a color change from blue to ve1;ow; however, the analyst correctly 

described the color change as being from blue to ween. 

The acceptance criteria for percent-hardener-in-resin and for 

hardener equivalent weight given on 1193472 were not in the C&n 

procedure for use by the analyst, although the analyst was familiar 

with them. 

No sample vortex characteristics were specified to ensure uniform 

mixing in the flask. 

L. 

The team observed two instances in which the analyst deviated from C&D 

procedure BT-4: 

Instead of the 250-1111 Erlenmeyer sample flask specified by BT-4, the 

analyst used a 125-ml flask as specified by the Lucent drawing; 

although the analyst was unaware of the drawing requirement. 

explained, correctly, that for the specified 50-ml sample size, the 

smaller flask provided the depth of liquid necessary to achieve an 

adequate vortex (using the specified magnetic stirrer) for proper 

mixing of the indicator and the titrant in the sample. 

He 

The C&D analyst stated that he typically adds more drops of 

indicator than that specified in BT-4, nearly twice as much 

indicator as specified in drawing L193472. 

In addition, the analyst did not employ the standard titration technique 

when near the expected endpoint of so-called "drop splitting" or "drop 

transfer" using a glass rod to take each drop allowed to form at the 

burette nozzle as the titrant level is lowered by the smallest burette 

graduation and placing the drop in the sample. Instead, the analyst 

simply slowed the expected near-endpoint rate of titration to about 

1 drop per second, which may not allow for complete mixing and reaction 

of the titrant with the sample and indicator so as to shoh the endpoint 

when it first occurs before the next drop falls into the sample. This 

practice, with a less experienced analyst, and also using the 250-1111 

flask if BT-4 were followed, could cause the actual endpoint to be 

missed, thus degrading the sensitivity of the analysis or the accuracy 

of the results. 

Audits and Surveillance of C&D-Leola bv AT&T/Lucent A review of the 

AT&T Qual ity Management & Engineering Department (QM&E) audits and 

surveys indicated that AT&T/Lucent performed on-site audits of C&D to 

observe their capabilities and performance as required by Lucent's 

quality manual and technical specifications and drawings. The team 

noted that the QM&E personnel identified numerous quality problems over 

a several year period. 

Appendix By quality assurance requirements which were imposed upon AT&T/Lucent, the team determined that Lucent did not take prompt action 

to address the effectiveness of C&D's control of product quality. That 

is, although Lucent identified numerous examples of conditions adverse 

However, contrary to the 10 CFR Part 50, 

9 

-17- 
to quality over a three year period at the C&D-Leola facility, Lucent 

did not promptly correct those conditions adverse to quality identified 

or assure that C&D addressed the ineffectiveness of its program 

implementation consistent with the importance and complexity of the 

round cell product. It was also notzd that the quality and reliability 

of the product that was shipped to nuclear power plants did not appear 

to be addressed in the documents reviewed by the team. 

Consequently, the team determined that Lucent did not asstire that the 

manufacturing services that it contracted from C&D, for fabricating its 

round cells for at least the McGuire station, consistently conformed to 

the procurement document specifications and requirements during the 

charging and condi ti on i ng manufacturi ng act i vi ti es . 

For example, a November 2, 1994, AT&T Bell Laboratories letter to the 

C&D Leo1 a faci 1 i ty qual i ty control (QC) Manager and associated documents; 

discussed the findings of a September 21, 1994, mini quality audit of 

the round cell manufacturing and quality systems. AT&T's correspondence 

indicated that 15 separate areas were investigated during a period when 

C&D was in the process of increasing production rates of its round 

cells. The correspondence indicated that 11 of the 15 areas audited did 

not conform to AT&T's specification reallirements. 

also indicated that in light of a recent problem encountered with 1992 

series 1SH cells sold to Arizona Public Service (Palo Verde), "it would 

behoove C&D to spend additional time and effort on training new 

employees and i ncreasi ng inspect i on" at those operat i ons where these new 

employees are located. The conclusion reached by the AT&T auditors was 

that the overall round cell quality was not up to usual standards. The 

letter to C&D itemized specific areas of concern and requested C&D's 

corrective action plan. 

A December 19, 1994, AT&T Bell Laboratories Trip report documenting a 

December 14, 1994, follow-up mini quality audit stated that C&D had not 

compiled or composed the required corrective action plan, and the 

overall manufacturing process qualit) was found to be lower than C&D's 

usual standards. The trip report stated that although the observed 

deficiencies were not critical, they indicated ''a general relaxation in 

standards. I' 

The correspondence 

The December 1994 report stated that on December 14, the inspection of 

round cell manufacturing process quality indicated that the 

manufacturing quality standards were continuing to trend downward. 

letter indicated that the observation had been made over a period of 

time, was not specific to any one area, and that C&D's emphasis on 

quality and the commitment to quality have diminished over the last six 

months. 

address both Companies' concerns and to ''refocus on the critical 

importance of quality to this product's continued success." 

The 

The staff engineer recommended that AT&T and C&D meet to 

Three months later on March 15, 1995, another mini quality audit was 

conducted and identified findings in 18 areas at the C&D-Leola facility. 

The overall conclusion reached was that "quality is poor." The AT&T 

10 
staff performing the audit stated that work instructions need to be 

reviewed with both the operators and C&D quality control inspectors and 

the importance of producing good quality parts needs to be reinforced. 

c. Conclusions 

No anomalies were noted by the team in the manufacturing process control 

activities that were observed during the performance of this inspection. 

However, in the charging and conditioning process area, the team 

concluded that: 

C&D charging and conditioning procedures and practices deviated from 

KS-20472 and only some of the departures were authorized by Lucent 

which had not updated its specification in several years to reflect 

the different practices 

C&D procedures were not always followed and records were not always 

fully accurate, nor apparently carefully scrutinized by Lucent, the 

ultimate user or its representatives, and 

the 64 cells selected by GPU Nuclear fro9 each of the 72-cell 

strings produced and tested, did not all meet the acceptance 

criteria following the final test discharge. 

In the chemical analysis of epoxy mixture area, the team concluded that 

the anomalies from the AT&T/Lucent drawing as well as the quantitative 

analysis practices employed could affect the accuracy and repeatability 

of the epoxy titration analyses for percent-resin-in-hardener and 

equivalent weight of hardener. 

Additionally, although there is a wide range in the acceptance criteria 

for the epoxy titration analyses, the team was given the understanding 

that neither Lucent nor its customers had identified any of the 

anomalies observed by the team in the titration process. 

effects of these anomalies had not been evaluated by Lucent or their 

customers to determine their significance. 

Therefore, the 

The team concluded that the AT&T/Lucent auditors identified negative 

quality aspects that would be indicative of an ineffectively established 

or implemented quality program. However, the team noted that neither 

Lucent nor C&D took appropriate measures to determine the cause of the 

problem and implement corrective action. 

Additionally, neither the product or process appeared to be 

appropriately questioned regarding quality and reliability of the 

product. 

Lucent’s Dallas, Texas facility, Inspection Report 99901309/96-01-02, 

These issues are also discussed in NRC’s inspection results of 

11 

-19- 
3.4 

3.5 

Entrance & Exit Meeting 

In the entrance meeting on January 6, 1997, the NRC Team Leader 

discussed the scope of the inspection, outlined the areas to be 

inspected, and established interfaces with C&D management. In the exit 

meeting on January 9, 1997, the inspectors discussed their findings and 

concerx. 

PERSONS CONTACTED 

C&D Charter Power Systems. Inc. 

G. Stoermer, P1 ant Manager-Leo1 a 

T. Kinden, Director, Quality Assurance 

D. Johnson, Q.C. Manager 

Lucent Technoloqies-Bell Laboratories, Murray Hill, New Jersey Staff 

M.C. Weeks, Technical Staff 

J.B. Baldasty, Technical Staff 

12 

-20- 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 205550001 

April 29, 1997 

Mr. James Dailey, Manager, 

Quality Assurance 

DuBose National Energy Services, Inc. 

900 Industrial Drive 

Clinton, NC 28328 

Subject: NRC INSPECTION REPORT 99900861 /97-01 AND NOTICE OF 

NONCONFORMANCE 

Dear Mr. Dailey: 

On January 24, 1997, the U.S. Nuclear Regulatory Commission (NRC) completed 

an inspection of DuBose National Energy Services, Inc. The enclosed report 

presents the results of that inspection. 

During this inspection, the NRC inspectors found that tttz implementatioh of your 

quality assurance program failed to meet certain NRC requirements imposed on 

DuBose by customer?. Specifically, the inspection identified instances where 

DuBose accepted material from qualified suppliers without receiving all 

documentation reqliired by the applicable ASME Code or your purchase order 

provisions. In several instances DuBose also supplied material under your ASME 

Quality Systems Certificate without providing all documentation required by the 

applicable ASME NCA-3800 quality assurance program. Additionally, the 

inspection identified that DuBose supplier audits did not consistently identify 

objective evidence to support the audit conclusions, and the audit conclusions 

were not always consistent with the approved vendor list scope restrictions or the 

purchase order conditions. 

These nonconformances are cited in the enclosed Notice of Nonconformance 

(NON), and the circumstances surrounding them a-e described in detail in the 

enclosed report. You are requested to respond to the nonconformances and 

should follow the instructions specified in the enclosed NON when preparing your 

response. 
In accordance with 10 CFR Part 2.790 of the NRC's "Rules of Practice," a copy of 

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

Sincerely , 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Docket No.: 99900861 

Enclosures: 7. Notice of Nonconformance 

2. Inspection Report 9990861 /97-01 

22- 
NOTICE OF NONCONFORMANCE 

DuBose National Energy Services,lnc. 

Clinton NC 

Docket No.: 99900861 

Based on the results of an inspection conducted on January 21 through 24, 1997,- 

it appears that certain of your activities were not conducted in accordance with 

NRC requirements. 

A. Criterion V, "instructions, Procedures and Drawings," of Appendix B to 10 

CFR Part 50 requires, in part, that activities affecting quality shall be 

prescribed by documented instructions, procedures, or drawings and shall be 

accomplished in accordance with these instructions, procedures, or 

drawings. 

Paragraph NCA-3861 (b) states that the Material Organization shall transmit 

all certifications received from other Material Organizations or approved 

suppliers to the purchaser at time of shipment. 

Paragraph NCA-3862.l(b) of Section 111 of the ASME Code states that 

"When required chemical analyses (including mill heat analysis), heat 

treatment, tests, examinations, or repairs are subcontracted, the approved 

supplier's certification for the operations performed shall be furnished as an 

identified attachment to the Certified Material Test Report. " 

The following examples demonstrate failure to comply with the above 

requirements and constitute Nonconformance 99900861 /97-01-01 

1. Contrary to the above, DuBose Certified Material Test Report (CMTR) 

for material supplied to San Onofre Generating Station (SONGS), 

under Purchase Order (PO) 6N236016, certified that this material was 

supplied in accordance with their ASME Quality Systems Certificate 

(QSC), but did not reference or include the mill heat analysis as an 

identified attachment. 

2. Contrary to the above, DuBose CMTRs for material supplied to Union 

Electric Co. under PO 094006, dated June 24, 1996, and to Carolina 

Power & Light Co. (CPL), under PO 8M1717 CJ, dated November 1, 

1996, certified that the material was supplied in accordance with their 

QSC, but did not reference or include certifications from approved 

suppliers as identified attachments to their CMTRs. 

-23- 
B. Crkerion VI!, "Control of Purchased Material, Equipment, and Services," of 

Appendix t3 to 10 CFR Part 50, requires, in part, that purchased material 

conforms to procurement documents. 

Paragraph 10.3.1.4 of DuBose quality systems program (QSP) requires, in 

part that QA will review the vendor documentation against requirements of 

the PO, material specification, and, when required, the ASME Code. When 

unacceptable documentation ,is received, the material will be segregated until 

proper documentation can be obtained. 

Paragraph 2.3.4.1 of DuBose QSP requires, in part, that POs shall require 

Certification that the material was furnished under a program accepted by 

DuBose. 

The following examples demonstrate failure to comply with the above 

requirements and constitute Nonconformance 99900861 /97-01-02 

1. Contrary to the above, DuBose accepted material supplied under PO 

2284-63, dated February 5, 1993, without the starting material 

supplier being identified on the CMTR as required by this PO. 

2. Contrary to the above, DuBose accepted material under POs 2284-63 

and 1546-63, dated February 5, 1993, and August 24, 1992, 

respectively, without documented evidence that all requirements of 

the material specification (macroetch test) had been completed. 

3. Contrary to the above, DuBose accepted material under POs 11645- 

66 and 1061 2-62, dated December 9, 1996, and December 12, 

1996, respectively, without certification that the material was 

furnished under a quality program accepted by DuBose. 

C. Criterion IV, "Procurement Document Control," of Appendix B to 10 CFR 

Part 50 requires, in part, that requirements necessary to assure adequate 

quality are included in the documents for procurement of material. 

Paragraph 2.3.4.1 of the DuBose QSP requires, in part, that POs shall 

require certification that the material was furnished under a program 

accepted by DuBose. 

2 

-24- 
Contrary to the above, DuBose issued POs 11 495-61, and 10649-68, dated 

November 1 1, 1996, and November 15, 1996, respectively, without 

including a requirement that the material must be furnished under a program 

accepted by DuBose. These POs had been reviewed and approved by 

Du6ose QA on November 12, 1996, and November 18, 1996, respectively. 

(Nonconformance 99900861 -01 -03) 

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

Appendix B to 10 CFR Part 50 requires, in part, that the effectiveness of the 

control of quality by contractors and subcontractors shall be assessed at 

intervals consistent with the importance, complexity, and quantity of the 

product 

Paragraph 7.4.1 of DuBose QSP states, in part, that evaluation (of vendor) is 

performed to adequately evaluate the quality and technical capability of the 

vendor. 

Paragraph 2.4.3 of DuBose QSP states, in part, that the approved vendor's 

list (AVL) shall be compiled on the basis of a site evaluation of the vendor's 

quality program and Paragraph 2.4.2 states that the AVL shall show the 

scope of activity approved with any limitations. 

Paragraph 2.3.2 states that quality related material and services shall only 

be purchased from vendors appearing on the AVL. 

The following examples demonstrate failure to comply with the above 

requirements and constitute Nonconformance 99900861 /97-01-04. 

1. Contrary to the above,.DuBose audit of Colonial Machine Company on 

April 11, 1996, did not document that all areas of the AVL approved 

scope of activity had been evaluated and determined to be adequately 

controlled. Specifically, the audit documentation .did not describe 

what was reviewed in the areas of NDE, heat treatment, upgrading of 

unqualified source material, and control of suppliers of source material 

and weld repair activities were not addressed. 

2. Contrary to the above, DuBose audit of Capitol Manufacturing 

Company on February 13, 1996, did not document that all areas of 

the AVL approved scope of activity had been evaluated and were 

adequately controlled. Specifically, the audit documentation did not 

3 
describe what was reviewed in the areas of material testing and NDE, 

and did not address heat treatment and the upgrading of unqualified 

source material. 

3. Contrary to the above, on November 11, 1996, and on November 15, 

1996, DuBose issued POs 1 1495-61 and 10649-68, respectively, to 

Marmon/Keystone Corporation for material that was not included in 

the product scope of the AVL. 

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

Commission, AlTN: 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 Rockville, Maryland 

this day of 1997 

Enclosure 1 

4 

-26- 
Report No: 

Organization: 

Contact: 

Nuclear industry 

Activity: 

Dates: 

Inspectors : 

e 

Approved by: 

U.S. NUCLfAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

99900861 /97-01 

DuBose National Energy Services, inc. 

James Dailey, Quality Assurance Manager 

(910) 590 2151 

Supplier of threaded fasteners, bars, tubular products, 

forgings, plates, flanges, fittings, and other items used 

primarily for nuclear applications. 

January 21 through 24, 1997 

Uldis Potapovs, Senior Reactor Engineer 

Steven M. Matthews, Quality Assurance Engineer 

Gregory C. Cwalina, Chief 

Vendor Inspection Section 

Special inspection Branch 

Division of Inspection and Support Programs 

Enclosure 2 

-27- 
1 

2 

3. 

3.1 

INSPECTION SUMMARY 

During this inspection, the inspectors reviewed the implementation of 

selected portions of DuBose National Energy Services, Inc. (DuBose) quality 

assurance (QA) program for supplying material under the American Society 

of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) 

requirements, and for providing dedicated commercial grade material. The 

inspection was focused on the review of DuBose activities related to the 

supply of nonconforming material to Southern California Edison for use in 

the San Onofre Generating Station (SONGS). 

The inspection bases were: 

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

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

Reaulations (10 CFR Part 50). 

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

ASME Code, Section 111, Subarticle NCA 3800. 

During this inspection, four instances where DuBose failed to conform to 

NRC requirements imposed upon them by NRC licensees were identified. 

These nonconformances are discussed in Sections 3.2 and 3.3 of this 

report. 

STATUS OF PREVIOUS INSPECTION FINDINGS 

0 

This was the first NRC inspection of DuBose at its present location and 

under the present management. The last NRC inspection of DuBose was 

conducted on May 21 through 25, 1984, at Roseboro, NC. The status of 

findings from that inspection were not reviewed. 

INSPECTION FINDINGS AND OTHER COMMENTS 

Descrbtion of Facilities and Activities 

DuBose has been accredited by the ASME as a Material Organization (MO), 

supplying a wide range of metallic materials, including bars, threaded 

fasteners, castings, forgings, plates, fittings, flanges, tubular products, 

structural shapes, and weld filler material. According to DuBose 

management, approximately 70-80% of sales volume consists of safety 

related material supplied to nuclear utilities. Of that amount, approximately 

50% is sold under their ASME Quality Systems Certificate as complying 

-2- 

-28- 
3.2 

a. 

with the requirements of NCA 3800. The company has an on site laboratory 

facility capable of performing spectrochemical analyses, tensile testing, 

hardness testing, and hydrostatic testing. No manufacturing operations, 

other than subdividing of material, are performed. 

Jollow UD on Nonconformina material SUDD li e d t oSONGS 

Backaround 

On May 10, 1996, DuBose issued a Certificate of Conformance/Compliance/ 

CMTR (CMTR) to SONGS for forty six 3/4-10 ASME SA-1 94, Grade 7, 

heavy hex nuts, supplied as item 2 of SONGS purchase order (PO) 

6fU236016. SONGS receipt inspection of these nuts resulted in a 

nonconformance report (NCR) for failure to meet System 22 dimensional 

inspection requirements, as specified in the PO. It was also identified that 

chemical analysis of ten nuts showed that, while they complied with the 

specification limits, the nuts apparently came from two different heats of 

material (all 46 nuts were certified as coming from a single heat). SONGS 

performed a 10 CFR PART 21 reportability evaluation and determined that 

this event was not reportable. 

Review of documentation obtained from SONGS indicated that the material 

was shipped from T&T Enterprises, Corona California, a supplier qualified by 

DuBose. DuBose provided the CMTR for this material, referencing their 

ASME QSC nufnber and certifying compliance with the PO and ASME Code 

requirements. Attached to the DuBose CMTR was a CMTR from T&T 

Enterprises, certifying the performance of dimensional inspection (System 

221, visual inspection, and macroetch (ASTM E-381 1, in accordance with 

their DuBose-approved QA program. Also furnished with the material was a 

CMTR from A&G Engineering 11, fnc. (A&G), dated February 8, 1993 and 

issued to DuBose. The CMTR certified that the material was provided under 

A&G's QSC and contained the results of chemical and mechanical tests, 

apparently transcribed from Korea Bolt Ind. Co., LTD., Inspection Certificate, 

dated October 10, 1991, attached to the A&G CMTR. The certifications did 

not include a CMTR from the producing steel mill, and there was no 

indication of who produced the steel. 

On September 12, 1996, NRC contacted DuBose and requested information 

concerning the apparent loss of traceability for this material. DuBose 

responded on September 19, 1996, by stating that, based on their review 

and additional chemical analyses, they determined that there was no 

conclusive evidence to indicate that the nuts were from mixed heats of 

material. This conclusion was based primarily on the fact that the heat code 

was forged into each nut by the manufacturer and that the chemical 

-3- 

-29- 
b. 

overcheck results (1 5 samples) were within acceptable variance limits of 

ASTM A-29. 

Observations and Findinog 

Documentation available at DuBose for the material discussed above did not 

contain a CMTR from the producing mill. Also, other than the macroetch 

that was apparently performed by T&T Enterprises, as discussed above, on a 

finished nut, there was no indication that macroetch testing was performed 

as a part of the steel producer’s quality procedures, as required by the 

applicable material specification, ASME SA-1 94. A macroetch test on a 

forged nut, performed after the completion of manufacturing is not 

considered an acceptable method of achieving compliance with the material 

specification requirement. With respect to the missing CMTR from the steel 

producer, the inspector noted that, for material supplied under DuBose’s 

QSC, paragraph NCA 3862.1(b) of the ASME Code requires that 

certifications of approved suppliers be furnished as identified attachments to 

the QSC holder‘s CMTR. In this case, the steel mill was considered an 

approved supplier, and the mill CMTR, or heat analysis, should have been 

available, and should have been provided to the recipient of the nuts. 

With respect to DuBose’s comment, as discussed above, that forging of the 

heat code into each nut during the manufacturing process assures heat 

traceability, the inspector noted that the forged-in heat markings could not 

be relied upon $0 provide full heat traceability for this situation, since a 

material mixup may have occurred at the producing mill or at an intermediate 

supplier to the nut manufacturer. 

With respect to DuBose statement that, since the results of their chemical 

analyses of 15 nuts were within acceptable product analysis limits ’of ASTM 

A-29, there was no conclusive evidence that the nuts were from mixed 

heats, the inspector noted the applicable variance limits in A-29 were not 

intended to verify heat homogeneity, and that multiple analyses of a single 

heat of material would be expected to show less variability than indicated in 

the tables referenced in A-29. 

Review of the nonconforming thread issue identified during SONGS receiving 

inspection showed that DuBose issued Non-Conformance Report (NCR) No. 

1039 on June 18, 1996, as a result of communication from SONGS. Based 

on the NCR, a Corrective Action Report was issued to T&T Enterprises on 

June 19, 1996, describing the unsatisfactory condition, and requesting a 

response by July 19, 1996. On September 6, 1996, T&T Enterprises 

provided a response by stating that the internal threads were gaged with 

GO-NOGO plug gages and GO-NOGO cylinder gages which satisfied the 

-4- 

-3 0- 
functional requirements of System 22. They identified the cause of the 

nonconformance as unavailabiiity of the cone and vee type internal variable 

gages to appraise the values of the internal pitch diameters over the full 

length of the thread. The described corrective action was to assure the 

proper equipment is used to make these measurements on future orders. 

DuBose accepted the proposed corrective action on September 9, 1996. 

The inspector also reviewed several procurement documents related to 

DuBose’s acquisition of material from A&G, as summarized below: 

The nuts discussed above were purchased as a part of DuBose PO 

2284-63, dated February 5, 1993 for approximately 85,000 ASME SA194 nuts in sizes ranging from 5/8-inch to 1 1/2- inches. The nuts were 

to be provided under A&G’s QSC. The inspector noted that PO 

contained an attached Form D-2. Paragraph 6. of this form stated that 

“The starting material manufacturer is to be identified on the CMTR 

produced by your company.“ Paragraph 7. of the same form stated that 

the use of foreign starting material is not permitted without DuBose’s 

approval and that, when approval is given, backup certification is to be 

provided. The PO was amended on February 8, 1993, to allow foreign 

starting material. All other conditions were to remain the same. There 

was no further explanation concerning “backup certification.” The 

inspector’s review of documentation supplied by A&G did not identify 

any reference to ”backup certification,” and did not show any evidence 

that the surting material manufacturer had been identified as specified 

in the attakhment to the DuBose PO for this material. 

On August 24, 1992, DuBose issued PO 1546-63 to A&G for 20 tons 

of threaded rod in accordance with ASME SA-193, Grade B7 in 12-foot 

lengths, and diameters ranging from 1/2 to 1 inch, to be supplied under 

A&G’s QSC. The A&G CMTR for this material contained apparently 

transcribed results of the required chemical and mechanical tests and 

certified the material as being provided in accordance with their QSC. 

The CMTR did not include certification from the steel mill or include any 

certifications by approved suppliers. Also, there was no indication that 

a macroetch test, as required by the material specification, was 

performed during the steelmaking process. 

C. Conclusions 

As discussed above, the inspectors determined that, for SONGS PO 

N236016, DuBose did not provide all of the documentation required by the 

ASME Code provisions governing the application of their QSC when 

certifying Code material. Specifically, contrary to the requirements of ASME 

-5- 

-31- 
3.3 

Code, SC Ill, paragraph NCA 3862.1(b), DuBose did not include the 

producing mill heat analysis as an identified attachment to their CMTR. This, 

item was identified as an example of Nonconformance 99908861 897-01 -01. 

The inspectors also determined that, contrary to Paragraph 10-3. I .4 of their 

QA Manual, A&6 certifications for materia! supplied to fill DuBose $Os 

2284-63 and 1546-63 were either not reviewed by DuBose QA, or the 

review was not effective, Specifically, this material was accepted without 

identification of "starting material manufacturer" on these certifications, as 

required by Form 0-2, attached to the DuBose POs. Additionally, the A&G 

certifications did not indicate that all of the testing, required by the material 

specification (macroetch), had been performed. These items were identified 

as examples of Nonconformance 99900861 /97-01-02. 

With respect to the mixed heat issue, based on review of the results of 

DuBose chemical analyses of 15 nuts, it appears likely that more than one 

heat of material was represented in the sample. However, as discussed 

above, since none of the samples tested by DuBose or SONGS failed to 

meet the requirements of the material specification, co-mingling of heats, if 

it did occur, appears to have been limited to the same grade of material. 

Additionally, a July 15, 1996, memorandum from the DuBose QA Manager 

to its President indicates that this material was being moved to the DuBose 

commercial inventory. The dimensional nonconformance of these nuts 

apparently resulted from the subcontractor's failure to perform all sf the 

required measurements, and was dispositioned through the DuBose 

corrective action process. 

QA Program lmdernentation 

3.3.7 Proaram DescriDtion 

The DuBose quality assurance program is described in their Quality System 

Program (QSP) and in Quality Control Procedures (QCPs). Revision 5 of the 

QSP, dated January 10, 1996, was in effect at the time of the NRC 

inspection. The QSP was written to comply with the ASME Code, Section 

161 requirements applicable to Material Organizations, and a9so indicates 

compliance with 10 CFR 50, Appendix B, MIL 145208, MIL 0 9858, NQA 

1, ANSI 45.2, And IS0 9002. 

Only selected areas of the QA program and its implementation were 

reviewed during this inspection. 

-6- 

-32- 
3.3.2 Procurement Control 

a. 

b. 

b.1 

JnsDect ion SCODQ 

?he inspectors reviewed DuBose approved vendors list IAVL) dated January 

21, 1997, and selected several suppliers from the AVL Po evaluate DuBose 

capability to qualify vendors for placement on the AVL. 

Observations and Findings 

Approved Vendors 

According to Section 2.4, "Approved Vendors List," of QuBose QSP, 

vendors are placed on the AVL either on the basis of a site evaluation of the 

vendor's QA program in accordance with Section 7.4, "Vendor Surveys and 

Audits," of the QSP, or on the basis of verification that the vendor has an 

ASME issued OSC or an N-type Certificate of Authorization. Vendors that 

are placed on the AVL based on a DuBose site evaluation are qualified to one 

of two levels: For material traceability only, or as a QuBose qualified MQ 

with QA program meeting the requirements of Subarticle MCA-3800 of 

Section Slli of the ASME Code. 

To evaluate QuBose capability to qualify vendors for placement on the AVL 

as an approved supplier, the inspectors reviewed the following audits: 

(1 1 MarmonKeystone Corporation, Charlotte, North Carolina 

The last DuBose audit of MarmonlKeystone was performed on October 

13, 1994. The scope of the audit identified Marmon/Keystone as a 

supplier sf ferrous and non-ferrous, seamless, and welded without filler 

metal, tubular products. Based on a review of the audit report, the 

inspectors determined that DuBose had adequately evaluated 

Marmon/Keystone, Charlotte, North Carolina, as a supplier of source 

material with controlled traceability. 

However, the AVL's approved scope for purchases from 

Marrnon/Keystone was, "MS - carbon welded tubular products in 

accordance with OA manual, 1st Edition, Revision 0, dated January 27, 

1988." Therefore, in this instance, the AVL approved scope for 

purchases was inconsistent with the scope of DuBose's audit of 

Marmon/Keystone since it did not include nonferrous or seamless tubular 

products 
(2) Alloy Rods Incorporated 

The last DuBose audit of Alloy Rods Inc., was performed on April 15, 

1994. The scope of the audit was, "MO - ferrous and nonferrous bare 

and covered wire." On the basis of their review of the audit, the 

inspectors determined that DuBose had adequately evaluated Alloy Rods 

Inc. as a qualified MO. 

However, the AVL's approved scope for purchases from Alloy Rods Inc. 

was, "MO - ferrous and nonferrous brazing material, bare and covered 

flux cored electrodes, and bare wire." Therefore, the AVL's approved 

scope for purchases was not consistent with, and was expanded 

beyond the scope of the DuBose audit of Alloy Rods Inc. Specifically, 

the audit scope did not include brazing material or bare and covered flux 

cored electrodes. This condition would permit DuBose buyers to 

procure material that was not included in the audited, and therefore, 

qualified capability of Alloy Rods lnc. 

(3) Colonial Machine Company, Inc., Pleasantville, Pennsylvania 

The last DuBose audit of Colonial Machine Company was performed on 

April 1 1 , 1996. The scope of audit was: "MO - ferrous and 

nonferrous fittings, flanges and pipe hangers - machining and 

proce.csing of DuBose supplied material. " However, the Vendor Quality 

Assurance. Survey/Audit Cover Sheet (Form D-461, also listed additional 

items inclcding the following: 

bars 

plates 

seamless fittings 

forgings 

flanges 

seamless tubing product 

NDE (nondestructive examination) 

HT (heat treatment) 

weld repair 

material testing 

upgrading unqualified source material 

approval & control of suppliers of source material 

approval & control of suppliers of services 

-8- 

-34- 
On the basis of their evaluation of DuBose April 1 1, 1996, audit of 

Colonial Machine Company, the inspectors determined that the audit 

report did not contain objective evidence that the audit had adequately 

evaluated the following areas: 

0 

0 

0 

for NDE, HT, upgrading unqualified source material, and 

approval & control of suppliers of source material, the DuBose 

auditor did not document objective evidence of what was 

audited 

weld repair was not addressed 

for approval & control of suppliers of services, the DuBose 

auditor did not document objective evidence that supplier audits 

were evaluated. 

Additionally, the AVL's approved scope for purchases from Colonial 

Machine Company was, "MO - ferrous and nonferrous fittings, flanges 

and pipe hangers - machining and processing of DuBose supplied 

material - welding in accordance with ASME 631.1 Code." Therefore, 

the AVL's approved scope for purchases was not consistent with and 

was expanded beyond the scope of DuBose's audit of Colonial Machine 

Company. Specifically, welding in accordance with ASME B 31.1 Code 

was not included in the audit scope. This condition would permit 

DuBose buyers to procure material that was not included in the audited, 

and therefore, qualified capability of Colonial Machine Company. 

(4) Capitol Manufacturing Company 

The last DuBose audit of Capitol Manufacturing Company was 

performed on February 13, 1996. The scope of audit was: "MO - 

manufacturing carbon and stainless fittings." However, DuBose Form 

0-46, also listed additional items including: 

seamless fittings 

NDE 

HT 

material testing 

upgrading unqualified source material 

approval & control of suppliers of source material 

approval & control of suppliers of services 

-9- 

-35- 
b.2 

On the basis of its evaluation of DuBose February 13, 1996, audit of 

Capitol Manufacturing Company, the inspectors determined that the 

audit failed to adequately document evaluation of the following areas: 

0 

0 

0 

for NDE and material testing, DuBose did not document 

objective evidence of what was evaluated 

HT was. not addressed 

for upgrading unqualified source material, DuBose did not 

document a review of upgrading activities. 

Procurement Document Control 

To evaluate DuBose capability to control procurement documents, the 

inspectors reviewed the foliowing purchase orders (POs): 

(1) PO 11495-61 

DuBose issued PO 1 1495-61, dated November 1 1, 1996, to 

Marmon/Keystone Corporation, Charlotte, North Carolina, for 20-feet 

(ft.), random length, 8-inch nominal diameter, scheduie 160, A-106 

Grade B, seamless pipe. The PO was reviewed and approved by 

DuBose QA on November 12, 1996. 

However, the PO failed to impose on Marmon/Keystone the requirement 

that material must be controlled in accordance with the quality control 

system and procedures audited by DuBose. Marmon/Keystone also 

failed to provide DuBose certification that the material was furnished 

under a quality program accepted by DuBose. 

Additionally, the AVL approved scope for purchases from 

Marmon/Keystone was, "MS - carbon welded tubular products in 

accordance with QA manual, 1st Edition, Revision 0, dated January 27, 

1988." Since this PO procured seamless pipe, the PO did not comply 

with the limitations of the DuBose AVL. 

(2) PO 10649-68 

DuBose issued PO 10649-68, dated November 15, 1996, to 

Marmon/Keystone Corporation, Charlotte, North Carolina, for 60-ft., 

W8-inch outside diameter (OD) x .083-wall x 20-ft., A-269 type 304 

stainless steel tubing. The PO was reviewed and approved by DuBose 

QA on November 18, 1996. 

-10- 

-36- 
However, the PO failed to impose on Marmon/Keystone the requirement 

that material must be controlled in accordance with the quality control 

system and procedures audited by DuBose. Marmon/Keystone also 

failed to provide DuBose certification that the material was furnished 

under a quality program accepted by DuBose. 

Additionally, the AVL approved scope for purchases from 

Marmon/Keystone was, "MS - carbon welded tubular products in 

accordance with QA manual, 1st Edition, Revision 0, dated January 27, 

1988." Since this PO procured stainless steel tubing, the PO did not 

comply with the limitations of DuBose AVL. 

DuBose issued PO 11 645-66, dated December 9, 1996, to TAD USA, 

Inc., for one piece of 2-inch, schedule 80 x lO-ft., SA-335, P22, 

seamless pipe, The PO was reviewed and approved by DuBose QA on 

December 11, 1996. 

Even though DuBose imposed on TAD USA the requirement that 

material furnished must be controlled in accordance with the quality 

control system and procedures audited by DuBose, TAD USA failed to 

provide DuBose certification that the material was furnished under a 

quality program accepted by DuBose. 

(4) PO 10612262 

DuBose issued PO 1061 2-62, dated December 12, 1996, to Castle, 

A.M. & Company for one piece of 2 3/4-inch thick x 35-inch wide x 36- 

inch long, SA-516 Grade.70, plate. 

Even though DuBose imposed on Castle, A.M. & Company the 

requirement that material furnished must be controlled in accordance 

with the quality control system and procedures audited by DuBose, 

Castle, A.M. & Company failed to provide DuBose certification that the 

material was furnished under a quality program accepted by DuBose. 

C. Conclusions 

The inspectors determined that the DuBose material traceability audits of 

suppliers appeared to be adequate for their intended purpose. 
However, the inspectors determined that the audits of DuBose-qualified 

NCA-3800 suppliers (MOs) were lacking in detail and programmatic in their 

conduct andlor documentation, The inspectors identified two instances 

(Colonial Machine Company, Inc., and Capitol Manufacturing Company) 

where DuBose audits of its vendors failed to (a) adequately assess the 

effectiveness of the control of quality by its vendors and (b) compile its 

AVL, for all vendors, on the basis of an adequate site evaluation of the 

vendors quality program for the AVL's approved scope for purchases. These 

audits often lacked objective evidence that upgrading and dedication 

activities were adequately verified or that the implementation of these 

activities was evaluated. These findings were identified as examples of 

Nonconformance 99900861 /97-01-04, 

Additionally, in certain instances identified by the inspectors and described 

above, DuBose AVL approved scope for purchases was expanded beyond 

the scope of DuBose's audit and therefore permitted DuBose buyers to 

procure material that was not included in the audited and qualified capability 

of the supplier. 

In other instances identified by the inspectors and described above, DuBose 

procured material form vendors on its AVL that was not included in the AVL 

approved product scope (POs 1 1495-61 and 10649-68). These instances 

were identified as examples of Nonconformance 99900861 /97-01-04. 

The inspectors identified two instances (POs 1 1645-66 and 1061 2-62) 

where DuBose-accepted material from its vendor without certification that 

the material was furnished under a quality program accepted by DuBose. 

These items were identified as examples of Nonconformance 99900861 /97- 

01-02. 

On the basis of its review of the POs described above, the inspectors 

identified two instance (POs 11495-61 and 10649-68) where Dubose failed 

to require its vendors to certify that the material was furnished under a 

quality program accepted by DuBose. These items were identified as 

examples of Nonconformance 99900861 /97-01-03. 

3.3.3 Review of Recent Material Sales To NRC Licensees 

a. ScoDe 

The inspectors reviewed selected documentation packages for safety related 

material recently supplied to NRC licensees under the DuBose's QSP to 

verify the implementation of the applicable QA program requirements and 

compliance with 10 CFR 50, Appendix B, provisions. 

-1 2- 

-38- 
b. Observations and Findinas 

Document packages for material supplied to Appendix B requirements, as well as 

material certified under Du%ose QSC, were reviewed. Significant observations and 

findings are summarized below. 

Union Electric PO 094006, dated June 24, 1996, Item (2) for twenty 

ASME SA-564 Grade 630 studs (3/4-10, 5 1/2in long), with 1100 F 

aging treatment, and 302 BHN minimum hardness. 

This material was certified under DuBose QSC as complying with the 

applicable Code and PO requirements. Appended to DuBose CMTR was 

a certification from Walker Bolt, a DuBose-qualified supplier. Attached 

to the Walker Bolt certification was certification from Ugine Savois 

(French company) containing chemical and tensile test results, 

satisfactory macroetch statement and statement that the material was 

provided in solution treated condition. No quality program statement 

was attached to this certification, indicating that Ugine Savois was not 

an approved supplier and that the material was upgraded by Walker 

Bolt. However, no test reports, indicating that upgrading of this 

material was performed were attached to the Walker Bott certification. 

There was also no indication of what organization performed the 

required aging treatment on this material, or the actual time and 

temperature of the heat treatment. 

The inspector noted that Paragraph 3.3.10 of the DuBose QSM section 

on utilization of unqualified source material states that "... the lab test 

report shall be attached to the DuBose Certificate of Conformance." 

Paragraph 3.3.1 1 states that the Certificate will contain a statement 

"upgraded per NCA 3855.5." 

Before the completion of this inspection, Walker Bolt telefaxed an 

amended certification to DuBose indicating that upgrading tests were 

performed at their facility and including a certification of the heat 

treatment. 

a Carolina Power & Light Co. PO 8M1717 CJ, dated November 1, 1996 

for thirty four 1-& 2 1/2inch long ASME SA-193, Grade B7 Cap screws. 

This material was certified under DuBose QSC, as complying with the 

applicable Code and PO requirements. Appended to the DuBose CMTR 

was a certificate from T&T Enterprises (T&T), a DuBose-qualified 

supplier, containing chemical analyses, results of tensile and hardness 

tests, description of heat treatment, visual inspection, and a statement 

-1 3- 

-39- 
indicating that a macroetch test had been performed with satisfactory 

resultsr The T&T certification also contained a quality program 

statement that all activities had been performed in accordance with their 

DuBose-approved QLA program. A Republic Steel Co. CMTR for this 

heat of material was attached. The Republic Steel certification was sent 

to Fry Steel Co., a distributor, indicating that T&T obtained this material 

as unqualified source material and upgraded it. 

The inspector noted that, since the starting material was in the dorm of 

1-inch bar, cold or hot heading operations, as well as heat treatment 

wouSd have been involved in the manufacturing process, requiring the 

use of approved suppliers. However, no certifications for these 

operations were in the data package. 

0 Public Service Electric & Gas Co. PO P2 0902993 1850 0000 for four 

10 gage, 4 by 8 foot steel sheets. 10 CFR 50,Appendix B QA 

requirements and 10 CFR Part 21 were invoked by the PO. 

DuBose certified this material as complying with their QSP and the PO 

requirements. DuBose obtained the material from NUCOR Steel Co., a 

commercial supplier, and dedicated it under their program. The 

dedication consisted of performing complete chemical analysis on each 

of the four sheets, and tensile and bend tests on one sheet. The 

laboratory report, including the test results and the NUCOR Steel Co. 

material certification were attached to the Dubose CMTR. The chemical 

analyses if samples from the four sheets showed close agreement to 

the NUCOR Steel Co. CMTR, supporting the bases for verifying 

mechanical properties of only one sample. 

C. Conclusions 

Failure to include certifications from approved suppliers as identified 

attachments to DuBose CMTRs issued to Union Electric (PO 8M17f 7 CJ) 

and Carolina Bower and Light (PO 8M1717 CJ), were identified as examples 

of Nonconformance 99900861 /97-01-01, 

3.4 Entrance and Exit Meetinas 

In the entrance meeting on January 21, 1997, the NRC inspectors discussed 

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

established interfaces with DuBose management. In the exit meeting, on 

January 24, 1997, the inspectors discussed their findings and concerns. 

-1 4- 

-40- 
PARTIAL LIST OF PERSONS CONTACTED 

Car! M. Rogers, President 81 CEO 

John Hottel, Vice President 

James DaiOey, Manager, Quality Assurance 

Ailen Jones, Manager, Sales & Marketing 

Laurie Meyer-Dickson, Certification engineer 

ITEMS OPENED, CLOSED, AND DISCUSSED 

99900861 197-01 -01 

59900861 197-01 -02 

99900861 /97-01-83 

99908861 657-01 -04 

NON Failure to provide required documentation 

NON Acceptance of material with incomplete 

documentation 

NON Inadequate procurement Documents 

NON Inadequate supplier audits 

-1 5- 

-41- 
UNITED STATES 

NUCLEAR REGULATORY COMMiaarGN 

WASHINGTON, D.C. 205550001 

June 11, 1997 

Mr. Craig P. Kipp 

P1 ant Manager 

General Electric Nucl ear Energy 

P.O. Box 780 

Wilmington, NC 28402-0780 

SUBJECT: NONPROPRIETARY VERSION OF NRC INSPECTION REPORT 99900003/97-01 AND 

NOTICE OF NONCONFORMANCE 

Dear Mr. Kipp: 

This letter transmits the nonproprietary version of the U.S. Nuclear 

Regulatory Commission’s (NRC’s) inspection report at the General Electric (GE) 

Nuclear Energy facility. Our letter to you dated May 20, 1997, transmitted 

the original (proprietary) version of the report. On the basis of our 

discussions and review of the information in your letter (RJR-97-076) of June 

6, 1997, attached affidavit, and attached Proprietary Information Summary 

Sheet, we have concluded that the specific items identified in your letter 

could be regarded as proprietary and, as such, were removed from the 

inspection report. In the revised nonproprietary (public) version of the 

report, we have deleted or revised the text. 

Your response to either this letter or our letter dated May 20, 1997, 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 Section 2.790(a) of the NRC “Rules of Practice,” of Title 

10 of the Code of Federal Rewlations, a copy of this letter and its 

enclosures will be placed in ths NRC Public Document Room. 

any questions concerning the matter, please contact Ani1 S. Gautam of my staff 

at 301/415-2988. 

Should you have 

$&- R ber M. 110. Chief 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Docket No. 99900003 

Enclosures: NRC letter to GE, May 20, 1997 

-42- 
UNITED STATES 

NUCLEAR REGULATOirY COMMISSIO?i 

WASHINGTON, D.C. -1 

May 20, 1997 

Mr. Craig P. Kipp 

P1 ant Manager 

General Electric Nuclear Energy 

P.O. Box 780 

Wilmington, NC 28402-0780 

SUBJECT: NRC INSPECTION REPORT 99900003/97-03 AND NOTICE OF NONCONFORMANCE 

Dear Mr. Kipp: 

On March 14, 1997, the U.S. Nuclear Regulatory Commission (NRC) completed an 

inspection at the General Electric (GE) Nuclear Energy facility. The enclosed 

report presents the results of that inspection. 

The inspection was conducted to assess GE’s design interface with licensees 

that implemented the average power range monitor-rod block monitor-technical 

specification (referred to as ARTS) modification, that allowed licensees to 

bypass the rod block monitor (RBM) for a rod withdrawal error (RWE) event if 

the core had sufficient minimum critical power ratio margin. The inspectors 

assessed RBM operabi 1 i ty and re1 ated technical specifications actions with 

respect to protecting the fuel cladding in the boiling water reactor cores of 

ARTS plants for an RWE event. The inspectors also assessed GE’s monitoring of 

the effectiveness of its design control program. 

During the inspection, the inspectors determined that GE did not adequately 

inform ARTS licensees of the need to consider the fuel cladding plastic strain 

limits and the associated mechanical overpower (MOP) limits in addition to the 

minimum critical power ratio limits when considering RBM operability for 

potential RWE events. This inadequate interface between GE and ARTS licensees 

contributed to (1) the failure of licensees to ensure through their technical 

specifications that the RBM was operable to protect fuel cladding at 

applicable plants, and (2) occasions during certain operational cycles of the 

Fermi, Hatch, Brunswick, and Duane Arnold nuclear plants when, based on GE’s 

RWE analyses, the fuel cladding had exceeded its MOP limits and had the 

potential of exceeding its plastic strain limits. 

The inspectors determined that GE modified certain design parameters used in 

the RWE analysis without procedures or other documented basis, and did not 

assess the overall effectiveness of GE’s design control program after Detroit 

Edison Company (DECo) identified deficiencies in design adequacy. Based on 

the above, the inspectors concluded that the implementation of GE’s quality 

assurance program did not meet certain NRC requirements imposed on GE by its 

customers. 

-43- 
C. Kipp -2- 

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

circumstances surrounding them are described in detail in the enclosed report. 

You are requested to respond to the nonconformances and should follow the 

instructions specified in the enclosed NON when preparing your response. 

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

this letter and its enclosures will be placed in the NRC's Public Document 

Room. 

Si ncerel y , 

Originel signed by 

Robert M. Gallo, Chief 

Special Inspection Branch 

Division of Inspection and Support Programs Office of Mucl ear Reactor Regul at ion 

Docket No. 99900003 

Enclosures: 1. Notice of Nonconformance 

2. Inspection Report 99900003/97-01 

-44- 
NOTICE OF NONCOdFORMANCE 

General El ectri c Nucl ear Energy 

Wilmington, NC 

Docket No. : 99900003 

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

Comnisslon (NRC) from March 10 through 14, 1997, it appears that the following 

activities were not conducted in accordance with NRC requirements: 

1 Criterion 111 of Appendix B to 10 CFR Part 50, "Design Control," 

requires, in part, that measures shall be established for the 

identification and control of design interfaces and for coordination 

among participating design organizations. These measures shall include 

the establishment of procedures among participating design organizations 

for the review, approval , release, distribution, and revision of 

documents involving design interfaces. 

Paragraph 3.3 of Section 3, "Design Control ,'I General Electric (GE) 

Nuclear Energy Qual i ty Assurance Manual NEDO-11209, dated March 31, 

1989, requires, in part, that GE design documents be furnished to the 

customer to provide for interface compatibility review and coordination 

by owner design organizations. 

Contrary to the above requirements, GE did not adequately inform 

licensee design organizations implementing the average power range 

moni tor-rod block moni tor-technical speci f ication (ARTS) modification of 

the need to consider the 1 percent fuel plastic strain limits and the 

associated mechanical overpower (MOP) limits in addition to the minimum 

critical power ratio limits when evaluating rod block monitor (RBM) 

operabi'l ity for a rod withdrawal error (RWE) event. GE's supplemental 

reload licensing reports for ARTS plants did not adequately address 

requirements for RBM operability with regard to the MOP limits. This 

inadequate interface between GE and ARTS licensees contributed to (1) 

the failure of licensees to ensure through their plant technical 

specifications that the RBM was operable to protect fuel cladding at 

applicable plants, and (2) occasions during Fermi Cycles 4, 5 and 6, 

Hatch Unit 1 Cycles 16 and 17, Hatch Unit 2 Cycles 13 and 14, Brunswick 

Cycle 10, and Duane Arnold Cycle 14 when based on GE's RWE analyses the 

fuel cladding had exceeded its MOP limits and had the potential of 

exceeding its plastic strain 1 imits. 

2 Criterion V of Appendix 8 to 10 CFR Part 50, "Instructions, Procedures, 

and Drawings," requires, in part, that activities affecting quality 

shall be prescribed by documented instructions and procedures of a type 

appropriate to the circumstances, and shall include appropriate 

quantitative or qualitative acceptance criteria for determining that 

important activities have been satisfactorily accomplished. 

(99900003/97-01-01} 

GE Nuclear Energy Quality Assurance Manual NEDO-11209, Section 5, 

"Instructions, Procedures, and Drawings," dated March 31, 1989, 

requires, in part, that documented instructions, procedures, and 

drawings be utilized to communicate quality requirements throughout all 

phases of design. 

Enclosure 1 

-45- 
Contrary to the above requirements, GE (1) modified peaking factors when 

the MOP limits were exceeded in the RWE analyses, (2) applied alternate 

rod patterns in addition to normal rod patterns in the RWE analyses, and 

(3) revised the theoretical density values used in the peak cladding 

temperature (PCT) analysis to reduce the calculated PCT, without 

documented instructions or procedures. (99900003/97-01-02) 

3 Criterion XVIII of Appendix B to 10 CFR Part 50, "Audits," requires, in 

part, that periodic audits shall be carried out to determine the 

effectiveness of the quality assurance program. Followup action, 

including reaudit of deficient areas, shall be taken where indicated. 

GE Nuclear Energy Quality Assurance Manual NEDO-11209, Section 18, 

"Audits," dated March 31, 1989, requires, in part, that the audit 

program provide for followup action, including any necessary reaudit of 

deficient areas. 

Contrary to the above requirements, Detroit Edison Company (DECo) 

audited GE in 1992 and 1993 and observed several deficiencies regarding 

design control, including GE's failure to inform DECo (and other ARTS 

licensees) that the MOP limits would be exceeded if the RBM was not 

operable during an RWE event. GE took corrective actions for specific 

deficiencies but did not conduct foilowup action, including reaudit of 

the design control area, to determine the effectiveness of the program. (99900003/97-01-03) 

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

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

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

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

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

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

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

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

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

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

Dated at Rockvi 11 e , Mary1 and 

this 20th day of May 1997 

-46- 
Report No: 

Organization : 

Contact : 

Nuclear Industry 

Activity : 

Dates: 

Inspectors: 

Approved by: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

99900003/97-01 

General Electric Nuclear Energy 

Wilmington, North Carolina 

Ralph 3. Reda, Manager 

Fuels & Facility Licensing 

910/675-5000 

Nuclear fuel assemblies and related reload core 

designs, core components, safety analyses, and 

services 

March 10-14, 1997 

Ani1 S. Gautam, Team Leader, NRR 

Laurence E. Phil 1 ips, NRR 

Kombiz Salehi, NRR 

Edward D. Kendrick, NRR 

Geoffrey R. Golub, NRR 

John F. Carew, Brookhaven National Laboratory 

Carl E. Beyer, Pacific Northwest National Laboratory 

Gregory C. Cwalina, Chief 

Vendor Inspection Section 

Special Inspection Branch 

Division of Inspection and Support Programs 

Enclosure 2 

-47- 
1 INSPECTION SUMARY 

During this inspection, the NRC inspectors reviewed activities associated wlth 

General Electric (GE) Nuclear Energy's design interface with 1 icensees that 

implemented the average power range moni tor-rod block moni tor-technical 

specifications (hereafter referred to as ARTS) modi f 1 cation. The ARTS 

modification allowed licensees to bypass the rod block monitor (RBM) for a rod 

withdrawal error (RWE) above 30 percent power (hereafter referred to as RWE 

event) if the core had sufficient minimum critical power ratio (MCPR) margin. 

The inspectors focused on evaluation and implementation of RBM operability to 

prevent exceeding fuel cladding 1 percent plastic strain limits (hereafter 

referred to as strain limits) in the boiling water reactor fuel cores of the 

ARTS plants during an RWE event. The inspectors examined technical 

specifications (TS) actions to ensure RBM operability during an RWE event to 

keep fuel cladding from exceeding the strain limits and associated MOP limits. 

The inspectors also assessed GE's monitoring of the effectiveness of its 

design control program. 

The inspection bases were as follows: 

0 

0 

0 

Appendix 8, "Quality Assurance Criteria for Nuclear Power Plants and Fuel 

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

ReQulations (10 CFR Part 50). 

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

GE Nuclear Energy Quality Assurance Program Description NEDO-11209-04A, 

Revision 8, dated March 31, 1989, and associated implementing procedures. 

During this inspection, the inspectors noted three instances in which GE 

failed to conform to NRC requirements imposed upon it by NRC licensees. 

nonconformances are discussed in Section 3.1 and 3.3 of this report. 

These 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

This inspection was limited in scope to address GE's design interface 

with licensees. Previous inspection findings were not evaluated. 

3 INSPECTION FINDINGS AND OTHER CONHENTS 

3.1 Pesian Interface with ARTS Licensees 

a. InsDection ScoDe 

The inspectors assessed the adequacy of GE's design interface with ARTS 

licensees regarding RBM operability during an RWE event. 

examined GE's RWE analyses and supplemental reload 1 icensing reports 

(SRLRs) and GE's interaction with licensees regarding the assumptions and 

results of its analyses. 

The inspectors 

2 

-48- 
b. Observations and Findinps 

The RWE event assumes erra eous withdrawal of the highest worth rod from 

full in to full out at any time during the operating cycle. To protect 

against this event and keep the fuel from reaching a 1 percent plastic 

strain during an RWE event, the RBM monitors control rod movements and 

initiates a rod block during a RUE in which the local power exceeds a 

predetermined setpoint. 

conservatively defined as the threshold below which fuel damage due to 

fuel clad overstraining is not expected to occur. 

The 1 percent plastic strain of the cladding is 

GE uses conservative mechanical overpower (MOP) 1 imits to assure 

compliance with strain limits. GE started using MOP limits in 1987 after 

a study showed that the fuel assemblies in the error cell in newer fuel 

designs could potentially exceed the strain limits during an RWE event. 

Using the MOP limits in the RWE analysis added a margin of protection 

against exceeding the strain limits. 

assemblies did not exceed the MOP limits during an RWE event, they would 

not exceed the strain limits. If the RWE analysis indicated that the MOP 

limits would be exceeded, licensees could take credit for RBM 

operability, perform an exact site-specific analysis to ascertain whether 

the strain limits would be exceeded without RBM operability, or 

reconfigure the core design to avoid exceeding the MOP limits during an 

RWE event. 

GE determined that if fuel 

The MCPR limits were imposed to avoid fuel damage due to severe 

overheating of the fuel cladding. The ARTS modification allowed 

licensees to bypass the RBM for an RWE event if the core had sufficient 

MCPR margin. GE determined that for certain ARTS plants during certain 

fuel cycles the RBM needed to be operable to prevent exceeding the MOP 

limits during an RWE event, even if the MCPR margin was within the design 

parameters established by the ARTS modification. 

The inspectors reviewed activities associated with GE’s design interface 

with ARTS licensees regarding RBM operability during a potential RWE 

event at ARTS plants. The inspectors assessed GE’s interaction with 

Detroit Edison Company (DECo) for Fermi Atomic Nuclear Plant Unit 2 

(Fermi); with Southern Nuclear Operating Company (SNOC) for Edwin I. 

Hatch Nuclear Plant, Units 1 and 2 (Hatch); with Carolina Power & Light 

Company (CP&L) for Brunswick Steam Electric Plant, Units 1 and 2 

(Brunswick); with Philadelphia Electric Company for Limerick Generating 

Station, Units 1 and 2 (Limerick); with Nebraska Public Power District 

for Cooper Nuclear Station (Cooper); with IES Utilities (IES) for Duane 

Arnold Energy Center (Duane Arnold); and with Boston Edison Company for 

Pilgrim Nuclear Power Station, Unit 1 (Pilgrim). The inspectors’ review 

of factors pertaining to RBM operability at ARTS plants is summarized 

bel ow: 

(1) FEW1 

Cycle 3 commenced June 1, 1991. Cycle 4 commenced November 1, 1992 

and concluded with the turbine trip event in December 1993. 

April 1992, DECo audited GE’s activities regarding fuel design control 

In 

3 

-49- 
for Fermi Cycle 4 (DECo Report 92-018, dated May 6, 1992). DECo 

determined that based on GE's RWE analysis there was a potential for 

the fuel cladding exceeding its MOP limits if the RBM was not operable 

and control rod patterns deviated from GE's recommended normal 

operation cycle patterns. GE had not informed DECo that the MOP 

limits at Fermi could be exceeded if the RBM was bypassed, and that 

the Fermi TS 3/4.1.4.3 (in accordance with the ARTS modification) 

allowed bypassing of the RBM. 

Cycle 3 and 4 cores to either affirm or refute the validity of the RWE 

analysis with respect to MOP limits in intermediate control rod 

patterns, and whether the Fermi TS 3/4.1.4.3 "contained adequate 

ope rat i on a1 rest r a i n t s . I' 

DECo recommended that GE analyze both 

GE analyzed Fermi Cycles 3 and 4 and informed DECo that for Cycle 3 

the MOP limits were not exceeded. 

Cycle 4, the MOP limits for GE-11 fuel would be exceeded but that the 

strain limits would not be exceeded because of a substantial 

conservative margin between the MOP limits and the strain limits. GE 

concluded that for Cycle 4 the RBM did not need to be operable, and 

that the TS 3/4.1.4.3 RBM operational requirements were adequate. 

GE also informed DECo that, for 

DECo audited GE on December 7 through 10, 1993 (DECo Report QA-94- 

1006, dated January 10, 1994), and again determined that GE had not 

adequately communicated design changes to DECo. 

DECo that Fermi's TS basis for RBM operability required consideration 

of the MOP limits. The inspectors determined that the Fermi reload 

analysis for Cycle 4 provided in the SRLR 23A7075 Revision 0, dated 

April 1992, did not address RBM operational requirements for not 

exceeding the MOP limits. In November 1994, as part of its response 

to DECo's audit observation, GE issued letters to inform all ARTS 

licensees that they should consider the strain limits and associated 

MOP limits when evaluating RBM operability for the ARTS modification. 

GE had not informed 

For Cycle 5, GE informed DECo that the MOP limits could be exceeded. 

In preparation for Cycle 5, GE's RWE analyses, dated November 11, 

1993, and June 22, 1994, indicated that the RBM needed to be operable 

to prevent the MOP limits from being exceeded. Cycle 5 commenced on 

December 17, 1994. In May 1995, DECo requested GE to further evaluate 

the need for RBM operability. On August 22, 1995, during Cycle 5, GE 

completed an RWE analysis using the actual core loading and control 

rod patterns and concluded that the MOP limits were not exceeded and 

that the RBM did not need to be operable. 

For Cycle 6, GE informed DECo that the RBM was required to be 

operable. 

MOP limits for fuel GE-11 would be exceeded within the error cell 

during an RWE event if the RBM was not operable. 

6, issued in November 1996, stated: "At least one RBM channel must be 

operable when moving rods in order to protect for mechanical overpower 

limits." The inspectors considered the SRLR to be inadequate in that 

it did not reference the 30% power level for RBM operability and 

addressed operability of only one RBM channel. 

Based on the PANACEA analyses, GE informed DECo that the 

The SRLR for Cycle 

4 

-50- 
During the inspection, GE informed the inspectors that the strain 

limits were not exceeded at Fermi during the above operational cycles. 

However, the inspectors concluded'that GE did not adequately inform 

DECo of the need to consider the strain limits and the associated MOP 

limits in addition to the MCPR limits when evaluating RBM operability 

for an RWE event. 

appropriate TS revisions for Cycles 5 and 6 are addressed in section 

3.2 b of this report. 

DECo actions regarding implementation of 

(2) HATCH 

For Hatch Unit 1 Cycles 16 and 17, and Unit 2 Cycles 13 and 14, GE 

informed SNOC that the MOP limits would be exceeded during an RWE 

event if the RBM was not operable, A GE memorandum to SNOC, dated 

October 31, 1994, stated that the "1 percent plastic strain is met, if 

one channel of the RBM remains operable." In addition, on October 31, 

1994, GE informed SNOC by letter that they should consider the strain 

limits and associated MOP limits when evaluating RBM operability for 

the ARTS modification. 

Hatch 2 Cycles 13 and 14 informed Hatch of necessary protective 

measures but failed to clearly address the conditions under which the 

RBM should be operable. During the inspection, GE informed the 

inspectors that based on its analyses the MOP limits had been exceeded 

for Hatch Unit 1 Cycles 16 and 17, Unit 2 Cycle 13, and Cycle 14. The 

strain limits were not exceeded during these cycles. 

regarding implementation of appropriate TS revisions are addressed in 

section 3.2 b of this report. 

GE's SRLR for Hatch 1 Cycle 16 and 17, and 

SNOC's actions 

(3) PILGRIH, LIMERICK, COOPER 

Based on the SRLRs for these plants, the RBM did not need to be 

operable for Pilgrim Cycle 12, Limerick Unit 1 Cycles 6 and 7, 

Limerick Unit 2 Cycle 4 and 5, and Co~per Cycle 18. 

determined that the MOP limits were not exceeded during these cycles. 

In November 1994, GE issued letters to these licensees indicating that 

they should consider the strain limits and associated MOP limits when 

evaluating REM operability for the ARTS modification. 

The inspectors 

(4) BRUNSWICK AND DUANE ARNOLD 

In January 1995, GE conducted an RWE analyses for the Brunswick Cycle 

10 and Duane Arnold Cycle 14 and determined that the MOP limits would 

be exceeded if the licensees bypassed the RBMs. 

inspectors review of GE's design documents, GE did not inform these 

licensees that the MOP limits would be exceeded. Based on GE's RUE 

analysis, the MOP limits were exceeded at Brunswick. On April 1, 

1997, CP&L issued a 50.72 notification (Event No. 32058) regarding RBM 

operabi 1 i ty . 

Based on the 

In March 1997 (before the NRC inspection), GE reviewed reload design 

documents to determine which ARTS plant cycles resulted in exceeding 

the strain limits. Based on its review of prior RWE analyses, GE 

5 

-5 1- 
notified CP&L and IES by letter dated March 11, 1997, that the 

Brunswick Cycle 10 and Duane Arnold Cycle 14 GE-10 fuel had exceeded 

the MOP limits. 

inspection that the MOP limits were exceeded for these two plants. 

GE did not inform the inspectors during the 

Subsequent to the inspection, GE performed a cycle-specific analysis 

for both plants. In a letter dated April 25, 1997, GE informed the 

NRC that the strain limits were not exceeded at Brunswick and Duane 

Arnold. On May 9, 1997, the inspectors contacted GE staff by 

telephone to ask why licensees were not informed in January 1995, and 

why the inspectors were not informed during the March 1997 inspection 

that based on GE’s RWE analyses, Brunswick and Duane Arnold had 

exceeded the MOP limits. GE stated that it did not inform the 

licensees or the NRC because GE’s “generic” RWE analysis had indicated 

that the strain limits would not be exceeded. CP&L and IES’s actions 

regarding implementation of appropriate TS revisions are addressed in 

section 3.2 b of this report. 

Based on the above review, the inspectors determined that GE did not 

adequately inform 1 icensee design organizations implementing the ARTS 

modification of the need to consider the strain limits and the associated 

MOP limits in addition to the MCPR limits when evaluating RBM operability 

for an RWE event. 

requirements for RBM operability with regard to the MOP limits. 

inadequate interface between GE and ARTS licensees contributed to (1) the 

failure of licensees to ensure through their TS that the RBM was operable 

to protect fuel cladding at applicable plants, and (2) occasions during 

Fermi Cycles 4 and 5, Hatch Unit 1 Cycles 16 and 17, Hatch Unit 2 Cycles 

13 and 14, Brunswick Cycle 10, and Duane Arnold Cycle 14 when based on 

GE’s RWE analyses the fuel cladding had exceeded its MOP limits and had 

the potential of exceeding its strain limits. This constitutes 

Nonconformance 99900003/97-01-01. 

GE’s SRLRs for ARTS plants did not adequately address 

This 

In its March 19, 1997, letter to the NRC, GE stated that its 

“communications to licensees in the SRLR have not been adequate.” GE 

also stated that, in part, it (1) had sent letters to all customers 

apprising them that the RBM is required to ensure that the MOP limits 

were not exceeded, (2) had revised technical design procedure (TDP) 0035 

to clarify required utility communications, and (3) planned to review the 

SRLR format and revise the GESTAR I1 standard format, if necessary, and 

make further revisions to TDP-0035 if necessary to be consistent with the 

final SRLR format. 

c. Conclusions 

The strain limits were not exceeded at the ARTS plants. 

interface with licensees was weak, as noted in the nonconformance herein. 

GE did not adequately inform ARTS licensees of the conditions for RBM 

operability, contributing to the ARTS licensees’ failure to implement RBM 

protection for fuel during potential RWE events. SRLRs did not clearly 

identify conditions for RBM operability. Based on GE’s RWE analyses, the 

GE’s design 

6 

-52- 
fuel cladding had exceeded its MOP limits and had the potential of 

exceeding its strain limits at Fermi, Hatch, Brunswick, and Duane Arnold. 

3.2 TECHNICAL SPECIFICATIONS (TS) ACTIONS 

a. InsDection ScoDe 

The inspectors examined TS actions taken by GE and ARTS licensees to 

ensure RBM operability during an RWE event to protect exceeding the 

strain limits and associated MOP limits. 

b. Observations and Findinqs 

DECo audited GE in 1992 and 1993 and observed GE's failure to inform DECo 

(and other ARTS licensees) that the MOP limits would be exceeded if the 

RBM was not operable during an RWE event. In November 1994, GE informed 

all ARTS licensees of the need to consider RBM operability to not exceed 

strain 1 imits. 

During Cycle 5, GE informed DECo that the RBM must be operable. 

Section 10 of the Cycle 5 SRLR, dated December 1993, GE stated that "at 

least one channel of the RBM must be operable when moving rods in order 

to meet the MOP limits for an RWE event." 

June 22, 1994, regarding the Cycle 5 analysis stated: "GE still 

recommends that at least one RBM channel be operable when moving rods." 

However, DECo did not revise the Fermi TS; rather, in May 1995 DECo 

requested GE to further evaluate the need for RBM operability. On August 

22, 1995, GE concluded that the MOP limits were not exceeded during Cycle 5. 

In 

GE's memorandum to DECo, dated 

In preparation for Cycle 6, OECo addressed RBM operability by placing a 

statement in the core operating limits report (COLR). The NRC resident 

inspectors became aware of DECo's use of the COLR in this manner and on 

November 22, 1996, the NRC informed DECo by telephone that the addition 

of this statement to the COLR without a corresponding request to amend 

the TS would not comply with the requirements of 10 CFR 50.36. The NRC 

permitted DECo to start up using administrative controls after the 

licensee submitted the TS amendment request on December 2, 1996, to 

require RBM operability above 30 percent power. 

112 to facility operating license No. NPF-43) was issued on May 15, 1997. 

The TS amendment (No. 

For Hatch Unit 1 Cycles 16 and 17, and Unit 2 Cycles 13 and 14, GE 

informed SNOC that the MOP limits would be exceeded during an RWE event 

if the RBM was not operable. A GE memorandum to SNOC dated October 31, 

1994, stated that the "1 percent plastic strain is met, if one channel of 

the RBM remains operable." SNOC did not revise the Hatch TS. 

In January 1995, GE's RWE analysis indicated that the Brunswick Cycle 10 

and Duane Arnold Cycle 14 GE-10 fuel would exceed the MOP limits. Based 

on documents provided to the inspectors, GE did not inform licensees of 

the potential for exceeding the MOP limits. In March 1997, GE informed 

7 

-53- 
CP&L and IES that based on its RWE analyses the fuel at Brunswick and 

Duane Arnold had exceeded its MOP limits. CP&L and IES did not revise 

the Brunswick and Duane Arnold TS. 

The inspectors asked GE why they did not explicitly inform ARTS licensees* 

to revise their TS to ensure RBM operability for an RWE event. On May 1, 

1997, during a telephone conversation with the inspector, GE agreed to 

inform licensees of the need to examine or consider revisions to their 

TS, as applicable, with respect to RBM operability. 

c. Conclusions 

Fermi and Hatch apparently knew that the RBM must be operable for a 

potential RWE but did not revise their TS to require RBM operability for 

an RWE event. Other ARTS licensees, including Brunswick and Duane 

Arnold, were informed by GE in November 1994 of the need to consider RBM 

operability to not exceed strain limits but did not revise their TS. 

GE’s failure to explicitly inform licensees to revise their TS to ensure 

RBM operability during an RWE is considered a weakness. 

3.3 DESIGN CONTROL 

a. Insoection Scope 

The inspectors reviewed the adequacy of GE’s RWE analyses and GE’s 

assessment of its design control program. 

b. Observations and Findinos 

The inspectors determined that GE modified peaking factors when the MOP 

limits were exceeded in the RWE analyses, and applied alternate rod 

patterns in addition to normal rod patterns in the RWE analyses without 

documented instructions or procedures. The inspectors also determined 

that GE changed design parameters for its peak cladding temperature (PCT) 

analysis by revising the theoretical fuel pellet density values without 

procedures for tracking the removal of conservatisms in the PCT analysis. 

GE’s use of the incorrect fuel pellet densification value had resulted in 

a lower peak cladding temperature. When GE corrected the densification 

value, the Loss of Coolant Accident analysis of the BWR-6 plant indicated 

an unacceptable increase of the PCT. 

values to reduce the PCT without issuing any documented instructions or 

procedures for using the revised density values. 

Nonconformance 99900003/97-01-02. 

GE revised the theoretical density 

This constitutes 

Subsequent to the inspection, in its March 19, 1997, letter to the NRC, 

GE stated that, as part of its corrective actions, it (1) revised TOP 

0035 to improve its communication with licensees on the need for RBM 

operabil i ty, (2) imp1 emented an expl ici t requirement for MOPS 

calculations for all fuel types, (3) plans to review TDPs for other areas 

of improvement, e.g., added guidance on local peaking factor adjustments, 

and (4) plans to document conservatisms used in its analyses. 

8 

-54- 
DECo audited GE in 1992 and 1993 and observed several deficiencies 

regarding design control, including GE’s failure to inform DECo (and 

other ARTS licensees) that the MOP limits would be exceeded if the RBM 

was not operable during an RWE event. 

took corrective actions for specific deficiencies but did not conduct 

followup action, including reaudit of the design control area, to 

determine the effectiveness of the program. This constitutes 

Nonconformance 99900003/97-01-03. 

it planned to perform an internal audit to address the effectiveness of 

its design program. In its March 19, 1997 letter to the NRC, GE stated 

that its corrective actions included (1) strengthening its requirements 

for annual adequacy reviews of TDPs; planning to perform adequacy reviews 

during the second quarter of 1997, and (2) performing a QA followup audit 

of a revised adequacy review process scheduled for late 1997. 

The inspectors determined that GE 

During the inspection, GE stated that 

The inspectors observed that GE’s generic analysis of MOP limits for RWE 

events at ARTS plants may not be valid for cores containing GE 8x8 fuel 

or GE-10, -11, -12, and -13 fuel not included in the 1987 assessment of 

generic MOP limits. 

began conducting cycle-specific RWE analyses for all plants and all fuel 

types. 

GE informed the inspectors that in March 1997 it 

No immediate safety concerns were identified. 

The inspectors reviewed GE’s PANACEA analysis and observed that GE 

evaluated UO, rods rather than Gadolinium (Gd) rods in an error cell for 

an RWE event. The generic GESTR-M strain analysis indicated that the Gd 

rods were more limiting than the UO rods. Consideration of the UO, rods 

but not the Gd rods in the error ce51 for the PANACEA analyses may not 

identify the limiting rod for the strain limits. 

observed that the PANACEA analyses considered planned rod patterns for 

certain cycles but did not include possible alternate rod patterns. 

of only planned rod patterns for certain analyses may not identify the 

most limiting operation within the error cell for the analyzed cycle 

because alternate rod patterns may be used. The inspectors determined 

that exclusion of Gd rods and alternate rod patterns was a weakness in 

GE’s PANACEA analyses. 

The inspectors also 

Use 

The inspectors assessed the initial conditions assumed in cycle-specific 

accident analyses. One example reviewed involved a Fermi turbine trip 

event with nominal reactor steam dome pressure assumed for the start of 

the transient. The inspectors examined whether the safety analysis was 

bounding and the MCPR limit was sufficient if a transient would be 

initiated from a higher pressure within the bounds of the TS operating 

limits. 

sensitivity study to evaluate the effect on the thermal margin of 

pressure variations of plus or minus 20 psi, including cases with GE-9 

and GE-13 fuel. 

delta critical power ratio (CPR)/initial critical power ratio (ICPR) over 

this pressure range. CPR/ICPR. 

assumptions regarding the initial conditions of the analyses and whether 

operating 1 imits derived from the analyses provide bounding protection 

for all operating conditions permitted by the plant TS. 

In response to questions by the inspectors, GE performed a 

The results indicated a pressure sensitivity of a small 

Higher initial pressure resulted in a lower delta 

In reviewing safety analysis methods, the NRC examines 

Since the 

9 

-55- 
C. 

3.4 

thermal margin is dependent on several operating parameters that do not 

vary independent of pressure, thermal margin sensitivity to pressure 

reduction is not necessarily indicative of the thermal margin sensitivity 

to overall off-nominal plant operating conditions for 1 arge pressure 

increments. No concerns were identified. 

The inspectors observed that in late February 1997, Tennessee Valley 

Authority determined that the Browns Ferry Nuclear Plant Emergency 

Operating Procedures (EOPs) revision for GE 9x9 fuel had not considered 

all the input parameters affected by the new fuel design. The input 

parameters are specific to fuel type and would be used by the licensee to 

calculate plant specific curves or limits to maintain Emergency Procedure 

Guide1 ines (EPGs) specified action levels. Input parameters for EOP 

calculations are contained in Appendix C of NE90-31331, Revision 4, 

issued March 1987. At the time of issuance of revision 4, GE 9x9 fuel 

was not in use by any 1 icensee. 

1 icensees apparently did not recognize that certain Parameters identified 

as ”generic data” in EPG Appendix C were affected by fuel type. The 

parameters were four steam-cooling-related input parameters and two 

shutdown boron weight-related input parameters. In response to the 

event, GE issued Services Information Letter (SIL) 529, Supplement 1, on 

March 14, 1997, informing BWR customers of the fuel type-specific nature 

of the input parameters and recommending that potentially affected plants 

reevaluate their EPG calculations if the fuel design has changed since 

the Appendix C calculations. 

calculations which may be impacted by revised input parameters. 

Upon incorporation of 9x9 fuel, 

The SIL also provided a list of 

Conclusions 

GE did not have documented instructions or procedures to control 

modifications in RWE analyses when MOP limits were exceeded. GE’s 

generic MOP limits evaluations need to be replaced with cycle-specific 

analyses. The PANACEA fuel strain analysis was weak. 

Entrance and Exit Meetinss 

In the entrance meeting on March 10, 1997, the NRC inspectors discussed 

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

established interactions with GE management. In the exit meeting on 

March 14, 1997, the inspectors discussed their findings and observations*. 

10 

-56- 
4 PARTIAL lIST OF PERSONNEL CONTACTED 

General Electric Nucl ear Eneray 

Chris Monetta, Manager, Nuclear Quality Assurance 

Glen Watford, Manager, Nuclear Fuel Engineering 

dames K1 approth , Manager, Product Definition 

Steve Congdon, Manager, Nuclear Techno1 ogy 

Chuck Papandrea, Manager, SC Quality 

Jerry Potts, Manager, Fuel Performance 

Kevin Theriault, Manager, Fuel QA 

Anne Sul 1 i van, Manager, Fuel Engineer, QA 

Paul Sick, Manager, Nuclear Qual i ty Assurance 

El wood Mob1 ey, Manager , Qual i ty Assurance 

Rendy Hannel t, Manager, Nuclear Fuel Projects 

Ralph Reda, Manager, Fuel & Facility Licensing 

Caroline Smith, Sr. Engineer, Fuel & Facility Licensing 

John Embley, Program Manager, Fuel & Facility Licensing 

Jim Rash, Sr. Engineer, Fuel & Facility Licensing 

Robert Rand, Principal Engineer, Nuclear Fuel 

Fran Bo1 ger , Pr i nci pal Engineer, Nucl ear Fuel 

Barry R. Fi scher, Technical Program Engineer Nuclear Fuel 

ITEMS OPENED, CLOSED, AND DISCUSSED 

ODened 

99900003/97-01-01 

99900003/97-01-02 

99900003/97-01-03 

Para 3.1 b NON 'inadequate design interface 

Para 3.3 b NON inadequate procedures 

Para 3.3 b NON inadequate assessment of 

design effectiveness 

ACRONYMS USED 

ARTS 

GE 

RBM 

RUE 

RUE Event 

MCPR 

TS 

SRLR 

MOP 

TDP 

CQLR 

PCT 

Gd 

UO 

ICPR 

EPOs 

EPGs 

SIL 

CPk 

Average Power Range Monitor-Rod Olock Moni tor-Technica 

General Electric 

Rod Block Monitor 

Rod Withdrawal Error 

Rod Withdrawal Error above 30 percent power 

Minimum Critical Power Ratio 

Technical Specifications 

Supplemental Reload Licensing Report 

Mechanical Over Power 

Technical Design Procedure 

Core Operating Limits Report 

Peak Cladding Temperature 

Gadol i n i urn 

Uranium Dioxide 

Critical Power Ratio 

Initial Critical Power Ratio 

Emergency Operating Procedures 

Emergency Procedure Guide1 ines 

Services Information Letter 

Spec cations 

11 

-57- 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WAS!:INGTON, D.C. a05650001 

April 22, 1997 

Mr. John M. Leslie, President 

SynTech Products Corporation 

520 East Woodruff Avenue 

Toledo, Ohio 43624 

SUBJECT: NRC INSPECTION REPORT 99901313/97-01 

Dear Mr. Leslie: 

On February 4-6, 1997, the U . S. Nuclear Regul atory Commi ssi on (NRC) conducted 

an inspection at your Woodruff Avenue facility in Toledo, Ohio. The enclosed 

report presents the results of that inspection. 

The NRC inspection team evaluated the commercial quality program that SynTech 

established and implemented to control the manufacture and supply of various 

cleaning and maintenance chemical products. Although SynTech supplies several 

of its products to NRC-licensed nuclear power plants, the inspectors 

determined that those products are not basic components as defined in Section 

21.3 of Part 21 of Title 10 of the Code of Federal Resulations (10 CFR Part 

21) because, for facilities licensed under 10 CFR Part 50, the products are 

not designed or manufactured under a quality assurance program meeting the 

provisions of 10 CFR Part 50, Appendix By nor are they treated as basic 

components for procurement purposes by the NRC licensees to whom you supply 

them. 

The inspection consisted of an examination of selected procedures, purchase 

orders, test reports, certificates of compliance and other associated records, 

interviews with personnel and observations by the inspectors. No violation of 

NRC requirements was identified. Also, during this inspection, the inspectors 

reviewed the circumstances surrounding SynTech’s submission to Consumers Power 

Company (Palisades) of what appeared to be a Detroit Edison Company chemical 

test report, but which reported data of indeterminate origin and cited a 

sample lot for which Detroit Edison or its subsidiary, Utility Technical 

Services, had no record of testing. 

avail able to the inspectors suggested a fraudulent SynTech report. However, 

your explanation of the circumstarves surrounding the creation of the report 

in its unusual form was plausible, if not conclusive, and was sapported by 

other information obtained by the inspectors. Although you couid not offer a 

satisfactory explanation for the report’s not being effectively screened 

through management review before issue, the incident appeared t:, have been 

isolated. The inspectors could not conclusively determine that there was any 

deliberate intent to mislead the recipient of the report and found no evidence 

of a pattern of similar activity. 

Prior to the inspection, the information 

-58- 
J. Leslie -2- 

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

this letter and its enclosures will be placed in the NRC's Public Document 

Room (PDR) . 

Sincerely, 

Robert M. Gallo, Chief 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Docket No. 99901313 

Enclosures: 1. Inspection Report 99901313/97-01 

-59- 
U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

Report No.: 99901313/97-01 

Organization: SynTech Products Corporation 

520 East Woodruff Avenue 

Toledo, Ohio 43624 

Contact: D. Moore, Quality Assurance Manager 

(419) 241-1215 

Nuclear Industry Manufactures and suppl ies industrial cleaners, 

Act i vi ty : solvents, and other maintenance products such as 

penetrating oils, cutting fluid, gasket/paint 

strippers, etc., in bulk quantities or in aerosol 

cans. 

Dates: 

Inspectors: 

Approved by: 

February 4-6, 1997 

Stephen D. Alexander, Reactor Engineer (Team Leader) 

Kamalakar R. Naidu, Senior Reactor Engineer 

Gregory C. Cwalina, Chief 

Vendor Inspection Section 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Enclosure 1 

-60- 
1 INSPECTION SUHMARY 

During this inspection of SynTech Products Corporation, the NRC inspectors 

reviewed selected procurement documents, test certificates and other 

documents, examined products and storage facilities, and interviewed SynTech 

staff regarding the supply of industrial cleaning and maintenance products to 

NRC-licensed nuclear power plants. The inspection bases were Part 21 of Title 

10 of the Code of Federal Reaulations (10 CFR Part 21) and Appendix B to 10 

CFR Part 50. 

Based on a SynTech purchase order (PO) to a licensee-operated testing 

laboratory for analysis of a SynTech product that invoked 10 CFR Part 50, 

Appendix B, and 10 CFR Part 21, SynTech may have supplied some of its products 

as basic components as defined in 10 CFR 21.3. However, the inspectors 

subsequently determined that SynTech products are not designed (devel oped) or 

manufactured under a 10 CFR Part 50, Appendix By quality assurance (PA) 

program and SynTech has not supplied, nor does it currently supply its 

products as basic components. A1 though many of SynTech’s nuclear plant 

customers required certificates of conformance (CoC) to their PO requirements 

and a certified chemical contaminant analysis test report traceable to the 

batches or lots of the products supplied, the inspectors found no instances in 

which NRC licensees imposed nuclear-unique design or QA requirements on 

SynTech or invoked 10 CFR Part 21 in their procurement documents. 

On the basis of a review of SynTech’s commercial quality controls and review 

of independent analyses of SynTech products, the inspectors concluded that the 

contaminant content of SynTech products is variable enough to require batchor lot-traceable certified chemical contaminant analysis reports. Such 

reports would confirm that levels of undesirable substances such as halides, 

sulfur and heavy metals are within acceptable limits before using SynTech 

products on safety-related equipment or allowing them to come in contact with 

safety-re1 ated systems. 

The inspectors a1 so reviewed the circumstances surrounding SynTech’s 1995 

submission of what appeared to be a Detroit Edison (DE) Company chemical 

analysis report, but which reported data of indeterminate origin and cited a 

sample lot for which DE or its subsidiary, Utility Technical Services (UTS), 

had no record of testing. Prior to the inspection, the information available 

to the inspectors suggested a fraudulent SynTech report. However, SynTech’s 

explanation of the circumstances surrounding the creation of the report in its 

unusual form was plausible, if not conclusive, and was supported by other 

information obtained by the inspectors. 

satisfactory explanation for the report’s not being effectively screened 

through management review before issue, the incident appeared to have been 

isolated. 

deliberate intent to mislead the recipient of the report and found no evidence 

of a pattern of similar activity. 

Although SynTech could not offer a 

The inspectors could not conclusively determine that there was any 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

This was the first NRC inspection of SynTech. 

2 

-61- 
3 INSPECTION FINDINGS AND OTHER COMMENTS 

3.1 Commerci a1 Qual i tv Control s/Chemical Contaminant Control 

a. ScoDe 

In order to evaluate SynTech commercial qual ity controls, particularly 

regarding control of chemical contaminants in SynTech products sold to NRClicensed facilities, the inspectors toured the SynTech facility, interviewed 

personnel , examined raw materials and finished products in various forms and 

in various stages of preparation. 

shipment either to customers or subcontractor manufacturing or packaging facilities. The inspectors also reviewed numerous chemical analysis reports 

on the contaminant content of SynTech products as well as correspondence with 

the materials engineer at General Electric Nuclear Energy (GE NE) regarding 

the eligibility of SynTech products for GE NE'S "nuclear grade" designation. 

Also examined were products awaiting 

b. Findinas and Observations 

SynTech manufactures several industrial cleaning and maintenance chemical 

products including the following which are used in nuclear power plants. 

SynTech supplies these products in aerosol cans, in bulk containers, or both: 

Easy Tap: Cutting and tapping fluid (aerosol and squeeze bottles) 

Parts Kleen: Solvent and degreaser (aerosol and bulk containers) 

N-Grade : 

NOC : (Non-organic C1 eaner) A1 1 purpose cleaner 

Rust Buster 11: Penetrating oil 

Titanic C: A1 1 purpose floor cleaner (bul k) 

Touch It Up: All purpose cleaner (aerosol and bulk) 

Gasket and paint stripper (aerosol) 

SynTech manufactures (blends) a batch of a product supplied in bulk from raw 

materials or receives finished product in bulk form from one of its suppliers 

or contract manufacturers. 

subcontractor 1 aboratory, currently Phil adel phi a Electric Company's (PECO 

Energy's) Wayne, Pennsylvania, laboratory, for chemical analysis to determine 

if the levels of total halogens/hal ides, sulfur, and other undesirable 

elements, mostly heavy metals (e.g., As, Bi, Sn, Pb, Hg, etc.), are within 

specified acceptance levels. For products in aerosol cans, the finished 

products are sent in bulk form to a SynTech subcontractor for packaging in the 

aerosol cans. Samples from numbered lots of aerosol cans (which are supposed 

to be filled from one batch of bulk product) are also sent to PECO's 

laboratory for analysis. 

after repackaging because SynTech has found that the packaging process or cans 

themselves may introduce some of the contaminants of concern. 

Batches or lots of products routinely sold to nuclear plants as well as other 

customers that are awaiting results of chemical analysis are stored in a 

locked cage designated the "Commercial Quality" area with QA hold tags or 

It then sends a sample of that numbered batch to ii 

SynTech has samples of aerosol can lots analyzed 

3 

-62- 
labels affixed. 

levels of the contaminants of concern are within acceptable limits, the 

batches of chemicals are released for shipment to nuclear power plants. 

the levels are above the acceptance limits, the products are removed from the 

“Commercial Quality Area” and stored in general storage for sale to nonnuclear faci 1 it i es. 

If the results of the chemical analysis indicate that the 

If 

As discussed in correspondence with and from GE NE’S materials engineer, 

SynTech established the acceptance 1 imits to maintain a “nuclear grade” rating 

in accordance with GE NE Material Specification D5OYP12 and Appendix A of 

NEDE-31295P, GE’s “BWR Operator’s Manual for Materials and Processes,” June, 

1986. However, individual licensees sometimes specify their own limits, 

either explicitly or by reference to a material standard or specification 

(e.9.) those of the American Society for Testing and Materials (ASTM)). For 

nuclear customers, SynTech stated that it checks that the batch/lot analysis 

report indicates the customer’s specific limits are met before shipping the 

product unless specifically requested otherwise by the customer. Review of 

numerous lot analysis reports on file at SynTech indicated that the amounts of 

contaminants or undesirable substances such as ha1 ides, sulfur, and heavy 

metals (e.g., mercury, bismuth, arsenic, lead, tin, etc.) are somewhat 

variable, sometimes being marginal or even exceeding GE NE nuclear grade or 

customer-specific limits. 

SynTech brought two such instances to the inspectors’ attention. In one case, 

SynTech had discovered that tin was leaching out of aerosol cans into the 

product. This was corrected by SynTech requiring their aerosol can packaging 

subcontractor henceforth to use cans with a non-metallic coating inside. 

Subsequent analysis reports indicated that this measure corrected the problem 

and did not introduce additional problems such as adding halides. In the 

other instance, the NRC licensee, Philadelphia Electric Company (PECo) 

requested that SynTech ship a batch of product for which the analysis results 

had not yet been obtained because it was PECo’s laboratory that was currently 

performing the contaminant analyses for SynTech. As Murphy’s Law would have 

it, sure enough, this particular batch of the product contained excessive 

sulfur and was of course captured and sent back to SynTech. Shipping records 

reviewed by the inspectors confirmed this and also indicated that, as stated 

by SynTech, it routinely provides copies of the batch/lot chemical analysis 

test reports to nuclear customers whether or not they are required. 

In the Commercial Quality Area, the inspectors noted information on the tags 

affixed to 55-gallon drums containing Titanic Cy NOC, and Parts Kleen, and a 

tag affixed to a box containing Part Kteen aerosol cans, and reviewed the 

documents associated with their manufacture. The inspectors observed that the 

access to this storage area was controlled with a lock and key. 

One of the products, Titanic C floor cleaner, was manufactured on February 28, 

1996, and stored in a 55-gallon drum. A tag attached to this drum identified 

the product as belonging to Lot 095, and Batch 1089. The batch manufacturing 

sheet for Lot 095, dated February 28, 1996, indicated the various ingredients 

that were used to manufacture this lot of Titanic C. The SynTech QA manager 

stated that all this material was stored in an area designated for commercial 

quality products. The inspectors observed some of the ingredients used in the 

4 

-63- 
manufacture of Titanic C stored in their original bags in the Commercial 

Quality Area. 

its current independent testing laboratory, to analyze a sample of Lot 095 for 

contaminant content including halogens, sulfur and heavy metals. 

SynTech issued PO 15201 on February 4, 1997, to PECO Energy, 

The tag attached to a 55-gallon drum of NOC (non-organic cleaner) was marked 

Lot NOC 01, Batch 10994. The tag indicated that a sample of the contents of 

the drum had been tested. SynTech received certification on October 3, 1996, 

that the levels of halogens, sulfur and heavy metals were within the 

acceptable limits. 

A 55-gallon drum of Lot 0408942 of Parts Kleen, blended for SynTech by Chem 

Central in Toledo had a hold tag attached to it indicating that it was 

awaiting the results of chemical analysis of a sample of it that had been sent 

to one of SynTech’s contract laboratories for chemical analysis. laboratory report, dated August 23, 1994, indicated the content of halogens, 

sulfur and heavy metals were within the usual prescribed limits, but the hold 

tag was left on so that the report would be checked against specific customer 

requirements prior to shipment. 

The 

A hold tag was attached to a pallet of 90 cases of Parts Kleen aerosol cans, 

Lot 611130, in the commercial quality area packaged for SynTech by American 

Jetway of Detroit, Michigan. 

can of Lot No. 6m30 to the PECO Energy laboratory for chemical analysis. 

The hold tag indicated that SynTech had sent one 

The inspectors reviewed the documents on a shipment of 1 dozen aerosol cans of 

Touch It Up for Niagara Mohawk Power Company’s Nine Mile Point site. 

documents included a SynTech CoC that certified that the contents were from 

Lot X4K31 which conformed to Specification 15586/94-004215, and the halides, 

sulphur and heavy metals were within acceptable limits. 

certification was based on a certified laboratory report providing the 

chemical contaminant analysis results. 

The 

SynTech’s 

c. Conclusions 

SynTech has imp1 emented commerci a1 qual i ty control s to provide reasonable 

assurance that its products shipped to NRC-1 icensed facilities typically contain less than the maximum acceptable levels of chemical contaminants of 

concern. SynTech had established measures to routinely control the release of 

products to nuclear plant customers pending receipt of certified chemical 

contaminant analysis reports from SynTech’s contract laboratory. According to 

SynTech, this had become their standard practice, regardless of whether such 

reports were required by a particular customer. No further concerns were 

identified in this area. 

3.2 Review of Procurement Documents. 

a. Scope 

The inspectors reviewed SynTech customer procurement document files to 

determine if NRC-licensees had imposed the requirements of 10 CFR 50, 

Appendix 8, or 10 CFR Part 21 on SynTech, i.e., if licensees had treated 

5 

-64- 
SynTech products as basic components. In addition, the inspectors examined 

selected f i 1 es of SynTech procurement documents to determine i f SynTech had 

passed these requirements to its suppliers of materials and services. 

b. Observation and Findinqs 

Prior to the inspection, the inspectors had obtained a copy of SynTech PO No. 

10022, dated January 28, 1993, and a letter accompanying a sample of Lot A050- 

303 of "Touch-It-Up" to a subsidiary of Detroit Edison (DE), Utility Technical 

Services (UTS), which stated that "Appendix 8 and Part 21 apply, 10 CFR 50" 

[sic]. This suggested that SynTech may have supplied some of its products to 

nuclear power plants as basic components as defined in 10 CFR 21.3. According 

to SynTech and documents on file at SynTech, when this PO was issued, SynTech 

had been contracting DE'S Technical and Engineering Services laboratory for 

several years to perform the contaminant analyses on its products being 

shipped to nuclear plants. 

of letters on DE letterhead, had covered multiple products and multiple 

samples. The inspectors found three such reports in SynTech's files, dated 

March 26, 1990, December 6, 1990, and January 15, 1992. In addition, there 

had historically been significant time delays in obtaining results from DE. 

However, the DE lab reports, notably in the form 

SynTech stated that at about the same time that DE'S Technical and Engineering 

Services laboratory was put under the new DE subsidiary organization, UTS, 

SynTech was advised by the QA manager at UTS that if SynTech invoked 10 CFR 

Part 50, Appendix B, and Part 21 in the PO, then UTS would be able to perform 

the analyses expeditiously in addition to the added benefits of a certified 

analysis report for one lot of one product at a time. 

these unfamiliar requirements, SynTech cited them incorrectly, the inspectors 

observed that subsequently, UTS test reports were for one lot of one product 

each and contained the statement "the work was performed in accordance with 

the Technical and Engineering Services quality assurance program." The 

inspectors identified no other instances in which SynTech had invoked 10 CFR 

Part 21 or imposed 10 CFR Part 50, Appendix B, on any of its suppliers or 

subcontractors. 

Although in translating 

The inspectors reviewed POs issued to SynTech by several NRC licensees 

including Boston Edison Company, Centerior Services Company, Niagara Mohawk 

Company, Detroit Edi son Company, Consumers Power Company, and Commonwealth 

Edison Company between 1992 and 1996 for the supply of various quantities of 

different SynTech products. 

customers required CoCs to their PO requirements or a certified chemical 

contaminant analysis test report, traceable to the batches or lots of the 

products supplied, the inspectors found no instances in which NRC-1 icensees 

had imposed nuclear-unique QA requirements on SynTech or had invoked 10 CFR 

Part 21 in their procurement documents to SynTech. 

Although many of SynTech's nuclear plant 

In addition, the inspectors determined through information obtained from some 1 icensees that they typically procure SynTech products as commercial grade 

material, although some handle them under their augmented quality programs in 

order to maintain certain controls, such as requiring analyses for contaminant 

exclusion, when they are used in maintenance of safety-related systems and 

equipment . 

6 

-65- 
Finally, during discussions with SynTech personnel and review of procedures 

and records, the inspectors also determined that SynTech had established a 

comnercial quality program, but did not have a 10 CFR Part 50, Appendix B, 

quality assurance program. SynTech was not familiar with the requirements of 

10 CFR Part 50, Appendix 6, and 10 CFR Part 21. 

c. Concl usi on 

The inspectors concluded that SynTech’s commerci a1 qual i ty program does not, 

nor was it intended to meet Appendix B QA requirements. 

inspectors concluded that SynTech products have not and are not being designed 

or manufactured under a 10 CFR Part 50, Appendix 6, quality assurance program. 

Further, SynTech products have no nuclear-unique design requirements, they are 

used in non-nuclear appl ications and they may be purchased on the basis of 

SynTech’s published product descriptions. Therefore, SynTech products meet 

the definitions of commercial grade items in 10 CFR 21.3 for both 10 CFR Part 

50 1 icensees and non-Part 50 1 icensees, and are not procured or suppl ied as 

basic components. 

Therefore, the 

3.3 guestionable SvnTech Analysis ReDort and CoC 

a. ScoDe 

In preparation for this inspection, the inspector reviewed documentation 

provided by UTS and Consumers Power Company (CPCo) at Pal isades relevant to 

the circumstances surrounding SynTech’s submission to Consumers Power Company 

(CPCo), Palisades’ licensee, of what appeared to be a DE chemical analysis 

report, dated May 15, 1995, that was not issued by DE. Although SynTech did 

not have a copy of this report in its files at the time of the NRC inspection, 

the inspectors had reviewed a copy along with other SynTech-related documentation obtained prior to this inspection. During this inspection, SynTech 

brought this issue to the attention of the inspectors before being asked about 

it specifically. 

SynTech and interviewed SynTech personnel on this issue. 

The inspectors reviewed relevant documentation on file at 

b. Observations and Findinqs 

b.1 Report Anomalies 

The report exhibited several anomal ies, most notably that , a1 though it 

appeared to have been printed on DE letterhead or stationery, unlike the 

genuine DE reports that SynTech had on file, this report contained results of 

analysis of only one lot of only one product, Specifically Lot 761-055 of 

SynTech N-Grade Gasket Remover and Paint Stripper. The report was in the form 

of a letter that was typed on what initially appeared to be DE letterhead, and 

contained a signature block indicating the name of the analyst who had signed 

the previous DE reports, but his signature itself was not present on the 

report in question. 

for which UTS stated it had no record of testing. 

In addition, the report provided data for a sample lot 

7 

-66- 
b.2 The Paljsades Response 

Upon receipt of the product, Palisades conducted its own analyses on a sample 

from Lot 761-055, as is their routine practice, but obtained unacceptable 

results, most notably excessive chlorides. This prompted Pal i sades 

Procurement Engineering to obtained SynTech's CoC and the attached analysis 

report for comparison. Although the report was on what appeared to be DE 

letterhead, the procurement engineer involved stated that she did not notice 

that the report was not in the usual format because normally only receiving 

sees them and they didn't identify that as an anomaly. Nevertheless, due to 

the significantly differing results, Palisades procurement engineering called 

the named analyst at DE/UTS in order to resolve the differences. 

telefaxed to the analyst a copy of the report. 

She also 

The Palisades procurement engineer further stated that she later contacted the 

QA Manager at SynTech who reportedly told her that the report was a mistake by 

a new person. The SynTech QA manager also reportedly requested that Palisades 

send back all the product and the original CoC. 

inspectors that they did this except for the one Lot 761-055 can from which 

the CPCo sample was taken. 

With respect to 10 CFR Part 21 evaluation, the Palisades procurement engineer 

explained to the NRC inspector that UTS had told her that they would make any 

reports required to the NRC; therefore CPCo did not evaluate the deviation 

under Part 21 procedures (a) because of this assurance from DE/UTS, (b) 

because the products were not considered basic components, and (c) because 

CPCo had never used this batch in Palisades. 

stated, would have been verified satisfactory or rejected. 

Palisades informed the 

Previous batches used, she 

b.3 The UTS Reply 

According to an August 24, 1995, handwritten memo to Palisades from the DE/UTS 

analyst named in the report, written in response to Palisades' call and fax, 

DE did not issue that report. The memo also stated that the report was not in 

the standard format of current UTS reports in that it did not have a report 

number or supervisory signatures, and if "nuclear related," was missing a 

traveller number and QA review signatures. However, the inspectors noted that 

the memo did not mention the fact that DE had provided similar reports 

(although covering multiple products and lots) on its letterhead until 1993. 

b.4 UTS Correspondence with SynTech 

In an August 29, 1995, letter to SynTech in response to the apparent 

unauthorized use of DE letterhead, and after a reported telephone conversation 

between a UTS QA representative and a SynTech sales representative, UTS 

demanded that SynTech stop using DE stationery, inform UTS of every instance 

in which Syniech had published product analysis results on DE stationery, and 

inform all its affected customers that DE did not perform the analyses. 

Again, there was no mention of the earlier reports DE had provided SynTech on 

DE letterhead before the establishment of UTS. 

a 

-47- 
b.5 SynTech Reply to UTS 

The inspectors also reviewed a copy of a SynTech letter, also dated August 29, 

1995, from the SynTech QA director, addressed to the UTS QA representative, in 

reply to the UTS letter of the same date, that denied any possession or use of 

DE stationery. 

questions. Its explanation of the circumstances surrounding the issuance of 

the report in question was vague and still did not explain how SynTech would 

have had legitimate possession of DE letterhead, not blank stationery. 

However, this letter also did not clearly answer the UTS 

b.6 UTS Contact with NRC 

In February 1996, in response to the inspector's inquiries to DE and UTS, a 

UTS QA representative informed the inspector that a SynTech sales 

representative had stated to him (in the telephone conversation mentioned 

above prior to issuance of the UTS August 29, 1995, letter) that SynTech had 

"routinely reported results of chemical analyses of its products to licensees 

and others on Detroit Edison letterhead." This statement was construed by UTS 

as an admission by SynTech. In addition, to compound the misunderstanding, 

during this reported conversation, the SynTech sales representative a1 legedly 

could not recall, when asked by UTS, how SynTech came into possession of DE 

stationery . 

b.7 NRC Review at SynTech 

When the NRC inspectors first learned of this issue, the inspectors were not 

aware of the DE letterhead reports on file at SynTech and the UTS QA 

representative did not mention them. 

other documents reviewed by the inspectors at SynTech, the telephone 

conversation statement attributed by UTS to SynTech's sales representative 

(confirmed during the inspection), was consistent with the circumstances. 

However, in light of the DE reports and 

During this inspection, the SynVech sales representative stated that he had 

admitted at the time of the reported telephone conversation that he could not 

offer a satisfactory explanation for the issuance of the analysis report in 

question (a) on what appeared to be DE letterhead during a period when UTS had 

taken over analysis of SynTech products, and (b) reported results on a lot of 

N-grade that neither UTS, nor apparently DE, had a record of analyzing. He 

stated that his remark about routinely issuing reports on DE letterhead was 

made in an effort to explain to the UTS QA representative how SynTech came to 

have possession of DE letterhead. However he failed to make it clear that the 

letterhead to which he was referring was what the DE reports had been printed 

on, not blank DE stationery. 

never had blank DE stationery. Nevertheless, the unclear explanation provided 

by the SynTech Sales representative, interpreted by UTS as an admission by 

SynTech, served to confirm UTS's suspicion that SynTech was routinely sending 

out fraudulent reports on DE letterhead (presumably using blank DE 

stationery), thus prompting UTS's August 29, 1995, letter. The inspectors 

found no evidence to suggest that SynTech had done more than send a single 

report on what appeared to be DE letterhead, and apparently only to Palisades 

on the one occasion in question. 

SynTech stated during this inspection that it 

9 

-68- 
As mentioned above, the inspectors found three reports on DE letterhead that 

had been sent to SynTech by DE, during the period before the formation of UTS, 

when DE'S Technical and Engineering Services group had been the name of the 

organization whose 1 aboratory was performing contaminant analyses of SynTech 

products. The reports, dated in 1990 and 1992, were all signed by the analyst 

whose name, but not his signature, appeared on the report in question. The 

inspectors noted that none of the analysis reports from DE on file at SynTech. 

contained results of analysis of N-grade, and the UTS report on file at 

SynTech that did contain N-grade analysis results was for lot 761-193, not lo3 

761-055 as in the SynTech report to Palisades. 

Review of associated records during this inspection confirmed that SynTech had 

in fact routinely reported its product test results to its customers simply by 

forwarding copies of the DE reports which included results for the products 

and lots being shipped, among others. 

given SynTech permission to use its stationery, letterhead, name or logo, but 

the inspectors found no evidence that either DE or UTS had ever prohibited 

SynTech from forwarding DE or UTS analysis results to SynTech customers on the 

original reports. UTS also confirmed the inspectors' findings at SynTech 

that, although the UTS laboratory had performed a number of analyses for 

SynTech on its products over the last two years, the lot listed on the 

apparently fraudulent report was never analyzed by UTS. 

UTS had also stated that it had never 

b.8 SynTech Internal Investigation and Explanation 

Presumably in anticipation of questions about this issue by the inspectors, 

SynTech disclosed the issue of the questionable report to the inspectors and 

offered their explanation of the circumstances. 

response to the August 29, 1995, letter from UTS, it had conducted an internal 

investigation to establish the facts and circumstances surrounding the 

issuance of the report in question. 

investigation, SynTech developed the following explanation: At the time the 

report in question was sent to Palisades, May 1995, SynTech was using UTS to 

perform its product contaminant analysis for nuclear customers. However, on 

this particular occasion, in response to an order from Palisades for some of 

the N-Grade Gasket Remover and Paint Stripper (hereinafter referred to as "NGrade"), SynTech discovered that it did not have a UTS analysis report on file 

for the lot of N-grade it had in stock, Lot 761-055. SynTech stated that upon 

informing Palisades of this fact, Palisades agreed to accept so-called 

"typical" results for this order, because they would be testing a sample of 

the received product themselves anyway. Therefore, according to SynTech, 

instructions were given to a clerk (who was reportedly new at the time and who 

is no longer with the company) to prepare a report for Palisades of typical 

contaminant analysis results for N-Grade Lot 761-055. 

produced examples from their files in support of the explanation) that after 

sending out DE reports with multiple product and lot results on them, it had 

become their practice for a while to scan into their computer the letter 

reports from DE. 

customized reports containing only the data for the product and lot being 

shipped under the report and the accompanying CoC. 

this practice would result in fewer, if any, transcription errors, less 

SynTech stated that in 

On the basis of the results of this 

SynTech explained (and 

The reports could then be conveniently edited to produce 

The rationale was that 

10 

-69- 
confusion, and admittedly, would not disclose unnecessarily the few results 

not pertinent to the shipment being certified that did not meet contaminant 

speci f i cat i ons . 

During this inspection, the inspectors examined some of these edited reports 

and found that they were printed out on plain paper and in a font not 

resembling the font of the original DE letters, but some were left partially 

in the form of letters from one of SynTech's contractor laboratories addressed 

either to the SynTech QA Manager or to the SynTech sales representative who 

was involved, depending on the vintage. The inspectors pointed out to SynTech 

that while it would have been appropriate to quote the results of the product 

and lot of interest from the test report and/or to refer to the report or even 

to attach a copy, it was not appropriate to edit the original report to 

isolate the desired set of results, yet leave it partially in the form of a 

letter report to SynTech which would result in a document that was 

fundamentally misleading as to its origin, if not the content. SynTech 

agreed, but stated that they had discontinued this practice entirely when 

their investigation to determine how the report in question could have been 

produced revealed that the scanning and editing practice was a significant 

contributing factor. 

SynTech stated that as nearly as they were able to determine, the clerk who 

had been told to produce a report for the Palisades N-grade order in question 

citing typical results for this product had called up an old DE report that 

had been scanned into the computer to obtain the results of analysis of a 

previous lot of the same product. However, instead of using the report (not 

scanned) provide by UTS in 1992 of Lot 761-193 of N-Grade that SynTech had on 

file, the clerk reportedly deleted the data for products not being shipped, 

leaving only the N-grade data and then edited the date of the report and the 

lot number to read May 15, 1995 and Lot 761-055 that was being shipped. 

SynTech stated that they were not able to find the original report from which 

the "typicals" presumably were taken so the origin or correctness of the data 

could not be verified. Then, SynTech explained, and as is apparent when 

examining the report, for an unknown reason, the clerk did not edit out the DE 

logo that appeared in the upper left hand corner of the page. The logo 

consisted of printed words in a stylized version of a bold block letter font 

and it had been scanned as part of the document by the optical character 

recognition (OCR) software being used by SynTech as part of the text of the 

letter instead of as a graphic, in a font as close to the one scanned as the 

software could reproduce. Upon cl ose examination, the inspectors were ab1 e to 

see that in fact, what had appeared to UTS and CPCo to be DE letterhead, was a 

computer reproduced facsimile of the logo that had reportedly been on the 

stationery on which the original report (similar to ones seen by the 

inspectors in SynTech's files) had been written. Close examination of the 

letters revealed that the reproduced font differed slightly from the actual DE 

logo in the thickness, spacing and proportions of the letters, the alignment 

of the words, and in particular, in the form of the letter "t" as it appeared 

in the word "Detroit" in the name and address of DE in the logo. 

actual DE logo, the stylized "t" is not fully crossed. In the custom DE font, 

the cross of the "t" extends from the vertical part only to the right, not to 

the left as on a conventional 'It." This feature, once noticed was evident on 

all the actual DE reports on file. Whereas, the scanned document, exhibited a 

In the 

11 

-70- 
similar font, but, as previously stated with slightly differently formed and 

aligned letters and with normal t’s on which the cross piece extended fully 

from the left to the right sides. 

SynTech’s explanation of how the report in question had been produced. Also, 

SynTech assured the inspectors that although it no longer had the scanning 

software used so that could not demonstrate the effect, it was named and 

described as readily available so that the NRC could test SynTech’s 

expl anat i on. 

This information tended to support 

Nevertheless, SynTech was not able to offer a satisfactory explanation of why 

the scanned and incorrectly edited report was sent out in that form and not at 

least edited to the extent that others had been. 

clerk would not explain why supervisors had not caught the error. SynTech 

could only postulate that in their haste to fill the order and deliver the 

paperwork quickly in response to customer pressure, they had failed to 

adequately review a1 1 the submitted documentation, including the accompanying 

CoC letter which also referenced the lot number instead of stating that the 

accompanying results were supposed to be typical. 

Even the error of a new 

c. Concl us i on 

Prior to the inspection, the information available to the inspectors, as 

discussed above, suggested a fraudulent SynTech report. However, SynTech’s 

explanation of the circumstances surrounding the creation of the report in its 

unusual form, was plausible, if not conclusive, and was supported by other 

information obtained by the inspectors. Although SynTech could not offer a 

satisafactory explanation for the report’s not being effectively screened 

through management review prior to issue, the incident appeared to have been 

isolated. 

deliberate intent to mislead the recipient of the report and found no evidence 

of a pattern of similar activity. 

The inspectors could not conclusively determine that there was any 

3.4 Procurement Status and Chemical Analysis Results 

a. ScoDe 

To dete 

content 

SynTec h 

761-055 

rmine the validity of SynTech reported contaminant content of Lot No. 

(as well as other products), and the consistency of contaminant 

in SynTech products, the inspectors interviewed analysts at two of 

’s former contract laboratories, both NRC licensee subsidiar es, had 

Branch, 

the 

neer i ng . 

the results evaluated by the NRC Chemical Engineering and Materials 

reviewed numerous analysis reports on SynTech products and reviewed 

nuclear-grade evaluation of SynTech products by GE NE Materials Eng 

b. gbservations and Findinas 

During a telephone conversation on July 15, 1996, between the inspector and a 

representative of Consumers Power (CPCo) , Pal isades Procurement Engineering, 

the Palisades procurement engineer stated that Syntech Products are treated 

under CPCo’s consumables program as Augmented Quality items because SynTech is 

not a qualified/approved supplier. CPCo had, for a number of years, purchased 

12 

-71- 
the product in question ("N-Grade Gasket Remover and Paint Stripper - 

Certified Non-Chlorinated, C1 inging Gel") among aiher SynTech products as 

non-safety-related. However, because the products are used on safety-related plant components, CPCo said it always independently tested a sample of each 

lot/batch in its own Trail Street Labs (owned by CPCo) to verify the product 

meets CPCo's safety-re1 ated or Category A requirements. For consumabl es used 

for cleaning, etc., even though they are not allowed to remain in safetyrelated systems, CPCo conservatively prefers not to allow any undesirable 

substances (e.g., fluorides, chlorides, sulfides, heavy metals, etc.) to come 

in contact with safety-related metals. 

lot is routinely sent to the CPCo lab by the Palisades receiving department 

for analysis. In September of 1995, CPCo's Trail St. Labs obtained results 

reported as "total chloride"[sic] of 10,492 ppm on the sample of lot 761-055. 

This was significantly higher than the 120 ppm "C1" [sic] results reported on 

the May 15, 1995 SynTech report in question. 

procedures, the issue was then referred to Procurement Engineering for 

resol uti on. 

Therefore, a sample of each received 

In accordance with CPCo 

On August 6, 1996, the inspector talked to a chemist at CPCo's Trail Street 

Labs, about the analysis report provided by SynTech to Palisades of Lot No. 

761-055 of its "N-Grade" solvent/ cleaner and gasket remover. 

the differences between her lab's analysis of the product, and the SynTech reported results, which differed widely as discussed previously. 

confirmed the inspector's concern that the SynTech results of heavy metal 

analysis may be questionable because many of the heavy metals (e.g., As, Bi, 

Pb, Sn, At, Hg, etc.) were reported by SynTech on the order of tl to t5 ppm, 

whereas, CPCo's analyses of these same elements were reported typically as (50 

or tlOO ppm because of the minimum sensitivity of the analyses. 

the inspector was concerned that unless SynTech (or its contract laboratories) 

had more sensitive instruments and processes for analyzing for these elements, 

the SynTech-reported results may have been below what is typically achievable, 

and would therefore, be questionable. 

Discussed were 

She 

At the time, 

During this inspection, however, the inspectors found comparable results on 

another lot of the same product in a test report by UTS. 

reviewed by the NRC's Chemical Engineering and Materials Branch who stated 

that the SynTech-reported results were not below the sensitivity of the more 

sophisticated techniques in common use in the industry. 

During subsequent conversations with the analyst at UTS who had signed the old 

DE reports to SynTech and also the analyst who had performed the analyses for 

UTS, they indicated that results of analyses could vary widely depending on 

analytical method used if only minimums are being reported due to minimum 

sensitivity or detectable qualities. Also, for halides, sulfur, it is 

necessary to know whether the analysis was for total amounts or leachable 

amounts. If being analyzed for "total" content, chlorinated or fluorinated 

compounds will release large quantities of halides when certain analytical 

methods are empl oyed. 

Also, the report was 

The second conversation the inspector had with the CPCo chemist at the Trail 

Street Labs confirmed that her analysis was for total chlorides, yielding 

10,492 ppm, indicating that the product contained chlorinated compound(s). On 

13 

-72- 
August 29, 1996, the Palisades procurement engineer faxed the NRC copies of 

previous reports of analyses of SynTech products done by the CPCo lab for 

themselves and said they would locate the records of analyses they used to do 

for SynTech several years ago, just as UTS had done until recently. The 

inspectors found some of these old (pre DE) CPCo reports on file at SynTech. 

c . Concl us i on 

On the basis of the information from the UTS and CPCo laboratories, the NRC 

Chemical Engineering and Materials Branch review, the GE NE nuclear-grade 

evaluation, and the analysis reports reviewed, the inspectors concluded that 

although N-Grade, Lot 761-055, contained excessive chloride, even for total 

(vice leachable) content, the results of the other substances analyzed for 

were consistent with the other analyses of N-Grade. The inspectors further 

concluded that the SynTech reported results, a1 though their specific origin 

was indeterminate, appeared to be typical of N-Grade and were consistent with 

the capabilities of the analytical methods used. 

However, on the basis of review of many SynTech product analysis reports and 

review of GE NE Materials Engineering recommendations in their nuclear-grade 

evaluation, the inspectors concluded that the contaminant content of SynTech 

products is variable enough such that "typical" analyses are not useful. 

Therefore, the practice of most of SynTech's nuclear customers to require 

batch-traceable or lot-traceable certified chemical contaminant analysis 

reports is prudent. Such reports would ensure that levels of undesirable 

substances such as halides, sulfur and heavy metals are within acceptable 

limits. 

4. PERSONS CONTACTED 

*+J. Leslie, President 

+M. Bramson, Executive Vice President 

*+D. Moore, Qual i ty Assurance Manager 

J. Broderick, Sales 

T. Mosier, Assistant QC Coordinator 

+ Attended the entrance meeting on February 4, 1997. 

* Attended the exit meeting on February 6, 1997. 

14 

-73- 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 2OSEAW01 

May 2, 1997 

Mr. Nicholas J. Liparulo, Manager 

Nuclear Safety and Regulatory Activities 

Nuclear and Advancr -' 'xhnolcgy Division 

Westinghouse Electric Corporation 

0.0. Box 355 

Pittsburgh, Pennsyivsnia 15230 

SUBJECT: NRC INSPECTION NO. 3S3 f 1 ~97-0i 

Dear Mr. Liparulo: 

On April 17, 1997, the U. S. Nuclear Regulatory Commission (NRC) completed an 

inspection of AP600 design control quality assurance activities at the Westinghouse 

Energy Canter in Monroeville, Pennsylvani:.. The enclosed report presents the results of 

that inspection. 

The purpose of the irlsoection was to evaluate the Westinghouse technical and quality 

oversight of Empresa Nacional de Ingeqieria y Tecnologia, S.A. (INITEC) design activities in 

general, and to assess Westinghouse's evaluation and disposition of the audit findings 

identified in the August 24, 1994, NRC "Summary of Audit of the AP600 Structural 

Design," report concerning NRC-identified errors in INITEC calculations. 

During this inspection, the NRC determined that the implementation of the Westinghouse 

quality assurance program for AP600 design certification activities failed to meet certain 

NRC requirements. Specifically, the team identified nonconformances with program 

implementation with respect to your: (1 failure to initiate appropriate root cause 

determinat'on and corrective actions for find'ngs described in the August 24, 1994, NRC 

Audit Summary report, and (2) failure to adequately evaluate and assess IFJITEC's 

performancs during your annual quality assurance review of AP600 suppliers, and to 

conduct an evaluation oi INITEC's February 7 5, 1995, response to the NRC audit findings 

as requested in Westinghouse's August 3, 1994, letter, during your triennial audit. During 

the required triennial supplier audit of INITEC performed in February 1995, Westinghouse's 

audit plan and scope did not include the calculation audit findings identified in the August 

1994 NRC Audit Summary report, and any associated INITEC corrective action activities, 

as issues that needed to be addressed to verify the effectiveness of INITEC's quality 

assurance program implementation. 

Based on the nonconformances identified above, the NRC is concerned that these quality 

assurance deficiencies may have introduced a level of uncertainty on the acceptability of 

design deliverables provided by AP600 technical cooperation agreement participants. Of 

particular concern to the NRC, is Westinghouse's failure to recognize and appropriately 

address a condition adverse to qcalirv, requiring a root cause evaluation and determination 

and appropriate corrective actions, even when such a condition was identified by an NRC 

audit and resulted in re-design of the AP600 foundation basemat. 

-74- 
Mr. Nicholas J. Liparulo -2- 

Accordingly, the NRC requests that Westinghouse: (1 ) determine and evaluate the impact 

of these nonconformances on completed or related design deliverables and/or activities 

performed by all AP600 technical cooperation agreement participants; (2) identify the 

steps that it has taken, or intends to take, to demonstrate that other design deliverables 

provided by AP600 technical cooperation agreement participants do in fact achieve the 

level of integrity in design verification and quality assurance necessary to satisfy the 

design certification provisions of 10 CFR Part 52; and (3) provide a list of all AP600 

technical cooperation agreement participants, including a description of their AP600 

program work scope and involvement. The NRC is identifying these concerns as an 

unresolved item. Please provide the above requested information within 30 days of receipt 

of this letter. 

The responses requested by this letter and the enclosed Notice of Nonconforrr . .ice 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. 96-51 1. 

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. 

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

them with you. 

Sincerely, 

'Robert M. Galio, Chief 

Special Inspect on Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Docket NO.: 52-003 

Enclosures: 

1. Notice of Nonconformance 

2. Inspection Report No. 99900404/97-01 

cc wlencls: See Next Page 

-75- 
Mr. Nicholas J. Lipard, 

Wastinghouse Electric Corporation 

cc: 

Mr. B.A. Mclntyre 

Advanced Plant Safety & Licensing 

Westinghouse Electric Corporation 

Energy Systems Business Unit 

P.O. Box 355 

Pittsburgh, PA 15230 

Ms. Cindy L. Haag 

Advanced Plant Safety & Licensing 

Westinghouse Electric Corporation 

Energy Systems Business Unit 

Box 355 

Pittsburgh, PA 15230 

Mr. M.D. Beaumont 

Nuclear and Advanced Technology Division 

Westinghouse Electric Corporation 

One Montrose Metro 

1 1921 Rockvilla Pike 

Suite 350 

Rockville, MD 20852 

Mr. Sterling Franks 

U.S. Department of Energy 

19901 Germantown Road 

Germantown, MD 20874 

NE-50 

Mr. S.M. Modro 

Nuclear Systems Analysis Technologies 

Lockheed Idaho Technologies Company 

Post Office Box 1625 

Idaho Falls, ID 83415 

Mr. Frank A. Ross 

U.S. Department of Energy, NE-42 

Office of LWR Safety and Technology 

19901 Germantown Road 

Germantown, MD 20874 

Docket No.: 52-003 

AP600 

Mr. Ronald Simard, Director 

Advanced Reactor Programs 

Nuclear Energy Institute 

1776 Eye Street, N.W. 

Suite 300 

Washington, DC 20006-3706 

Ms. Lynn Connor 

Doc-Search associates 

Post Office box 34 

Cabin John MD, 20818 

Mr. James E. Quinn, Projects Manager 

LMR and SBWR Programs 

GE Nuclear Energy 

175 Curtner Avenue, M/C 165 

San Jose, CA 95125 

Mr. Robert H. Bucholz 

rJE Nuclear Energy 

175 Curtner Avenue, M/C 781 

San Jose, CA 35125 

Barton Z. Cowan, Esq. 

Eckert Seamans Cherin & Mellott 

600 Grant Street 42nd Floor 

Pittsburgh, PA 1521 9 

Mr. Ed Rodwell, Manager 

PWR Design Certification 

Electric Power Research Institute 

341 2 Hillview Avenue 

Palo Alto, CA 94303 

Mr. Charles Thompson, Nuclear Engineer 

AP600 Certification 

U.S. Department of Energy 

Washington, DC 20585 

NE-50 

-76- 
NOTICE OF NONCONFORMANCE 

Westing house Electric Corpora tion 

Pittsburgh, Pennsylvania 

Docket Nos.: 52-003 

99900404 

Based on the results of a Nuclear Regulatory Commission (NRC) inspection conducted on 

April 17, 1 997, of activities supporting Westinghouse Electric Corporation's AP6Or) design 

certification, it appears that certain activities were not conducted in accordance with NRC 

requirements. 

A. Criterion XVI, "Corrective Action," of Appendix B to 10 CFR Part 50, requires, in 

part, that measures be established to assure that conditions adverse to quality, such 

as failures, malfunLtion-, deficiencies, deviations, defective material arld kquipment, 

and nonconformances are promptly identified and corrected. 

WCAP-12600, "AP600 Quality Assurance Program Plan," Revision 2, dated 

December 15, 1993, Section 16, "Corrective Action," states, in part, "Application of 

WCAP-8370 to AP600 activities includes the following: 

0 The root cause of significant conditions adverse to quality are determined 

and documented, and the impact of such conditions on completed or related 

items and activities is evaluated." 

WCAP-8370, Quality Assurance Plan (QA Topical Report), Revision 12A, dated April 

1992, Section 16, "Corrective Astion," stales, in part, in: 

Section 16.0, "General," that conditions adverse to qualitv such as failures, 

mal;unctio x, nonconformancss, and out-of control processes (including failure to 

follow procedures) shall be identified. These adverse conditions are also analyzed, 

documented, and corrected commensurate with their importance to safety. 

Section 1 6.1, "Corrective Ac In," that personnel performing activities in accordance 

with this plan identify conditions adverse to quality and suggest, recommend or 

provide solutions to the conditions as appropriate. For significant conditions adverse 

to quality, the causes are determined and documented and the impact of such 

conditions on 'terns and services is evaluated for significant trends and reported to 

the appropriate level of management. 

Section 1 5.2, "Follow-up," that for corrective action resulting from reports (e.g., 

nonconformance reports, audit reports, computer software error reports, NRC 

inspection reports, customer audit reports, etc.) quality assurance participates in 

verifying that appropriate corrective action is documented and implemented. 

Enclosure 1 

-77- 
Contrary to the above, Westinghouse: 

1. Did not identify, analyze, document, 3,id correct conditions adverse to 

quality as required by the AP6CO Quality Assurance program. During a July 

1994 NRC structural audit of the nuclear island foundation mat, errors were 

identified in calculations performed by INITEC that resulted in significant 

re-design of the AP600 foundation basemat. The findings described in the 

August L-, 1994, NRC "Summary of Audit of the AP600 Structural 

Design," report were not identified as a condition adverse to quality requiring 

or rec5ving quality assurance participation in verifying that appropriate 

corrective action is documented ar,d implemented. 

2. Did not adequately determine and document the root cause of INITEC'S 

basemat zalculation errors nor evaluate the impact of such a condition 

adverse to quality on completed or related INITEC AP600 design dsliverables 

and activities. (99900404/97-3: -01 1 

6. Criterion VII, "Control of Purchased Material, Equipment, and Services," Of 

Appendix B to 10 CFR Part 50, requirds, in part, that measures shall be established 

to assure that purchased material, equipment, and services, whether purchased 

directly or through contractors or subcontractors conform to the procurement 

documents. 

WCAP-12600, "AP600 Quality Assurance Program Plan," Revision 2, dated 

December 15, 1993, Section 7, "Control ol Purchased Items and Services," states, 

in part, "Application of WCAP-8370 to AP600 activities includes, but is not limited 

to, the following: 

0 The initial qualification and subsequent performance evaluation of suppliers 

to which technicaf cooperation agreements [emphasis addedl apply is 

performed in the same manner as for suppliers of purchased items and 

services. 

0 The perfsrmance of each supplier is evaluated on an annual basis, 

commensurate with the complexity and importance to safety of items or 

services provided. The evaluation is documented and includes evidence, 

based on direct observation of work performed by the supplier, that the 

supplier's quality assurance program is continuing to operate successfully." 

WCAP-8370, Quality Assurance Plan (QA Topical Report), Revision 12A, dated April 

1992, Section 7, "Control of Purchased Items and Services," states, in part, in: 

Section 7.3, "Supplier Performance Eval.uation," that "A formal evaluation of 

suppliers is performed each year to determine if additional actions such as audits are 

required during the upcoming year. This evaluation includes a review of some or all 

of the following: prior quality program audits, supplier surveillance activities, 

, . .resuJts of audits from other sources (customers, ASME, NRC, etc.1 if available 

[emphasis added], ... and the suppliers's respcnsiveness and cooperat& in resolving 
quality questions or problems. As a result Jf thi5 evaluation, suppliers requiring a 

complete quality program reauait are identified.. .Regardless of the results of the 

evaluation, suppliers are reaudited every three years." 

Contrary to the above, Westinghouse did not provide appropriate QA oversight of 

design activities performed by INITEC. Afte the basemat calculation errors were 

identified by the NRC in July 1994, Westinghouse did not evaluate or assess the 

impact of the errors on other work performed by INITEC. Specifically: 

1. 

2. 

rvestinghouse failed to adequately evaluate or assess INITEC's annual 

performance, as required by WCAP-8370, Part 8, Section 7.3, "Supplier 

Performance Evaluation," for a supplier of AP600 design deliverables that 

had been the subject of an adverse NRC audit finding. 

In its February 1995 triennial audit of INITEC, Westinghouse failed to 

conduct an evalLlation of INITEC's response to Westinghouse's Aug.-.;t 3, 

1994, letter, and any associated corrective actions taken. The letter to 

INITEC described the basemat design calculation issues identified by the 

NRC during the July 1994 struc~ur~l Jesign audit. (99900404/97-01-02) 

Please provide a written statement or explanatim 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 I, ispection 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 shoull include for each nor,conformance: (1 1 a description of tl.- 

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

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

preventative measures were or will be completed. 

Dated at Rockville, Mar land 

This m, day of day , 1997 

-3- 

-79- 
Report no.: 

Organization: 

Contact: 

Nuclear industry 

Activity: 

Date: 

Inspectors: 

Approved: 

U. S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

99900404/97-0 1 

Westinghouse Electric Corporation 

Nuclear and Advanced Technology Division 

Pittsburgh, Pennsylvania 15230 

;\.r. Nicholas J. Liparulo, Manager 

Nuclear Safety and Regulatory Activities 

Nuclear steam supply system design, components and 

services 

April 17, 1997 

Richard P. Mclntyre, ?SIB/DISP 

Juan D. Peralta, HQMB/DRCH 

Goutam Bagchi, ECGB!DE 

Diane Jackson, PDST/DRPM 

Gregory C. Cwalina, Chief 

Vendor Inspection Section 

Special Inspection Branch 

Division of inspection and Support Programs 

Enclosure 2 

-80- 
1 INSPECTION SUMMARY 

The purpose cf the inspection was to evaluate the Westinghouse Lechnical and quality 

oversight of Empresa Nacional de lngenieria y Tecnologia, S.A. (INITEC) design activities in 

general, and to assess Westinghouse's thoroughness in its evaluation and disposition of 

the audit findings identified in the August 24, 1994, NRC "Summary of Audit of the 

AP600 Structural Design" concer;?ing NRC staff-iuentified errors in INtTEC calculations: as 

described in Section L.&. 

The inspection bases were: 

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

Reprocessing Plants," to Part 50 uI I ilie 10 of the Code of Federal Reaulations 

(10 CFR Part 50). 

APSO0 SSAR, Revision 1 1, Section 17.3, "Quality Assurance" 

0 WCAP-12600, Revision 2, dated December 1993, "AP600 Quality Assurance 

Program Plan" 

During this inspection, two instances where Westinghouse failed to conform to NRC 

requirements were identified. Also, one unresolved item was identified. 

1.1 Nonconformance 

Nonconformance 99900404/97-01-01 was identified and is discussed in Section 3.3 

of this report. 

0 Nonconformance 99900404/97-01-02 wes identified and is discussed in Section 3.4 

of this report. 

1.2 Unres.,- 

Unresolved Item 9?900404/97-01-03 was identified and is discussed in Section 3.5 

of rnis repcrt. 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

No previous inspection findings were reviewed during this inspxtion. 

-2- 

-81- 
3 INSPECTION FINDINGS AND 01 HER COMMENTS 

3.1 AP600 Qua litv Assuiance Proaram 

Chapta 17 of the AP600 standard safety E ?alysis report (SSAR) describes the 

Westinghouse Electric Corporation quality assurance (CIA) program for the design 

phase of the APE,, AdvancGd Light Water Reactor (ALWR) Plant Program. 

AP600 SSAR, Revision 1 1 , Section 17.3, "Quality Assurance," states that activities 

performed prior to March 31, 1996, were performed in accordance with the quality 

assurance pian described in Westinghome topical report WCAP-8370, "Energy 

System Business Unit - Power Gerk, oram Business Unit, Quality Assurance Plan," 

Revision 1 2a, dated April 1992. WCAP-8370 applied to all Westinghouse activities 

affecting ,dality of items and services supplied to nuclear power plants <,nd 

establishes Westinghouse's compliance with the provisions of Appendix B to 

10 CFa 50. 

WCA?-12600, "AP600 Quality Assurance Program Plan," dated December 1 SS3, a 

project-specific QA plan, was developed by Westinghouse to enhance WCAP-8370 

in specific areas and to establish additional cornmitlnents needed to support the 

AP600 Design Cerrification and First-Of-A-Kind (FOAKE) programs. WCAP-12600 

establishes the responsibility of the Advance Technology Business Area of the 

Energy Systems Business Unit fo. AP600 Design Certification and FOAKE programs 

arld for control of the technical interface between Westinghouse and engineering 

groups and suppliers providing engineering services under such programs. 

Additionally, WCAP-1260 1, "AP600 Program Operating Prxedures," Revision 13, 

dated July 8, 1994, was develcped by Westinghouse to establish requirements and 

responsibilities for developing, approving, implementing, revising, and maintaining 

operating procedures to meet the QA and administrative requirements of the AP600 

proyra n. WCAP-12601 includes an "Ai 600 Program Procedure Matrix," currently 

Revision 15, dated April 4, 1995, which identifies the correlation between the 

Westinghouse commitments to the OA requirements of (1) ANWASME NQA-1, 

" Qualicy Assurance Program Requirements for Nuclear Facilities," 1 983 Edition (as 

endorsed by Regulatory Guide 1.28, Revision 3) and (2) ANWASME NQA-1 1989 

Edtion through NQA-1 b-199 1 Addenda, and the corresponding implementing 

guidance embodied in WCAP-9565, "Nuclear and Advanced Technology Division 

(NATD) Quality Assurance Program" currently Revision 34, dated May 2, 1994, and 

in WCAP-12601, WCAP-9565 governed the implementation of all NATD activities 

related to areas within the scope of WCAP-8370. 

3.1.1 AP600 Desian Certification and FOAKE Proarams Control of Purchased Items and 

Services 

WCAP-12600, Section 7, "Control of Purchased Items and Services," describes the 

application of WCAP-8370, Part B, Section 7, provisions to AP600 activities and 

clarifies that the initial qualification and subsequent performance evaluation of 

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suppliers to which technical cooperation agreements apply [emphasis added] is 

performed in the same manner as for suppliers of pi mhased items and services. 

WCAP-8370, Part 6, Section 7.3, "Supplier Performance Evaluation," requires, in 

part, "A formal evaluation of suppliers is performed each year to determine if 

adddtional actions such as audits are e .uired dJring the upcoming year. This 

evaluation includes a review of some or ail of the following: prior quality program 

audits, supplier surveillance activities, nature and frequency of hardware 

discrepancies, results of audits from other sources (customers, ASME, NRC, etC.l 

(emphasis added] .....I' This section also specifies that as a result of this evaluation, 

suppliers requirinq a complete quality program reaudit are identified. 

3.1.2 AP600 ParticiDation Proaram Between INITEC and Westinahouse Electric 

Coroora tion 

On March 5, 1992, Westinghouse and INITEC signed an agreement uwler which 

INITEC would perform engineering, management and execution of certain structural 

and piping tasks associated with the AP600 Program. The following tasks were 

identified as INITEC's work srope: (1 1 Analr,,; and Detail Design of Pressurizer 

Safety and Relief Valve (PSARW Module, (2) Structural Steel Framing (SSAR Section 

3.8.41, (3) Floor Slabs (SSAR Section 3.8.41, (4) Nuclear Island Basemat (SSAR 

Section 3.8.51, (5) Nuciear Island Shear Walls (SSAR Section 3.8.41, (6) Analysis 

and Detail Design of Piping, and (7) Structural Steel Framing (SSAR Section 3.8.4). 

Additionally, the agreement specified that design inputs, methods, required formats 

and other design information for structural tasks would be provided by Bechtel 

Power Corporation (Bechtel) and allowed for technical correspondence and interfaces 

for these tasks to take place directly between INITEC and Bechtel. 

In Section 3.0, "Quality Assurance," the agreement specified that INITEC shall 

establish, implement, and maintain a quality assurance grogram meeting the 

requirements of ANSl/ASME NQA- 1 , "Qualit\ Assurance Program Requirements for 

Nuclear Facilities" (1 986 edition). The agreement also specified that a Quality 

Assurance Program Plan (QAPP) would be prepared by INITEC and be subject to 

review and approval by Westir,,nouse prior to performing any quality-related work. 

3.2 Nuclear Island Basemat Desian - Bacl:armnd 

On July 1 1 through 14, 1994, the NRC performed an audit of the structural design 

of the AP600 at the Bechtel offices in San Francisco, California. The results of this 

audit were documented in a letter to Westinghouse dated August 24, 1994, 

"Summary of Audit of the AB600 Structural Design." 

One of the key issues identified during the audit were errors found by the audit team 

in design calculations performed by INITEC (Calculation No. 101 0-CCC-001, Rev. A). 

The NRC audit team identified the following deficiencies: (1) errors in shear and 

flexural rebar assessment, in the use of punching shear formula, and in the use of 

finite element dimension; (2) no consideration of accident pressure loads, and loads 

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3.3 

a. 

b. 

The inspectors reviewed an INITEC letter to Westinghouse (INI/FOKO175), dated 

February 15, 1995, in which INITEC provided its response to address the root 

causes of the quality issue relative to the Nuclear Island Basemat Calculation, 

identified by NRC in its July 1994 structural design audit, as well as the measures 

taken by INITEC in order to avoid the occurrence of similar situations during the 

performance of present and future structural analysis. The inspectors noted that 

INITEC's response had been formulated almost contemporaneously with the triennial 

audit being conducted by Westinghousp nn February 20 through 22, 1995, at 

INITEC's facilities. Yet Westinghouse's triennial audit report (QLA/IN10007) does not 

from con,.truct;t,l, "equence, and (3) out of phase overturning moment from shield 

and containment buildings. 

In its August 2, 1994, letter to the MC, Westinghouse acknowledged its 

commitment to: (1) perform an independent review of the basemat design 

calculations, (2) verify the Zdequacv r f l"ltTFC's in-house post-process computer 

programs used for the foundation mar aesign, (3) perform simplified analyses as 

appropriate to confirm the existing design results, and (4) provide the results cf this 

independent review to the NRC. (This issue was identified by the NRC in the AP600 

DSER as DSER Open Item 3.8.5-21). 

Review of Westinahouse Corrective Actions for Julv 1994 NRC Structural Audit 

Findinas 

Scose 

The inspectors reviewed Westinghouse's evaluation and disposition of the audit 

findings identified in the August 24, 1994, NRC "Summzry of Audit of the AP600 

Structural Design" report concerning NRC-identified errors in INITEC calculations. 

Observations and findinas 

As indicated above, in its August 2, 1994, letter to the staff regarding errors 

identified in INITEC post-process computer programs used for the AP600 foundation 

mat design, Westinghouse made a commitment to conduct an independent review of 

such calculations. In a letter to INITEC (FOK/lNlO181) dated August 3, 1994, 

Westinghouse forwarded: (1 ) comments presented by an NRC consultant at the Jillly 

1994 meeting, (3! additional interpretation and comments by Westinghouse technical 

staff, and (3) a copy of the August 2, 1994, letter to NRC. Westinghouse also 

stated that a Bechtel employee had been selected to perfom the independent review 

and the sc8ipe of basemat review was attached to the letter (Attachment 4). 

The inspectors reviewed the independent review report, "Independent Review of 

AP600 Nuclear Island Basemat Design," dated November 11, 1994, performed by 

Bechtel to address the basemat calculation issues identified by the staff. The 

inspectors noted that I I Section 6.0, "Summary and Recommendations," of the 

report, Bechtel made several observations and recommendations related to 

uncertainties in the INITEC calculation that required further evaluation or study. 

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C. 

3.4 

a. 

b. 

provide any evidence that IKITEC had idenlified this issue as a conditiori adverse to 

quality requiring root cause determination or corrective actions in accordance with 

INITEC’s Westinghouse-approved QAPP (sea Section 3.4.a.). 

During the inspection, the inspectors inquired as to whether Westinghouse had 

initiated my root cause determinations or vvective actions, as required by 

WCAP-12600, Section 16, “Corrective Aciion,” since this issue was first identilied 

by the NRC in July/August 1994 or whether Westinghouse had formally acczpted or 

rejected INITEC’s proposed corrective actions identified in INITES’s February 1995 

letter. We@+‘nghouse representatives responded that, as of April 17, 1997, this 

issue had not been identified as a condition adverse to quality requiring a root cause 

determination or corrective actions in accordance with WCAP-12600 nor had 

Westinshouse formally responded to INITEC’s February 1995 letter. 

Conclusi onS 

Based on the above, the inspectors concluded that Westinghouse: (1 1 failed to 

identify or recognize the NRC-identified basemat design deficiencies as a condition 

adverse to quality requiring a root cause d2;ermination or corrective actions in 

accordance with WCAP-12600, Section 16, and (2) failed to determine and evaluate 

the impact of such design deficiencies on completed or related INITEC AP600 design 

deliverables and activities. This issue was identified as Nonconformance 

99900404/97-0 1 -0 1. 

Review of Westinahouse Oversiaht of INITEC AP600 Desiai Activities 

ScoDe 

The inspectors reviewed Westinghouse technical and quality oversight of INITEC 

desjyn activities in general, including the documents describing the contractual 

agreements between WestinGhouse, INITEC and Bechtel for AP600 program design 

activities. 

Observations and findinas 

Westinghouse appropriately specified all technical and quality assurance 

requirements consistent with WCAP-12600. As required by the agreement signed 

on March 5, 1992, INITEC provided a QAPP for ?Vestinghouse‘s review and 

approval. 

As described in Section 3.3 above, in September 1994, Westinghouse contracted 

the services of Bechtel for an independent review of INITEC’s analysis and design 

methodology for the nuclear island basemat. The results of Bechtel‘s independent 

review were documented and made available to Westinghouse in a report dated 

November 1 1, 1994. 

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Although Westinghouse had identified the xbmittal of this independent revie\v 

report as a commitment in its August 2, 1994, response to the NRC, the inspectors 

were informed that a submittal of the report on the AP600 docket was still 

outstanding. No explanation was given as to why this report had not yet been made 

available for NRC review. 

The inspecfws learned that on August 3, 1994, Westinghouse had sent a letter to 

!NITEC (FOK/INlO 18 1 ) requesting that INITEC provide its response to the NRC audit 

findings {see Section 3.3.a.). However, it appeared that Westinghouse did not 

consider the results of the NRC structural audit findings and concerns in its Janoar:, 

1995 annual review of INITEC's performance. 

Additionz!ly, the inspectors reviewed hestinghouse's Audit Report QLA/IN10007, 

dated March 20, 1995, that documented a triennial audit conducted on February 20 

through 22, 1S95, at INITEC's facilities in Madrid, Spain. Upon reviewing the report, 

the inspectors found that: (1) an evaluai:isri of INITEC's response to Wc-tinghous;e's 

August 3, 1994, letter had not been included in the audit scope, and (21 the audit 

did not identify any evidence to suggest that INITEC had initiated any internal root 

cause analysis, 2nd evaluation. Further, th + inspectors determined that no corrective 

actions had been formally identified by INITEC's OA organization to determine the 

cause, and document the impact of the design deficiencies identified in 

Westinghouse's August 3, 1 994, letter, on INITEC's AP600 design deliverables. 

Finally, the inspectors questioned Westinghouse's conclusion, in its March 22, 1995, 

letter to INITEC, "Westinghouse AP600 Audit WES-95-2 7 1 ," that "INITEC Design 

Quality Assurance Program and implementing prOCedtJreS meet the applicable NQA- 1 

requirements for INITEC's AP600 work," in light of the fact that the audit report 

identified 10 findings of programma5c deficiencies in INITEC's implementation of its 

QAPP. 

c. Conclusions 

Based on the above, the inspectors concluded that Westinghouse's (1) failed to 

adequately evalcrsie or assess INITEC's performance, as required by WCAP- 1 2600, 

for a supplier of AP600 c'--ign deliverables that had been the subject of an adverse 

NRC audit finding, and (2) failed to conduct an evaluation of INITEC's response to 

Westinghouse's August 3, 1994, letter, and any associated corrective actions taken. 

This issue was identified as Nonconformance 99900404197-01 -02. 

3.5 Westinahouse Oversiaht of AP600 Desian Activities 

Based on the nonconformances identified above, the NRC is cnncerned that these 

quality assurance deficiencies may have introduced a level of uncertainty on the 

acceptability of design deliverables provided by AP600 technical cooperation 

agreement participants. Of particular concern to the NRC, is Westinghouse's failure 

to recognize and appropriately address a condition adverse to quality, requiring a root 

cause evaluation and determination and appropriate corrective actions, even when 

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such a condition was identified by an NRC audit 3nd resulted in re-design of the 

AP600 foundation basemat. 

Westinghouse's failure to address this design and quality assurance program 

deficiency in a timely manner has raised the issue of whether this is an isolated case 

and thar other design delivsrables provided b$ AP600 technical cooperation 

agreement participants do in fact possess the level of integrity in design verification 

and quality assurance necessary to satisfy the design certification provisions of 10 

CFR Part 52. This issue was iaentified as Unresolved Item 99900404/97-01-03. 

3.6 Entrance and Exit Meetinas 

Since this was a one-day inspection, the entrance and exit meetings were both held 

on April 17, 1997. In the entrance meeting the NRC inspectors discussed the scope 

Of the inspection and outlined the areas to be inspected. In the exit meeting the 

inspectors discussed the inspection findings and unresolved item. 

4 PERSONNEL CONTACTED 

Wesiinahouse Electric CorDoration 

David Alsing, AP600 Quality Systems Manager 

Ken Kloes, Projects Quality Assurance Engineer 

Robert Tupper, Advanced Technology, Project Engineer 

Richard Orr, AP600 Structures Advisory Engineer 

Narenda Prasad, Engineering Technology Department, Fellow Engineer 

Donald Lindgren, AP600 Licensing 

ITEMS OPENED, CLOSED, AND DISCUSSED 

99900404/97-0 1-01 NON inadequate corrective action 

99900404/97-0 1-02 NON inadequate quality and technical oversight of INITEC 

99900404/97-01-03 URI acceptability of AP600 design deliverables 

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Selected Generic Correspondence on the Adequacy of 

Vendor Audits and the Quality of Vendor Products 

Identifier - Title 

Information Notice 97-1 5 Reporting of Errors and Changes in Large-Break Loss-of-Coolant 

Accident Evaluation Models of Fuel Vendors and Compliance with 

1 OCFR50.46(a)(3) 

Information Notice 97-32 Defective Worm Shaft Clutch Gears in Limitorque Motor-Operated 

Valve Actuators 

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2-89) 

JRCM 1102, 

Clol, 3232 

~~~~~ ~ 

US. NUCLEAR REGULATORY COMMISSION 

BIBLIOGRAPHIC DATA SHEET 

(See instroctions on the reverse) 

2. TITLE AND SUBTITLE 

Licensee Contractor and Vendor Inspection 

Status Report 

Quarterly Report 

April - June 1997 

5. AUTHOR(S) 

1. REPORT NUMBER 

(Assigned by NRC, Add Vol., Supp., Rev., 

and Addendum Numbers. if any.) 

NUREG-0040 

Vol. 21, No. 2 

3. DATE REPORT PUBLISHED 

MONTH I YEAR 

November 1997 

1. FIN OR GRANT NUMBER 

6. TYPE OF REPORT 

Quarterly 

7. PERIOD COVERED (Inclusive Deks) 

April - June 1997 

3. PERFORMING ORGANIZATION - NAME AND ADDRESS (IfNRC, provrde Divism, Office or Region, US Nuclear Regulatory Commission, and mailing address, rfconkaclw, 

provide name and marlifg address J 

Division of Reactor Controls and Human Factors 

Office of Nuclear Reactor Regulation 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

I. SPONSORING ORGANIZATION - NAME AND ADDRESS (/fNRC, lype .Same as above" if coobacl~ provide NRC Division, Offa or Regron, U S Nuclaar Regulatory Commissron, 

and marling address ) 

Same as 8. above 

IO. SUPPLEMENTARY NOTES 

11. ABSTRACT (200 wwds or less) 

This periodical covers the results of Inspections performed between April 1997 and June 1997 by the NRC's Special Inspection 

Branch, Vendor Inspection Section, that have been distributed to the inspected organizations. 

13. AVAILABILITY STATEMENT 2. KEY WORDSlDESCRlPTORS (List words or phrases lhat will assist researchers in haling the repet.) 

Vendor Inspection unlimited 

14. SECURITY CLASSIFICATION 

(This Page) 

unclassified 

(This Rem) 

I 

NRC FORM 335(2-89) This form was electronically produced by Elite Federal Forms, Inc