NUREG-0040 

Licensee Contractor 

and Vendor Inspection 

Status Report 

Quarterly Report 

October - December 1994 

U.S. Nuclear Regulatory Commission 

Office of Nuclear Reactor Regulation 
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2. 

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A year's subscription of this report consists of four quarterly issues. 
NUREG-0040 

Vol. 18, No. 4 

Licensee Contractor 

and Vendor Inspection 

Status Report 

Quarterly Report 

October - September 1994 

Manuscript Completed: February 1995 

Date Published: February 1995 

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

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

employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or 

process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial praiuct, process, or service by trade name, trademark, 

manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views 

and opinions of authors expressed herein do not necessarily state or reflect those of the 

United States Government or any agency thereof. 

Division of Technical Support 

Office Of Nuclear Reactor Regulation 

Washington, DC-2055510001 - 

s 
DISCLAIMER 

Portions of this document may be illegible 

in electronic image products. Images are 

produced from the best availabie original 

document. 

.- 

I 
ABSTRACT 

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

Special Inspection Branch, Vendor Inspection Section, that have been 

distributed to the inspect.ed organizations during the period from October 1994 

through December 1994. 

TABLE OF CONTENTS 

PAGE 

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

Preface .................................................................. vii 

Inspection Reports ....................................................... ix 

Index .................................................................... x 

Selected Bulletins, Generic Letters, and Information Notices 

Concerning Adequacy of Vendor Audits and Qual ity of 

Vendor Products .......................................................... 153 

Correspondence Re1 ated To Vendor Issues .................................. 154 

V 

PREFACE 

A'-fundamental premise of the Nuclear Regulatory Commission (NRC) 1 icensing and 

inspection program is that licensees are responsible for the proper construction and safe and efficient operation of their nuclear power plants. The 

total government-industry system for the inspection of commercial nuclear 

facilities has been designed to provide for multiple levels of inspection and 

verification. Licensees , contractors , and vendors each participate in a 

quality verification process in compliance with requirements prescribed by the 

NRC's rules and regulations (Title 10 Code of Federal Regulations). 

performs an overview of the commercial nuclear industry by inspection 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 ongoing quality verification programs. 

The NRC 

The licensee is responsible for developing and maintaining a detailed quality 

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

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

is responsible for assuring that suppliers, contractors and vendors also have 

sui table and appropriate quality programs that meet NRC requirements, guides, 

codes and standards. 

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

and inspects nuclear steam system suppliers (NSSSs), architect engineering (AE) firms, suppliers of products and services, independent testing 

laboratories performing equipment qualification tests, and holders of NRC 

licenses (construction permit holders and operating licenses) in vendorrelated areas. 

of reported vendor-related problems are determined and appropriate corrective 

actions are developed. The inspections also review the vendors' conformance 

with applicable NRC and industry quality requirements, the adequacy of 

licensees' oversight of their vendors, and that adequate interfaces exist 

between licensees and vendors. 

These inspections are performed to assure that the root causes 

The VIS inspection emphasis is placed on the quality and suitability of vendor 

products, licensee-vendor interface, environmental qualification of equipment, 

and review of equipment problems found during operation and their corrective 

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

the inspected organization is required to take appropriate corrective action 

and to institute preventive measures to preclude recurrence. 

implications are identified, NRC assures that affected licensees are informed 

through vendor reporting or by NRC generic correspondence such as information 

notices and bulletins. 

When generic 

NUREG 0040, "Licensee Contractor and Vendor Inspection Status Report," is 

pub1 ished quarterly and contains copies of all vendor inspection reports 

issued during the calendar quarter for which it is published. 

inspection report lists the nuclear facilities to which the results are 

applicable thereby informing licensees and vendors of potential problems. 

addition, the affected Regional Offices are notified of any significant 

problem areas that may require special attention. 

Each vendor 

In 

NUREG 0040 also contains a 

vi i 
list of selected bulletins, generic letters, information notices, and other 

pertinent correspondence involving vendor issues. 

Correspondence with contractors and vendors relative to inspection data 

contained in NUREG 0040 is placed in the USNRC Public Document Room, located 

in Washington, D.C. 

viii 
INSPECT I ON REPORTS 

ix 
INDEX 

FAC I L ITY 

ABB Combustion Engineering, Inc. 

Windsor, Connecticut 

ABB Service Company 

C1 eve1 and, Ohi o 

Cardinal Industrial Products 

Las Vegas, Nevada 

GE Nuclear Energy 

San Jose, California 

Tennessee Valley Authority 

Watts Bar Nuclear Plant, Unit 1 

Chattanooga, Tennessee 

REPORT NUMBER - PAGE 

99900102/94-01 1 

99901281/94-01 48 

99901076/94-01 59 

99900403/94-02 79 

50-390/94-201 99 

X 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 205554001 

December 29, 1994 

Mr. Donald E. Allen 

Vice President, Total Quality 

Combustion Engineering, Inc. 

ABB Combustion Engineering 

Windsor, Connecticut 06095-0500 

Nuclear Operations 

P.O. Box 500 

SUBJECT: NRC INSPECTION NO. 99900002/94-01; 99900102/94-01 

Dear Mr. Allen: 

This letter transmits the report of the U.S. Nuclear Regulatory 

Comm'is~on (NRC) inspection conducted October 3 through 7, 1994, 

at the Fuel Operations-Windsor (CENFO) facility of Asea Brown 

Boveri (ABB) Combustion Engineering Nuclear Operations (ABB CENO) 

in Windsor, Connecticut; and the inspection conducted October 17 

through 21, 1994, at the ABB CENO Fuel Operations;, Nuclear Fuel 

Manufacturing-Hematite (CENFM) facility in Hemati-te, Missouri. 

The inspection of CENFO in Windsor, Connecticut, also included 

inspection of the fuel related activities of ABB CENOs 

Engineering Operations and Field Services organizations. 

inspection team, led by Steven M. Matthews, conducted a 

performance-based evaluation of the CENFO and CEbIFM management, 

staff, and quality programs and the implementation of those ~ 

programs related to pressurized- and boiling-water reactor core 

reload analysis design, fuel assemblies, and other fuel-related 

services supplied to the U.S. nuclear industry. 

The NRC 

The team (a) examined technical documentation, procedures, and 

representative records, (b) interviewed ABB CENO personnel, 

(c) held discussions with ABB CENO personnel, (d) listened to 

presentations by ABB CENO personnel, and (e) observed work 

activities. On the basis of this inspection, the staff 

determined that the implementation of the ABB CENO Topical 

Report, document CENPD-210A, Revision 7A, which had been approved 

by the NRC on February 17, 1993, as meeting the requirements of 

Appendix B to Part 50 of Title 10 of the Code of Federal 

Resulations (Appendix B to 10 CFR Part 50), was deficient in 

certain areas and, on this basis, we are issuing a notice of 

nonconformance (Enclosure 1) to ABB CENO. The irispection report 

(Enclosure 2) contains a detailed discussion of t:he areas 

examined during the inspection and our findings. 

1 
D. Allen -2- 

The most important inspection finding was ABB CENOs failure, in 

several instances, to implement established procedures for 

verifying or checking the adequacy of design data and, in other 

instances, to establish adequate measures for the control of 

design interfaces. Regarding the adequacy of design data, 

certain unvalidated physics biases and uncertainty data were used 

by Arizona Public Service Company (APS) in its Palo Verde Unit 1 

Cycle 4 reload safety analysis. 

and final verification of the reload design analysis did not 

recognize the use of the unvalidated data before its release to 

and use by APS. 

Regarding control of design interfaces, ABB CENOs design 

interface activities were based on the assumption that data, 

as core reload specification data, may be referenced as design 

input without verification. In the instances described in the 

report, ABB CENOs design interface activities and controls were 

not -?equate to ensure that the information in thc core reload 

specification data was consistent with the existing core analysis 

and safety analysis calculations for which ABB CENO is 

responsible. 

The team observed that the joint engineering effort between ABB 

CENO and APS was a strength of both organizations. A significant 

part of certain Palo Verde reload analyses were performed by APS 

nuclear fuel personnel. The first reload in this arrangement was 

jointly analyzed by APS and Engineering Operations engineers. 

For the next reload, APS engineers performed the work at APS with 

Engineering Operations engineers reviewing the work. 

Please respond to this letter following the instructions in the 

enclosed notice of nonconformance. Your response should document 

the specific actions taken and any additional actions you plan in 

order to prevent recurrence. 

within 30 days from the date of this letter. 

In accordance with 10 CFR 2.790 (a) of the NRCs "Rules of 

Practice," a copy of this letter and its enclosures will be 

placed in the NRC Public Document Room. 

by this letter and the enclosed notice are not subject to the 

clearance procedures of the Office of Management and Budget as 

required by the Paperwork Reduction Act of 1980, Public Law 

Engineering Operations review 

such 

Please send us written response 

The responses requested 

NO. 96-511. 

2 
D. Allen -3- 

Should you have any questions concerning this inspection, we will 

be pleased to discuss them with you. 

cooperation during this process. 

Thank you for your 

Sincerely, 

Special Inspection Branch 

Division of Technical Support 

Office of Nuclear Reactor Regulation 

Docket Nos.: 

99900002 (CENFO) 

99900102 (CENFM) 

Enclosures: 

1. Notice of Nonconformance 

2. Inspection Report: 

99900002/94-01; 99900102/94-01 

3 
NOTICE OF NONCONFORMANCE 

Combustion Engineering, Inc. 

ABB Combustion Engineering 

Windsor, Connecticut, and 

Hematite, Missouri 

Nuclear Operations 

Docket Nos.: 99900002 

99900102 

Report No.: 94-01 

On the basis of the results of U.S. Nuclear Regulatory Commission 

(NRC) inspections conducted on October 3 through 7, 1994, at the 

facilities of Fuel Operations-Windsor (CENFO), Engineering 

Operations, and Field Services in Windsor, Connecticut, and on 

October 17 through 21, 1994, at the Nuclear Fuel ManufacturingHematite (CENFM) facility in Hematite, Missouri, it appears that 

certain of your activities were not performed in accordance with 

NRC rqLirements. 

A. Section 111.3, “Design Control,” of the Asea Brown Boveri 

(ABB) Combustion Engineering Nuclear Operations (ABB CENO) 

Topical Report, document CENPD-210A, Revision 7A, approved 

by the NRC on February 17, 1993, as meeting the requirements 

of Appendix B to Part 50 of Title 10 of the Code of Federal 

Resulations (Appendix B to 10 CFR Part 50), hereafter 

referred to as the “QA topical report,” required, in part, 

that the procedures cover the total design process to assure 

that design interfaces among participating organizations are 

identified and controlled, and that design work be verified 

or checked for adequacy. 

System 3, “Design Control,n of the ABB CENO Quality 

Assurance Manual, QAM-lOO-NF, First Edition, Revision 2, 

dated June 1, 1994 (QAM-lOO-NF), required, in part, that the 

cognizant engineer shall review incoming design information 

to ensure that it shows evidence of approval and is adequate 

for the design effort; and that where the independent 

reviewer uses design review as the method of design 

verification, assumptions necessary to perform the design 

have been adequately described and are reasonable, or, where 

necessary, are identified for subsequent reverification when 

the detailed activities are completed. 

Contrary to the above, ABB CENO failed in the following 

instances to implement established procedures for verifying 

or checking the adequacy of design and establish adequate 

measures for the control of design interfaces (94-01-01): 

(1) For the core reload specification data from Baltimore 

Gas & Electric Company (BG&E) for its Calvert Cliffs 

Enclosure 1 

4 
Unit 1 Cycle 12 core reload, and from Arizona Public 

Service Company (APS) for its Palo Verde Unit 3 Cycle 5 

core reload, Engineering Operations failed to comply 

with the requirements of QCP-3.6, "Design Interface 

Control," of QCM-lOl-NF, "Quality Control Manual, Fuel 

Engineering and Fuel Development," Revision 1, dated 

November 15, 1993, (QCM-101-NF). Specifically, 

Engineering Operations failed to comply with the 

requirements of QCP-3.6 by not assuring the 

identification of (a) the quality program under which 

the licensees' core reload specification data was 

verified and (b) the verification status of the 

licensees' core reload specification data. For the 

Palo Verde Unit 3 Cycle 5 reload design, Engineering 

Operations also failed to follow the interface 

procedure documented in QC 93-01, "Quality Plan for 

Activities Performed under the Arizona Public 

Service/ABB Engineering Partnership Agreement." Since 

QC 93-01 also imposed the requirements of QCP-3.6, 

Engineering Operations did not comply with the 

requirements of QC 93-01 and QCP-3.6 in that it failed 

to assure the identification of (a) the quality program 

under which the APS design specification data was 

verified, and (b) the verification status of the APS 

design specification data. 

In addition, Engineering Operations design interface 

activities and controls were based on the assumption 

stated in QCP-3.3, "Design Input," of QCM-lOl-NF, which 

provides, in part, that customer fuel reload 

specification data may be referenced .as input and 

assumed to be correct. 

interface activities and controls were based on the 

assumption stated in QCP-3.3 and, therefore, ABB CENOs 

design interface activities and controls did not ensure 

the information in the core reload specification data 

was consistent with the existing core analysis and 

safety analysis calculations for which ABB CENO is 

responsible. (Report paragraph 4.3.3.2(1)) 

(2) The physics biases and uncertainties data, generated by 

computer code, "ROCS/.NEM," and used i:n the Palo Verde 

Unit 1 Cycle 4 reload safety analysis, were not 

(a) validated data when it was issued to APS on 

December 12, 1991, in reload safety a:nalysis V-91-371 

(the data was eventually validated after September 

1992), (b) identified in the reload safety analysis as 

unvalidated data as required by QCP-3.10, "Other Design 

Documents," of QCM-lOl-NF, and (c) identified by 

Engineering Operations during the author's review and 

the independent reviewer's final verification of the 

reload design analysis before the ROC;S/NEM biases and 

Engineering 0:perations design 

-2- 

5 
uncertainties were released to and used by ApS, as 

required by QCP-3.2, “Design Analysis,” and QCP-3.4, 

“Design Document Verification,” of QCM-101-NF. 

paragraph 4.3.3.3) 

TCR on 14x14 Grid Pin Pitch,” used to support BG&E‘s 

Calvert Cliffs Unit 1 Cycle 12 core reload fuel 

mechanical design, included a neutronics analysis to 

determine the effect of increasing spacer grid 

dimensional tolerances beyond the current limits. 

neutronics analysis did not document an adequate 

justification for borrowing the rod-bowing penalty to 

accommodate the increased grid spacer tolerances as 

required by QCP-3.2 of QCM-101-NF; which required that 

the methods used in an analysis be documented and 

justified. (Report paragraph 4.3.3.5) 

All five coefficient input errors to the computer code, 

“CECOR,” for the Calvert Cliffs Unit 1 Cycle 11 reload 

and the Cycle 12 CECOR coefficient input error were 

missed by Engineering Operations during its review and 

final verification of the reload design analysis as 

required by QCP-3.2 and QCP-3.4 of QCM-101-NF. 

addition, Engineering Operations had not established an 

input specification for the CECOR code inputs for 

Calvert Cliffs Unit 1 Cycle 11 and Cycle 12 core reload 

CECOR coefficients as required by QCP-3.13, ”Computer 

Software,” of QCM-101-NF. (Report paragraph 4.3.3.7) 

(Report 

(3) The physics analysis report, “Physics Justification of 

The 

(4) 

In 

B. Section 111.16, “Corrective Action,” of the QA topical 

report required, in part, that conditions adverse to quality 

be documented and reported to cognizant parties; 

corrective action be identified and implemented in a timely 

manner; and that procedures require identification of cause, 

corrective action to prevent recurrence, and verification of 

implementation. 

System 16, “Corrective Action,” of QAM-lOO-NF, required, in 

part, that conditions adverse to quality shall be identified 

promptly and corrected as soon as practical; that the cause 

of the condition be investigated and appropriate corrective 

action taken to preclude recurrence; and that the adverse 

condition, the cause or causes, and the corrective action 

taken be documented and reported to appropriate levels of 

ABB CENO management. 

Contrary to the above, ABB CENO failed in the following 

instances, where significant conditions adverse to quality 

existed, to assure that the cause of the condition was 

determined; the adverse condition was promptly identified 

and corrected; or the corrective action taken precluded 

that 

-3- 

6 
repetition. In other instances cited below, ABB CENO failed 

to establish adequate measures to assure that conditions 

adverse to quality were promptly identified and corrected 

(94-01-02). 

(1) Zirconium (Zr) alloy, Zr4, fuel clad tubing lot DHS12R 

supplied to CENFO by Sandvik Special Metals Corporation 

(SSM) of Kennewick, Nashington, was reworked by SSM 

because the material lots initial burst/elongation test 

specimens failed the circumferential elongation 

requirements as the result of, accordfing to SSM, 

microcracks on the outside surface of the tubing. 

CENFO approved the lot of fuel clad tubing for use and 

SSM shipped the lot to CENFM without determining the 

cause of the microcrack condition and submitting to 

CENFO its assignable root cause, as required by QCP516.1, “Corrective Action,” of QCPM-500, “Quality 

Control Procedures Manual, Nuclear Fuel ManufacturingWindsor,” Revision 3, dated September 15, 1994 (QCPM500). (Report paragraph 4.3.2.1(1)) 

Corrective Action Request (CAR) IQS-94-002 that 

identified that no procedures were established for the 

verification, control, and use of computer software for 

certain numerically controlled machines in the cage 

fabrication factory was issued with a response due date 

of August 8, 1994. Even though QCP-53.6.1 of QCPM-500 

required the corrective action analyses and planning to 

be completed and the response returned to the 

originator within 30 days, the response to CAR IQS-94- 

002 was not written until September 20, 1994. (Report 

paragraph 4.3.2.3) 

(3) For the five computer code, CECOR, coefficient input 

errors reported for Calvert Cliffs Unit 1 Cycle 11, 

described in A.(4) above, Engineering Operations found 

the assignable root cause to be that the CECOR code at 

ABB CENO was different than the version at BG&E. 

Engineering Operations corrective action was the 

development of an electronic interface between the BASS 

and CECOR computer codes. The corrective action from 

Cycle 11 was the creation of an electronic interface in 

the form of a template file which was processed by the 

BASS-P5 code to create the BASS coefficients for the 

plant. However, the BASS-P5 code did not trap this 

type of input error because the CECOR coefficient input 

error reported for Cycle 12, was also the input 

parameter in error in Cycle 11. 

Even though QCP-16.1, “Corrective Action,” of QCM-101- 

NF required the corrective action to identify action 

taken to prevent recurrence, Engineering Operations 

corrective action taken for the Cycle 11 input errors 

-4- 

7 
C. 

did not prevent a recurrence of an input error for the 

same input parameter in Cycle 12. 

4.3.3.7) 

(Report paragraph 

(4) QCP-516.1 of QCPM-500H assigned responsibility to the 

cognizant quality coordinator in each of CENF'Ms focused 

factories for the issuance and control of CARS for 

traceable quality problems identified by P/DNs. 

However, QCP-516.1 did not address the responsibilities 

of the cognizant quality coordinators in CENFMs focused 

factory environment where the corrective action 

analysis to determine assignable root causes involve 

interface activities between more than one focused 

factory. 

responsibilities between cognizant quality coordinators 

contributed to the delay in its issuance of CAR H-1-94. 

(Report paragraph 4.4.1.1(2)) 

CENFMs failure to address the interface 

(5) In October 1991, CENFM identified a malfunctioning 

pulse/pause timer as the assignable root cause of 

enrichment variations that resulted f,,m nolihomogeneous lots or batches of uranium oxide powder. 

CENFMs preliminary problem analysis of the fuel pellet 

isotopic deviations, described in B.(4) above, had 

determined the assignable root cause was a malfunction 

in the powder blenders' pulse/pause timer; a recurrence 

of the 1991 event. Therefore, CENFMs corrective 

actions taken for the 1991 event were not adequate to 

prevent recurrence as required by QCP-516.1 of QCPM500H. (Report paragraph 4.4.1.1(3)) 

Section 111.5, "Instructions, Procedures, and of 

the QA topical report required, in part, that activities 

affecting quality be prescribed by and performed in 

accordance with documented instructions, procedures, or 

drawings, and that these documents include or reference 

quantitative acceptance criteria for determining tnat the 

activity has been satisfactorily accomplished. 

System 5, "Instructions, Procedures, and Drawings," of QAMlOO-NF, required, in part, that operation sheets (OSs) shall 

be used to provide detailed instructions which describe the 

technique, equipment, tooling, and acceptance criteria to be 

used to perform specific manufacturing or inspection 

activities, and which may also provide detailed instructions 

for the accomplishment of other activities affecting 

quality. 

Contrary to the above, ABB CENO failed in the following 

instances to prescribe certain activities that affect 

quality and assure that activities that affect quality are 

accomplished in accordance with instructions or procedures 

(94-01-04) : 

-5- 

8 
PQ-lC, Revision 7, dated October 12, 1994, "Process 

Qualification Plan and Report for Enrichment Blending 

Unit Process (UPIC)," the procedure used to requalify 

the enrichment blenders, addressed (a) an increase in 

the qualified enrichment spread and (13) the sampling 

plan to requalify the blenders after recovering from 

the problems that caused the isotopic deviations 

described in CAR H-1-94. However, PQ-1C did not 

provide guidance for a qualified procedure or establish 

controlled conditions to assure process qualification 

as required by QCP-509.2, "Control of Special 

Processes," of QCPM-500H and QCH-2, "I?rocedure for Unit 

Process Qualification," Revision 2, dated June 23, 

1994. (Report paragraph 4.4.1.1(4)) 

The prestack key plan and rod pushing key plan for Part 

AEK000, Shop Order CC2NN6, were issued to the rod line 

without the approval of the cognizant quality 

organization as required by Operating Sheet (OS) 508, 

"Document Control," Revision 4, dated October 10, 1994. 

(Report paragraph 4.4.2.2) 

The pushing table grooves were not bejag cleaned at the 

beginning of each shift as required by OS 3320, "Rod 

Loading - Bundle Assembly," Revision 5, dated June 8, 

1994. (Report paragraph 4.4.2.4) 

The pushing table water system was not: being changed 

after approximately every 10 fuel assemblies as 

required by OS 3320. (Report paragraph 4.4.2.4) 

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

Nuclear Regulatory Commission, ATTN: Document Control Desk, 

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

Inspection Branch, Division of Technical Support:, Office of 

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

letter transmitting this Notice of Nonc0nformanc:e. 

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

Nonconformance" and should include the following: (1) a 

description of steps that have been or will be t:aken to correct 

this item; (2) a description of steps that have been or will be 

taken to prevent recurrence; and (3) the dates your corrective 

actions and preventive measures were or will be completed. 

This reply 

Dated at Rockville, Maryland 

this 29th day of December, 1994 

-6- 

9 
ORGANIZATION: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF TECHNICAL SUPPORT 

REPORT NO.: 

CORRESPONDENCE 

ADDRESS : 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY : 

INSPECTION DATES: 

LEAD INSPECTOR: 

Combustion Engineering, Inc., 

Asea Brown Boveri (ABB) Combustion 

Engineering Nuclear Operations (ABB CENO); 

Fuel Operations-Windsor (CENFO); and 

Nuclear Fuel Manufacturing-Hematite (CENFM) 

99900002/94-01; 99900102/94-01 

Donald E. Allen 

Vice President, Total Quality 

Combustion Engineering, Inc. 

ABB Combustion Engineering F?clear Operations 

P.O. Box 500 

Windsor, Connecticut 06095-0500 

W. Carlton Coppersmith, PhD. 

Director, Total Quality 

ABB CENO provides pressurized- and boilingwater reactor (PWR and BWR) core reload 

analysis designs, fuel assemblies, and 

nuclear-fuel-related services to the U.S. 

nuclear industry. 

October 3 through 7, 1994 - CENFO 

October 17 through 21, 1994 - CENFM 

Vendor Irkpection Section (TVIS) 

Special Inspection Branch (SIB) 

Division of Technical Support (DOTS) 

Office of Nuclear Reactor Regulation (NRR) 

OTHER INSPECTORS: David H. Brewer, TVIS/SIB/DOTS/NRR 

John F. Carew, Brookhaven National Laboratory 

Ramon L. Cilimberg, Parbmeter, Inc. 

Rodney L. Grow, Parbmeter, Inc. 

Edward D. Kendrick, Reactor Systems Branch 

(SRXB), Division of Systems Safety and 

Analysis (DSSA), NRR 

Kombiz Salehi, SRXB/DSSA/NRR 

Enclosure 2 

10 
REVIEWED BY: 

APPROVED BY: 

. Cwalina, Chief Date 

I 

Robert M. Gallo, Cwef 

SIB/DOTS/NRR 

-2- 
1 

2 

INSPECTION BASES: 

Appendix A, “General Design Criteria for Nuclear Power 

Plants,” General Design Criterion (GDC) 10, “Reactor 

Design,” and GDC 12, “Suppression of Reactor Power 

Oscillations,” to Part 50 of Title 10 of the Code of 

Federal Resulations (10 CFR Part 50) 

Appendix B, “Quality Assurance Criteria for Nuclear 

Power Plants and Fuel Reprocessing Plants,” to 10 CFR 

Part 50. 

Part 21, “Notification of Failure to Comply or 

Existence of a Defect,” of 10 CFR 

Asea Brown Boveri (ABB) Combustion Engineering Nuclear 

Operations (ABB CENO), Topical Report, document CENPD210A, Revision 7A, approved by the NRC on February 17, 

1993, as meeting the requirements of Appendix B to 10 

CFR Part 50, hereafter refereed to as the “QA topical 

report” 

SUMMARY OF INSPECTION F’INDINGS: 

2.1 Violations 

No violations were identified during this inspection. 

2.2 Nonconformances 

2.2.1 Nonconformance 94-01-01 

This nonconformance, described in Sections 4.3.3.2(1), 4.3.3.3, 

4.3.3.5, and 4.3.3.7, of this report, identifies 4 instances 

where the design control activities and established measures of 

ABB CENO failed to comply with requirements of the QA topical 

report and the ABB CENO Quality Assurance Manual, QAM-lOO-NF, 

First Edition, Revision 2, dated June 1, 1994 (QAM-100-NF). 

2.2.2 Nonconformance 94-01-02 

This nonconformance, described in Sections 4.3.2.1(1), 4.3.2.3, 

4.3.3.7, 4.4.1.1(2), and 4.4.1.1(3), of this report, identifies 5 

instances where corrective action activities and established 

measures failed to comply with the requirements of the QA topical 

report and QAM-100-NF. 

-3- 

12 
2.2.3 Nonconformance 94-01-03 

This nonconformance, described in Sections 4.4.1.1(4), 4.4.2.2, 

and 4.4.2.4 (2 places) of this report, identifies 4 instances 

where certain quality activities were not accomplished in 

accordance with instructions or procedures, and established 

measures for certain quality activities failed to comply with the 

requirements of the QA topical report and QAM-100-NF. 

2.3 Own Items 

2.3.1 Open Item 94-01-04 

As described in Section 4.3.2.1(1) of this report, the NRC 

inspection team (team) was given a root cause analysis report for 

the failed burst/elongation test specimens from :Lot DHS12 of Zr4 

fuel clad tubing supplied by Sandvik Specialty Metals Corporation 

(SSM) of Kennewick, Washington. However, during a subsequent 

inspection of SSM on December 5 through 7, 1994, the team was 

adviepd by SSM that its evaluation was not complete and that it 

would be submitting to ABB CENO a final report wyith the assigned 

root cause. The team requested that NRC be notified when ABB 

CENO has completed action on this item. 

2.3.2 Open Item 94-01-05 

As described in Section 4.4.1.1 of this report, the team observed 

ABB CENOs response to an enrichment variation event identified by 

CENFM when fuel pellet isotopic deviations of varying degrees 

were identified by rod scanner profiles. The team requested that 

NRC be notified when corrective action has been c:ompleted. 

2.3.3 Open Item 94-01-06 

As described in Section 4.4.2.1 of this report, the team observed 

CENFMs response to surface indications on Zr4 fuel clad tubing 

identified as gouges or pitting on the outside surface. The team 

requested that NRC be notified when the corrective action has 

been completed. 

2.3.4 Open Item 94-01-07 

As described in Section 4.4.2.1 of this report, t:he team learned 

that since July 1994 CENFM had observed tubing surface 

indications of blue, pink, black, and gray ink; brown stains; and 

white deposits. The team requested that NRC be notified when the 

corrective action has been completed. 

-4- 

13 
3 STATUS OF PREVIOUS INSPECTION FINDINGS 

Previous inspection findings were not addressed during this 

inspection. 

4 INSPECTION FINDl[NGS AND OTHER COMMENTS 

4.1 Backmound 

ABB CENO was formed in October 1993, when ABB Combustion 

Engineering Services and ABB Combustion Engineering Nuclear Fuel 

were combined. According to ABB CENO, as part of the 

reorganization, Engineering and Development in both of the former 

organizations were combined into a new organization, Engineering 

Operations. As a result of the reorganization, the major groups 

within the ABB CENO organization include: 

e Total Quality 

e Fuel Operations (CENFO) 

e Engineering Operations 

e Field Services 

A sister company, ABB Atom AB, located in Vaster&, Sweden, is 

also involved in the activities of the U.S. nuclear industry, 

chiefly BWR related. ABB Atom is a subsidiary of Asea Brown 

Boveri AB of the ABB Asea Brown Boveri Group. 

4.2 Entrance Meetinm, Interim Exit Meetins and Final Exit Meeting 

During the entrance meeting in Windsor, Connecticut on October 3, 

1994, the team met with members of ABB CENO management and staff, 

discussed the scope of the inspection, and established contact 

persons for the team within the management and staff of the 

CENFO, Engineering Operations, and Field Services organizations. 

On October 7, 1994, the team held an interim exit meeting to 

outline to ABB CENO management and staff the major concerns 

identified by the team thus far. 

During the entrance meeting in Hematite, Missouri, on October 17, 

1994, the team met with members of CENFM management and staff, 

discussed the scope of the inspection, and established the team’s 

contact persons within the management and staff of CENFM. 

October 21, 1994, the team conducted an interim exit meeting to 

outline to CENFM management and staff major concerns identified 

by the team during this portion of the inspection. 

During both inspection periods described above, the team 

conducted a performance-based inspection of the ABB CENO CENFO, 

Engineering Operations, and Field Services organizations through 

technically directed observations and evaluations of processes, 

activities, and documentation. The team (a) examined technical 

On 

-5- 

14 
documentation, prccedures, and representative records, 

(b) interviewed ABB CENO personnel, (c) held discussions with ABB 

CENO personnel, (d) listened to presentations by ABB CENO 

personnel, and (e) made other observations. 

The specific areas examined, the documentation reviewed, and the 

team's findings are described in this report. The persons who 

participated in and who were contacted during this inspection are 

listed in the Appendix to this report. 

During its exit meeting in Windsor, Connecticut, on November 2, 

1994, with ABB CENO management and staff, the team summarized the 

inspection findings, open items, weaknesses, and observations. 

4.3 October 3 through 7. 1994 - Windsor, Connecticut 

During this period of the inspection, the team evaluated ABB 

CENOs nuclear-fuel-related activities in Windsor, Connecticut. 

4.3.1 Total Quality 

Established by the authority of the President of ABB CENO, the 

quality assurance program for nuclear fuel operations is assigned 

to the Vice President, Total Quality. According to ABB CENO, 

Total Quality is the organization within ABB CENO that is 

responsible for the quality assurance (QA) functions of 

developing and maintaining quality assurance manuals and 

conducting independent internal quality audits. For nuclearfuel-related activities, quality control (QC) functions are 

assigned to CENFOs focused factories in Windsor, Connecticut, and 

Hematite, Missouri; to Supply Management; and to Engineering 

Operations. Quality coordinators in the CENFO focused factories, 

Supply Management, and the Total Quality representative in 

Engineering Operations (analysis development) and CENFOs Fuel 

Mechanical Design group have unimpeded access to Total Quality 

and have stop-work authority from Total Quality. 

Total Quality is responsible for preparing and maintaining the 

quality assurance program, as that program is expressed in the QA 

topical report and the ABB CENO Quality Assurance Manual, QAMlOO-NF, First Edition, Revision 2, dated June 1, 1994 (QAM-100- 

NF). The following documents, prepared and maintained under the 

direction of Total Quality and used to implement the QA topical 

report and QAM-100-NF in the various ABB CENO organizations, were 

used by the team during its inspection of ABB CENOs nuclear-fuelrelated activities: 

0 QSP-lOO-NF, "Quality Systems Implementing Procedures," 

Revision 1, dated June 1, 1994 (QSP-100-NF) 

-6- 

15 
a 

e 

9 

QCM-lOl-NF, “Quality Control Manual, Fuel Engineering 

and Fuel Development,” Revision 1, dated November 15, 

1993 (QCM-101-NF) 

QCPM-500, “Quality Control Procedures Manual, Nuclear 

Fuel Manufacturing-Windsor,” Revision 3, dated 

September 15, 1994 (QCPM-500) 

QCPM-500H, “Quality Control Procedures Manual, Nuclear 

Fuel Manufacturing-Hematite,” Revision 2, dated 

October 1, 1994 (QCPM-500H) 

QAM-100, “Quality Assurance Manual, Nuclear Services,” 

Revision 0, dated November 12, 1994 (QAM-100) 

QAM-101, “Quality Assurance Procedures Manual,” 

Revision 2 (QAM-101) 

4.3.2 Fuel Operations 

ABB CENOs Fuel Operations maintains focused fartory facilities in 

Windsor, Connecticut, and Hematite, Missouri. 

he1 Operations-Windsor (CENFO) 

The CENFO focused factories facility in Windsor, Connecticut, 

fabricates components such as upper and lower end-fittings, 

rod end caps, control element assemblies (CEAs), spacer grids, 

and grid cage assemblies. CEAs are shipped directly to the 

licensee, whereas, the end-fittings, fuel rod end caps, and 

completed grid cage assemblies are shipped to CENFOs facility in 

Hematite, Missouri. The nuclear-fuel-related activities of CENFO 

were conducted by the following focused factories and 

departments: 

fuel 

Supply Management 

Grid and Cage Manufacturing 

Component Manufacturing 

Fuel Mechanical Design 

BWR Operations 

PWR Projects and Planning 

Total Quality 

Fuel Operations, Nuclear Fuel Manufacturing-Hematite (CENFM) 

In the CENFM facility uranium hexafluoride (UF,) is converted to 

uranium dioxide (U02); the UO, is pelletized; fucl rods are 

loaded using components shipped from CENFO; final fuel assembly 

takes place using grid cages shipped from CENFO; and completed 

-7- 

16 
fuel assemblies are shipped to licensees. 

of the nuclear-fuel-related activities of the CENFM facility is 

described in Section 4.4 of this report. 

The team's evaluation 

4.3.2.1 Supply Management 

The team selected the following purchase orders (Pos) to evaluate 

CENFOs procurement of certain materials used in fuel assemblies: 

PO 9491114-D5332, Supplement No. 3, to Teledyne Wah 

Chang (TWC) of Albany, Oregon, for Zr a.lloy strip 

material for spacer grids 

PO 9491170-D5332, Supplement No. 00, to Ulbrich 

Stainless Steel for Inconel strip material for 

Guardian, grids for 16x16 and System 80 fuel 

assemblies 

PO 9392152-D5602, Supplement No. 00, to TWC for 

centerless ground Zr4 rod to machine upper and lower 

fuel rod end caps 

PO 9491063-D5502 to SSM for Zr4 fuel clad tubing 

In each of these Pos, the requirements of 10 CFR Part 21 and 

QC-14-09, "Quality Assurance Program Requirements for Suppliers 

to ABB Combustion Engineering Fuel Operations," Revision 6, 

Class A, were invoked. QC-14-09 is a graded quality program; 

Class A is intended to meet the requirements of Appendix B to 10 

CFR Part 50. The certified material test reports for each PO 

contained data showing that the materials met all requirements. 

(1) Fuel Clad Tubing Material 

For fuel clad tubing, the team reviewed the data packages 

for Zr; fuel clad tubing lots and determined+that, with the 

exception of lot DHS12R, the material, represented by the 

data packages reviewed, met the requirements of CENFOs Pos. 

The material lots initial burst/elongation test specimens 

from lot DHS12 failed the circumferential elongation 

requirement of 2 12% with a test value of 7%; the 

burst/elongation test was used to characterize the biaxial 

mechanical properties of the Zr4 fuel clad tubing. 

According to SSMs request for approval or review 9491063- 

002, dated June 15, 1994, lot DHS12R, before reworking, 

apparently contained surface imperfections (described by SSM 

as microcracks) on the outside diameter (OD) of the Zr4 fuel 

clad tubing. After receiving approval to rework from CENFO, 

SSM removed the surface imperfections from the tubing by 

belt-abrading the OD surface. After .0004 inches were belt 

sanded from the OD of the tubing to remove the surface 

-8- 

17 
imperfections, burst/elongation tests was performed on 9 

samples of the reworked material. Samples of lot DHS12R 

(R standing for reworked) met the burst test circumferential 

elongation requirements with values that ranged from 15.1 to 

62.6%. 

As stated in its June 30, 1994, memo, to SSM approving the 

request for approval or review, CENFO approved the reworked 

fuel clad tubing material on the condition that SSM perform 

a root cause analysis on the failed specimens. However, SSM 

shipped the material to CENFM as acceptable for use without 

performing a root cause analysis. The team noted that the 

root cause analysis had not been performed, as required by 

the corrective action requirements of paragraph 5.3.4 of 

QCP-516.1, “Corrective Actions,” of QCPM-500. The team 

determined that CENFO had approved the lot of fuel clad 

tubing for use and that SSM shipped the lot to CENFM without 

determining the cause of the microcrack condition and 

submitting to CENFO a root cause analysis report. 

concluded that CENFOs failure to comply with QCP-516.1 by 

*7ot determining the assignable root cause oL the microcracks 

before approving the lot of fuel clad tubing for use was a 

nonconformance. As a result, instance (1) of Nonconformance 

94-01-02 was identified during this part of the inspection. 

The team 

After the team determined that no root cause evaluation had 

been performed, CENFO requested CENFM to place a hold tag on 

the lot of fuel clad tubing material. On October 20, 1994, 

during its inspection at CENFM, the team was given a root 

cause analysis report that, according to ABB CENO, was the 

final report of SSMs analysis. 

However, during a subsequent inspection of SSM on December 5 

through 7, 1994, the team was advised by SSM that its 

evaluation of the failed burst/elongation specimens was not 

complete and that it would be submitting to ABB CENO a final 

problem analysis with assigned root cause. Some confusion, 

therefore, appears to exist regarding the status of the 

analysis report given to the team on October 20, 1994. The 

team identified ABB CENOs evaluation of the failed 

burst/elongation specimens to determine assignable root 

cause as an open item of concern and requested that NRC be 

notified when ABB CENO has completed its problem analysis of 

the failed burst/elongation test specimens, identified 

possible causes, evaluated the possible causes, confirmed 

the true cause, and identified corrective actions to be 

taken. (Open Item 94-01-04) 

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18 
(2) Rod and Strip Material 

The team reviewed certain specifications for the materials 

used in manufacturing order (MO) 5403-94-1156, Supplement 1, 

to provide fuel assemblies for Southern California Edison 

Companys San Onofre Unit 2 core reload Batch K. The fuel 

assembly was prescribed in "Standard Specification for Fuel 

Assemblies," 00000-F'MDE-0100, Revision 08; supplemental 

purchasing information 1-08; and bill of materials S2K-FMDEBOM1, Revision 1. 

The team reviewed CENFOs materials specifications for Zr4 

rod material for upper and lower end caps (00000-PD-304, 

Revision 3); Zr4 strip for spacer grids (00000-PD-302, 

Revision 3); and Inconel strip for spacer grids (ASTM B 

443-84 and SPI-5R). The Zr4 rod material specification 

contained requirements for documented agreement between 

CENFO and its supplier regarding methods and standards for 

evaluating materials. However, the team found that CENFO 

had no documented agreement for the methods to be used for 

anaiyzing hydrogen, nitrogen, and oxygen gases. The team 

also found a similar condition existed for Zr4 strip 

material. 

Before the team's interim exit meetina on October 7, 1994, 

CENFO contacted its suppliers of Zr4 rod and strip haterial 

and obtained documentation correcting the condition 

described above. 

approval or review; 9392494-2 for Zr4 rod and 9394176-1 for 

Zr4 strip. 

CENFO gave the team two request for 

4.3.2.2 Grid and Cage Manufacturing 

(1) Grid Fabrication 

The team observed CENFOs laser welding operations on spacer 

grids for the San Onofre Unit 2 core reload Batch K. The 

team evaluated two laser welder qualification reports to 

verify qualification of the processes being used to laserweld the 16x16 Zr4 spacer grids. Two qualification 

procedures were required, because a second, :Longer, spacer 

grid was used to give the additional seismic stability 

required for fuel bundle assemblies supplied to San Onofre. 

The team concluded that both laser-welding qualification 

procedures had been developed according to CENFOs procedure 

and had been reviewed and approved by appropriate 

disciplines and levels of management. Process variables and 

laser motion were controlled by a computer-driven controller 

and an approved program. Applicable travelers and 

- 10 - 
procedures were present at the workstation, and the work 

observed by the team was being performed according to 

procedures. 

Qualification of the process for laser-welding Guardian, 

grids of Inconel material for the San Onofre order was in 

progress and witnessed by the team. 

The team reviewed QC-17-14, “Qualification Procedure for 

Determining the Measuring Capability of a Non-Contact 

Measuring Machine - View Bazic 12,” and determined that 

QC-17-14 had been reviewed and approved by appropriate 

disciplines and levels of management. 

intersection locations in grid spacers was performed by an 

electro-optical device referred to as the VIEW measuring 

machine. The VIEW measuring machine performed a 100% 

electro-optical measurement of the intersection locations, 

verified all critical dimensions within each spacer grid 

cell, and printed a report of the dimensions verified and 

conformance of the grid spacer to requirements. All grid 

spacers were examined by this method. The team also 

reviewed the adequacy of the calibration of the VIEW 

measuring machine and determined that standards used to 

calibrate the VIEW measuring machine were properly stored 

and calibrated and traceable to National Institute of 

Standards and Technology standards. The VIEW measuring 

machine was scheduled for calibration annually and its 

calibration was current. 

QC inspection of the 

The team also observed QC personnel visually examining the 

spacer grids and measuring critical dimensions with vernier 

calipers and gages. 

were present at the workstation, and the inspection 

activities observed by the team were being performed 

according to procedures. 

Applicable travelers and procedures 

(2) Cage Fabrication 

The team observed the fabrication of fuel assembly cages and 

their packaging for shipment to CENFM for use in fuel 

assemblies for Baltimore Gas & Electric Companys (BG&Es) 

Calvert Cliffs Unit 1. Each cage consists of five Zr4 guide 

tubes, spacer grids, the cast lower end-fitting, wear tubes, 

plugs, and flanges. The team observed the welding of guide 

tubes to spacer grids and bottom spacer grids to lower endfittings and determined the welding to be performed in 

accordance with written procedures and travelers at the work 

stations, The team determined that the welders’ 

qualifications and certifications were up to date and met 

- 11 - 

20 
procedural requirements. Assemblers were observed 

inspecting spacer grid cells with plug gages; according to 

written procedures and travelers. 

4.3.2.3 Component Manufacturing 

The inspectors observed the numerically controlled (NC) machining 

of a metal bar by a qualified machinist. The team observed that 

micrometers 140012, 340037, 340019, 70006, and 250014 were in use 

at the machining workstation and determined that these measuring 

instruments were within the calibration due dates. A subsequent 

check with the CENFO calibration technician also determined that 

the calibrations had been performed according to written 

procedures. 

In its discussions with machinist, the team learned that the NC 

machine is controlled by computer software on a tape which the 

machinist installs in the machine before starting the operation. 

The team learned that the use of computer software in the NC 

machines was not controlled by procedures. 

that this problem was identified as Finding 1 in Total Quality's 

internal audit report TQ94/MMG/006, dated July 8, 1994. 

Corrective Action Report (CAR) IQS-94-002 was issued for this 

finding with a response due date of August 8, 1994. 

The team also learned 

However, a response to CAR IQS-94-002 was not written until 

September 20, 1994. The team examined the CAR response and 

discovered that the corrective action planned would not be 

completed until late November 1994. QCP-516.1 of QCPM-500 

required the corrective action analyses and planning to be 

completed and the response returned to the originator within 30 

days. 

516.1 and respond to the CAR in the time period required was a 

nonconformance. As a result, instance (2) of Nonconformance 

94-01-02 was identified during this part of the inspection. 

The team concluded that CENFOs failure to comply with QCP4.3.2.4 Fuel Mechanical Design 

Although the Fuel Mechanical Design department belongs to the 

CENFO organization, the results of the team's evaluation of 

CENFOs mechanical design activities are described in Section 

4.3.3.5 of this report. 

4.3.2.5 BWR Operations 

The activities of CENFOs BWR Operations with respect to nuclearfuel-related activities were conducted by the following groups: 

0 

0 

0 

Core Engineering 

Safety f Transient Analysis 

Project Management 

- 12 - 

21 
In addition to evaluating QCPM-500 and QSP-lOO-NF, which CENFO 

followed to implement the QA program, the team also evaluated BWR 

Fuel Operations Quality Plan, BFO-QP-01, "Design Control for 

Analyses Performed in Support of Fuel Reloads, Licensing Reports, 

and lead fuel assembly (LFA) Programs," Revision 0, dated June 9, 

1994, in conducting its evaluation of CENFOs BWR Operations. 

The team listened to a presentation by CENFOs BWR Operations 

personnel to determine its current activity level, its 

organizational structure, and its interface and working 

relationship with ABB CENOs sister company, ABB Atom AB, 

in Vaster&, Sweden. From the presentation, the team learned 

that BWR Operations is currently working on the Cycle 12 core 

reload for Washington Public Power Supply Systems Washington 

Nuclear Unit 2 (WNP-2). During this discussion, the following 

key points were made: 

located 

e 

e 

e 

e 

e 

The CENFM facility will make the BWR UO, pellets and 

ship them to ABB Atom for use in manufacturing BWR fuel 

for the WNP-2 Cycle 12 core reload. 

In accordance with quality plans, reload analysis 

design work may be performed by either BWR Operations 

or ABB Atom, using the same computer systems and codes 

to perform analyses. 

Safety-related BWR work performed at ABB CENO will be 

reviewed by ABB Atom for technical content. 

Safety-related BWR work performed at ABB Atom will be 

reviewed by ABB Atom to the requirements of its QA 

program. 

All licensing submittals will be prepared and submitted 

by ABB CENO; however, the supporting technical analyses 

may be performed by either BWR Operations or ABB Atom. 

4.3.3 Engineering Operations 

The activities of ABB CENOs Engineering Operations with respect 

to nuclear-fuel-related activities were conducted by the 

following groups: 

Fuel Development 

Core Analysis Technology 

Systems and Transients 

Nuclear Licensing 

Total Quality 

- 13 - 

22 
Although the Fuel Hechanical Design group belongs to the CENFO 

organization, the results of the team's evaluation of this area 

are described in this Section of the report. 

Quality Control Manual, QCM-lOl-NF, Revision 1, dated 

November 15, 1993 (QCM-101-NF), used by the team in conducting 

its evaluation of Engineering Operations, contained the quality 

control procedures (QCPs) that Engineering Operations followed to 

implement the QA topical report and QAM-100-NF. 

4.3.3.1 Organization and Management 

The performance, interfaces and documentation of the reload 

safety analysis (RSA) processes were reviewed as part of the 

overall evaluation of the ABB CENO organization and management. 

The RSA is managed by an assigned reload project engineer who has 

the responsibility for producing the project deliverables and 

insuring that the RSA conforms to the QCPs specified in QCM-101- 

NF. Engineering Operations employs a matrix-like management 

scheme in which the managers of the technology groups also manage 

specified project teams that perform the RSA under the direction 

of the reload project engineers. As an example, the manager of 

the Core Analysis Technology group also manages several project 

teams that provide the Reload Analysis Reports 

of licensees. In addition, an assigned project manager provides 

the interface with the individual licensee. 

(RZILRs) for a group 

The individual component RSAs for the RARs are performed by 

engineers in the technology groups that have the responsibility 

for the specific analysis being performed (e.g., -the CEA analysis 

is performed by an engineer in the Systems and Transients 

analysis group). The computer codes and methods developers act 

as consultants to the project. 

were performed were specified in QCM-101-NF. 

The QCPs under which the RSAs 

4.3.3.2 Licensee Interface 

(1) Groundrules 

In discussions with the project managers concerning the 

organizational interfaces both within Engineering Operations 

and with the licensee, the team noted that the design and 

operational data that provided the basis and input for the 

RSA was developed iteratively with the licen., e-ee over a 

period of time. 

specifies the plant conditions and basic data that defines 

the plant analyzed and documented in the RAR., 

The resulting "groundrules" document 

The team concluded, therefore, that the groundrules data 

underlies and is fundamental to the RSA. Changes in the 

- 14 - 

23 
groundrules data can substantially change the results and 

consequences of the RSA. 

The team also evaluated the interface between Engineering 

Operations and licensees with regard to the process of 

reviews and approvals of (a) technical change requests 

(TCRs), initiated by CENFO or an external supplier, to 

document requested changes to technical requirements in a MO 

or a PO, and (b) deviation from contract requirements 

(DCRs), initiated by CENFO or an external supplier to 

request acceptance of materials, equipment, or services 

which do not meet requirements in a MO or a PO. The team 

noted that neither QCP-3.14, “Technical Change Requests,” 

nor QCP-7.4, “Deviation from Contract Requirements,” 

specifies requirements for the licensees‘ review or approval 

of these processes. Additionally, QCP-503.1, “Design 

Control,” requires, in part, that the basis for a design 

change request should be submitted to the appropriate 

project manager, who will decide whether the change has been 

justified for the contract and whether licensee concurrence 

is needed. The team noted tnat there is nCJ documented 

procedural guidance by which the project manager decides 

whether licensee concurrence is needed. However, this 

situation appears to be adequate for the current fuel 

reloads (i.e., ABB CENO-desigred fuel reloads for ABB CENOdesigned plants). The team concluded that, for future 

applications (e.g., mixed cores, BWRs, and other PWRdesigned plants), a set of standards for licensee 

concurrence would formalize the process and reduce the 

potential for errors. The team concluded that Engineering 

Operations lack of establish guidance by which the project 

manager decide the need for the licensee to concur was a 

weakness in Engineering Operations reload design process. 

In reviewing the BG&Es groundrules document, “Calvert Cliffs 

Unit 1 Cycle 12, Final Groundrules,” dated August 27, 1993, 

the team noted that the plant data was not validated. The 

team also reviewed Arizona Public Services (APSs) reload 

specification documents for its Palo Verde Unit 3 Cycle 5, 

Section 5, “Unit 3 Cycle 5 Groundrules Document,” which 

serves as a source of design input for the reload analysis 

and it also was not validated. 

documents for both Calvert Cliffs Unit 1 Cycle 12 core 

reload, and Palo Verde Unit 3 Cycle 5 core reload were 

transmitted from the licensees to their respective 

Engineering Operations project managers. 

reviewed quality plan (QP) 93-01, “Quality Plan for 

Activities Performed Under the APS/ABB Engineering 

Partnership Arrangement,” Revision 0, dated July 15, 1993, 

and found that QCP-3.3, “Design Input,” and QCP-3.6, “Design 

Interface Control,” of QCM-101-NF apply to QP 93-01. 

The reload specification 

The team also 

- 15 - 

24 
QCP-3.6 requires that (a) requested information identify the 

QA program under which the information is to be verified and 

(b) the status of verification of transmitted information be 

identified. 

data which is documented and traceable “may be assumed to be 

correct. 

In addition, QCP-3.3 provided that groundrules 

On the basis of its evaluation of the reload specification 

documents described above, the team determined that 

Engineering Operations, for the Calvert Cli.ffs Unit 1 

Cycle 12 core reload, and Palo Verde Unit 3 Cycle 5 core 

reload, had not complied with the requirements of QCP-3.6 

and identified the QA program under which the reload 

specification documents were verified. 

concluded that Engineering Operations had not complied with 

the requirements of QCP 3.6 for transmitting the information 

in a controlled and documented manner including the 

verification status. 

specification documents for both Calvert Cliffs Unit 1 Cycle 

12 and Palo Verde Unit 3 Cycle 5, the team could not find 

evidence that the design input data was documented and 

verified in a controlled manner suitable for data used as 

design input. 

personnel that they assumed the data to be correct, as 

provided for in QCP-3.3. 

The team also 

In the case of the reload 

The team was told by Engineering Operations 

The groundrules documents reviewed did not document review, 

modification, verification, approval, and control in a way 

necessary to ensure that the groundrules documents reflect 

the current, licensed condition of the plant. Although the 

responsibility to ensure the adequacy of the groundrules 

documents is shared with the licensees, Engineering 

Operations design interface activities and controls were 

based on the assumption stated in QCP-3.3 that groundrules 

data may be assumed to be correct. Therefore, Engineering 

Operations design interface and controls were not adequate 

to ensure that the information in the groundrules document 

was consistent with existing core analysis and safety 

analysis calculations for which ABB CENO is responsible. 

The team concluded that Section 3.4.3 of QCP-3.3 is not 

adequate to ensure the information in the groundrules 

document was consistent with existing core analysis and 

safety analysis calculations; the assumption that 

groundrules data is correct could produee errors in the 

reload analysis. The team also concluded that the 

instructions of QCP 3.6 were not being followed by 

Engineering Operations in its handling of the groundrules 

data and in the case of Palo Verde Unit 3 Cycle 5, the 

interface procedure QP 93-01 also was not followed by 

Engineering Operations. 

- 16 - 

25 
Recognizing that the groundrules information is critically 

important and that the use of invalid groundrules data will 

invalidate the RSA, the team concluded that the lack of 

adequate procedures to assure the validation of the reload 

design groundrules data is a nonconformance. As a result, 

instance (1) of Nonconformance 94-01-01 was identified 

during this part of the inspection. 

Strengths 

The team observed that a significant part of the Palo Verde 

Unit 3 Cycle 5 reload analysis was performed by APS nuclear 

fuel personnel. The team discussed this with Engineering 

Operations engineers to determine the impact on interface 

concerns and the quality of the work. 

examined training records for the APS personnel and 

calculation files which were prepared by APS and reviewed by 

Engineering Operations personnel. 

The team was told that for the first reload that was jointly 

analyzed by APS and Engineering Operations engineers, two 

APS engineers were in residence at ABB CENO for 6 months. 

For the next reload, APS engineers performed the work at APS 

with Engineering Operations engineers reviewing the work at 

APS or at ABB CENO. In one case, the Engineering Operations engineer performed an independent duplicate calculation and 

then used the duplicate analysis to validate the APS 

results. 

effort for the Palo Verde Unit 3 Cycle 5 reload analysis was 

a strength. 

The team also 

The team concluded that the joint engineering 

Licensee Weaknesses 

The team determined that a potential source of error in the 

engineering reload analysis process is Engineering 

Operations assumption that the groundrules data is correct. 

The team discussed the source of this data with Engineering 

Operations engineers and determined that this data must come 

from the licensee because the licensee best knows the 

reactor changes from cycle to cycle. The team also observed 

that the data contained in the output from the RSA and the 

RAR was documented and verified in an unique calculation 

folder. 

as part of the verification, and the calculation file 

contained a design analysis title page. 

incorporated by the licensee in the facility change 

evaluation. 

The output was documented and traced to its source 

The RAR will be 

The team identified certain weaknesses, as described above, 

in the licensees' design interface activities and pointed 

out that Criterion 3, "Design Control," of Appendix B to 

- 17 - 

26 
10 CFR Part 50 required, in part, that established measures 

include procedures among participating design organizations 

for the review, approval, release, distribution, and 

revision of documents involving design interfaces. 

Because the reload specification documents supplied by 

licensees contain the plant parameters that nuclear fuel 

vendors will use to perform the safety evaluations of fuel 

reloads for the licensee‘s reactor core, the licensee‘s 

reload specification documentation must reflect the current, 

licensed condition of the plant and current information and 

guidelines presenting cycle- or unit-specific information. 

To ensure the information presented in the reload 

specification documentation is consistent with anticipated 

plant configuration, operation, and licensing impact (e.g., 

plant technical specifications and applicable portions of 

the updated final safety analysis report), licensees may 

wish to review their methodology for the review, 

modification, approval, and control of reload specification 

data. 

4.3.3.3 Reload Analysis Report 

Engineering Operations performs a detailed and complete RSA for 

each reload core. The analysis was documented in the RAR that 

was given to the licensee. Core reload data and information was 

given to the licensee in the form of hard copy as well as 

computer disk. 

documented in topical reports that had been approved by the NRC 

staff. According to Engineering Operations, methods submittals 

were made when a fundamental change in methods was introduced or 

based on the nature of the uncertainties changes. Engineering 

Operations stated that these issues were discussed at periodic 

meetings with the staff. Analyses were performed for both analog 

and digital plants; the digital plants generally require a more 

complex analysis. During discussions with the team, Engineering 

Operations pointed out they were not presently providing fuel for 

cores having multiple fuel vendors (”mixed cores”). 

The RSAs were a very large and complex analyses efforts performed 

by many technical groups within the Engineering Operations 

organization. The analysis began with a three-dj-mensional. 

coupled neutronics/thermal-hydraulics core analysis performed 

with the “ROCS” simulator computer code. A series of steadystate core simulations were then performed which provide input to 

the core limits, set point, and transient analyses. These 

analyses include steady-state thermal-hydraulics, CEA ejection, 

steamline break, loss-of-coolant accident (LOCA), CEA withdrawal, 

and the determination of the “CECOR” computer code input data. 

Each of these analyses may require several weeks to perform, and 

an analysis may receive input from as many as four related 

The RSA was performed with methods that were 

- 18 - 
analyses. 

and well-coordinated schedule to allow timely input to the 

component evaluations and optimize the milestone schedule. 

licensee is given the RAR report and the results and data from 

the component analyses. 

As a result of discussions with Engineering Operations staff 

concerning methodology submittals, the team raised a question 

concerning the required validation of the nodal expansion method 

(NEM), which has recently been incorporated in the ROCS simulator 

code. The ROCS/NEM bias and uncertainties associated with this 

new approach were documented in ABB CENO topical report CE-CES129, "Methodology Manual for Physics Biases and Uncertainties," 

Revision 1-P, dated 1991. Although this report was issued in 

August 1991, the validation of the ROCS/NEM bias and 

uncertainties was not performed until after September 1992 and 

was not released until January 1993 in Revision 2-P of the 

topical report CE-CES-129. During this period, before the 

nuclear physics biases and uncertainties were validated, this 

data was transmitted to APS in the Palo Verde Unit 1 Cycle 4 

reload safety analysis (V-91-371) on December 12, 1991l. This 

unvalidated nuclear physics data was also released by ABB CENO in 

its TECHNOTE No. 92-01. The sequence of events concerning the 

release of this data is presented in Table 1 below. 

The evaluations are performed according to a detailed 

The 

The ROCS/NEM biases and uncertainties used in the Palo Verde 

Unit 1 Cycle 4 reload and included in the ABB CENOs December 12, 

1991, transmittal were not identified as unvalidated data as 

required by Section 3.5.2 of QCP-3.10, "Other Design Documents," 

of QCM-101-NF. 

during Engineering Operations reload evaluation design review and 

final verification. 

Operations release of the unvalidated ROCS/NEM nuclear physics 

data and its failure to identify the unvalidated data during its 

design review is a nonconformance. As a result, instance (2) of 

Nonconformance 94-01-01 was identified during this part of the 

inspection. 

This unvalidated data was also not identified 

The team concluded that Engineering 

'"UIC4 RAR Revision Pages," V-91-371, letter dated December 12, 1991. 

- 19 - 
Table 1, Sequence of Events Associated with 

the Release of the Unvalidated ROC/NEM Biases and Uncertainties: 

.VENT 

1. Unvalidated ROCS/NEM 

uncertainties and biases 

released in the ABB CENO 

topical report CE-CES-129, 

Revision 1-P 

2. ABB CENO transmitted the 

unvalidated nuclear physics 

data to Palo Verde Unit 1 

Cycle 4 

3. Unvalidated nuclear physics 

data released in ABB CENO 

TECHNOTE NO. 92-01 

4. ROCS/NEM nuclear physics 

data validated 

IJATE 

August 1991 

December 12, 1991 

March 9, 1992 

After September 1992 

4.3.3.4 Reload Design Analysis Process 

A primary objective of this inspection was to evaluate 

Engineering Operations reload design analysis process and 

activities. As part of this evaluation, a detailed review of 

BG&Es Calvert Cliffs Unit 1 Cycle 12 (CC1-12) reload analysis was 

performed. The CC1-12 reload design analysis did the following: 

(a) determined the core loading (enrichment: and absorber 

(b) developed the groundrules agreement, 

(c) performed the licensing (transient and set point) 

(d) performed the as-built analysis 

settings), 

analysis, 

The reload evaluations included the fuel assembly and fuel rod 

mechanical designs, determination of the fuel stored energy and 

pressure, the LOCA and thermal-hydraulic analysis, analysis of 

the core and system transients, and the determination of the 

departure from nucleate boiling and linear heat rate limiting 

safety system set points and limiting conditions for operation. 

The reload design analysis results are documented in the 

licensing report and the core operating limits report. 

After initially reviewing the approximately fifteen individual 

CC1-12 component reload evaluations, the team chose the CECOR 

coefficient library generation analysis (A-CCl-FE;-0047, 

Revision l), the CEA ejection analysis (A-CCl-FE-0030), and the 

- 20 - 
set point analysis (A-CC1-FE-0036) for detailed review because of 

the complexity of the analyses and the interfaces involved. The 

team reviewed each of these reload analyses for conformance with 

the design analysis procedures and requirements of QCP-3.2, 

”Design Analysis,” of QCM-101-NF. These analyses were sampled 

and found to fully conform with QCP-3.2. 

Paragraph (a) 1 of Section 3.2.5, “Performing and Documenting the 

Analysis,” of QCP-3.2 required that the assumptions made in the 

RSA be stated. The assumptions section of both the CEA ejection 

and the set point analysis did not identify the assumptions made 

in these analyses. Since each of these large, complex analyses 

involved many assumptions, the team concluded that these analyses 

did not satisfy the requirements of Section 3.2.5 of QCP-3.2, and 

that this failure to explicitly state the assumptions was a 

weakness in Engineering Operations documentation of the RSA. 

To determine if the quality of the reload analyses varied 

significantly from plant to plant, the team compared (a) the 

CC1-12 and the Palo Verde Unit 3 Cycle 5 (PV3-5) (A-PV3-FE0066-CT, Revision-00) CEA ejection analysis and (b) the CC1-12 

and Florida Power and Light Companys St. Lucie Unit 2 Cycle 8 

(A-SL2-FE-0035) set point analysis. These comparisons indicated 

that the quality of all these analyses was essentially the same. 

The final step in the RSA was the release of the RAR. 

Reload Report’ was issued by the project engineer and reviewed 

in accordance with Section 3.4, “Final Approval,” of QCP-3.10. 

The CC1-12 

In reviewing the Engineering Operations reload analysis process, 

the team discovered that no written procedures defined the stepby-step process for performing the RSA. 

system of using the previous cycle reload RSA as a “template” to 

define the process for the present cycle appears to be adequate 

for the current fuel reloads (i.e., ABB CENO-designed fuel 

reloads for ABB CENO-designed plants), the team concluded that, 

for future applications (e.g., mixed cores, BWRs, and other PWRdesigned plants), a set of standardized detailed reload 

procedures appeared warranted to standardize the analyses, 

formalize the process, and reduce the potential for errors. The 

team concluded that Engineering Operations lack of establish 

detailed design procedures was a weakness in Engineering 

Operations reload design process. 

Although the current 

2”Calvert Cliffs Unit 1, Cycle-12 Reload Design Report for 88 Assembly 

Reload Design,” letter CC1-FE-0050, Revision 1, dated January 28, 1994. 

- 21 - 

30 
4.3.3.5 Fuel Mechanical Design 

The reload batches typically involve changes in the fuel 

mechanical design. 

these design changes were compared to a set of fuel design 

criteria (FDC) that have been established to define an envelope 

of design acceptance. 

FDC, a design-specific evaluation was performed. The design 

criteria compliance summary (DCCS) report documented the 

evaluation of the fuel design changes relative to the design 

criteria. The fuel design engineering analysis 'was based on the 

groundrules document which specified the fuel design information 

and assumptions. 

As part of the mechanical design evaluation, 

When design parameters fall outside the 

The evaluation of the CC1-12 fuel mechanical design was performed 

by the fuel Mechanical Design group of the CENFO organization. 

The team reviewed the documentation supporting this reload 

analysis. The reload evaluation consisted of approximately 15 

individual analyses and/or tests. 

following analyses was reviewed and found to generally conform 

with the requirements of QCP-3.2 of QCM-101-NF: 

The documentation of the 

(a) evaluation of the CC1-12 core reload Batch P plenum' 

(b) 

(c) 

(d) 

spring design 

evaluation of the removal of the upper alumina spacer 

in the fuel rod design 

evaluation of the Batch N reload fuel assembly upper 

end-fitting holddown spring and bundle design for 

incore service 

power distribution transmittals from Engineering 

Operations to Fuel Mechanical Design 

The CC1-12 reload document, "14x14 TIG Welded Ziirc Grid Pitch 

Study," dated November 11, 1993, did not include any validation 

documentation. This deficiency was pointed out by the team, and 

Fuel Mechanical Design stated that the data from this document 

was not actually used in the CC1-12 reload analysis, but was 

included as supporting information. Nevertheless, the team 

concluded that the inclusion of unvalidated data in the CC1-12 

reload analysis was a weakness in the CC1-12 fuel mechanical 

design. 

The "SIGCREEP" code, version-CW1, was used to determine the guide 

tube growth and shoulder gap for the.CC1-12 reload design. The 

team reviewed the SIGCREEP code documentation, which included 

benchmarking against measured data and other associated 

documentation. The team found this documentation to be 

consistent with the requirements of QCM-101-NF. 

The team performed a detailed review of the report entitled 

"Physics Justification of TCR on 14x14 Grid Pin ]?itch," letter 

- 22 - 

31 
GM-FE-0125, dated September 29, 1993, which was included in the 

CC1-12 reload analysis. This analysis provided the neutronics 

calculations for the effect of increasing spacer grid tolerances 

beyond current fuel mechanical design criteria. The effect of 

this design change was to increase the local power peaking and 

decrease the margin to local peaking and departure from nucleate 

boiling ratio limits. 

compensates for this nonconservative local power increase by 

borrowing conservatism, which was assumed to be available in the 

rod-bowing penalty. 

The analysis described in the above report 

On the basis of this review, the team concluded that adequate 

justification had not been given for using the rod-bowing penalty 

to accommodate the increased power peaking that results from the 

CC1-12 increased spacer grid tolerances. The team concluded that 

Fuel Mechanical Designs failure to adequately justify borrowing 

conservatism constitutes a nonconformance. As a result, instance 

(3) of Nonconformance 94-01-01 was identified during this part of 

the inspection. 

4.3.3.6 Indoctrination and Training 

The qualification and training of the Engineering Operations 

personnel was included as part of the evaluation of the CC1-12 

reload desl?n process and activities. The procedure in QCP-2.1, 

"Indoctrination and Training," of QCM-101-NF specified the 

training required of the engineering staff performing the reload 

analysis. In practice, a junior-level engineer (with a B.S. or 

M.S. degree and possibly several years of experience in an 

unrelated area) will collaborate with a senior-level engineer 

(acting as a "mentor") in performing the analysis and will 

coauthor the reload analysis report. 

of experience with the mentor, the junior-level engineer may 

perform the reload analysis independently. The training records 

of the Engineering Operations staff that were involved in the 

CC1-12 reload analysis were reviewed (Forms QCF 2.1-1 2nd 

QCF 2.1-2). The training records indicated the courses attended 

and the self-study material used. However, on all of the 

(approximately 30) training records examined, no record was kept of the on-the-job training (OJT), although Section 4 of QCP-2.1 

required such records to be kept. ABB CENOs lack of documented 

OJT training was a concern to the team since OJT was the primary 

means of training the engineering staff and without records of 

this training, the staff may not be properly assigned to reload 

tasks. 

After gaining several years 

The current system of Engineering Operations indoctrination and 

training appears to be adequate for the current fuel reload 

activities (i.e., ABB CENO-designed fuel reloads for ABB CENO 

designed plants). However, since the assignment of the 

appropriate staff to the specific reload tasks can have a 

- 23 - 

32 
substantial effect on quality, the team concluded that, for 

future applications (e.g., mixed cores, BWRs, and other 

PWR-designed plants), the lack of OJT documentation was a 

weakness in Engineering Operations training process. 

4.3.3.7 Error Reports 

The team reviewed error reports, root cause and CARS and audit 

reports from audits recently performed by Total Quality, as 

internal audits, and by licensees or by licensees participating 

in the Nuclear Fuel Users Forum (NFUF) external audits. The team 

reviewed corrective actions to assess their effectiveness. When 

appropriate, the corrective actions were traced back to the 

originating error report or audit finding. 

The team observed that five input errors were reported for the 

Calvert Cliffs Unit 1 Cycle 11 core reload CECOR coefficients. 

review of the root cause and corrective actions report showed 

that the root cause of one error was that the CECOR code at ABB 

CENO (CDC version) was different than the version at BG&E (plant 

coaputer version). The intermediate/long-term corrective action 

was the development of an electronic interface between the "BASS" 

code and the CECOR code. The BASS code is used for on-line 

monitoring of thermal margin, and the BASS coefficient data was 

included in the CECOR input geometry deck for Calvert Cliffs. 

The team observed that an error report for Calvert Cliffs Unit 1 

Cycle 12 core reload reported that a CECOR input error was made. 

The team observed that the input parameter that was in error was 

also in error in the previous cycle. For Cycle 12, the 12th 

input value on card 102000 was 2.000, which was interpreted as 

200% by the BASS/CECOR codes. 

1.02, interpreted as 2%. 

A 

The correct entry should have been 

The corrective action for the input errors from Cycle 11 was to 

create a template file, which the BASS-P5 code processed to 

create the BASS coefficients for the plant. However, the BASS-P5 

code did not trap this type of input error. 

preprocessor code to the CECOR input preparation process was not 

an effective corrective action. Thus the corrective action from 

Cycle 11 did not prevent a recurrence of an input error for the 

same input parameter in Cycle 12. 

(a) the corrective action from Cycle 11 was ineffective because 

the root cause had been incorrectly identified and (b) the 

addition of a preprocessor code to the CECOR input preparation 

process was not an effective corrective action. 

concluded that Engineering Operations failure to correctly 

determine the root cause and take effective corrective action for 

the input errors that were reported for the Calvert Cliffs Unit 1 

Cycle 11 and Cycle 12 core reload CECOR coefficients constitutes 

The addition of a 

The team determined that 

The team 

- 24 - 

33 
a nonconformance. As a result, instance (3) of Nonconformance 

94-01-02 was identified during this part of the inspection. 

The team reviewed the associated documentation pertaining to 

these CECOR input problems and discussed this material with 

Engineering Operations personnel. 

validation report for BASS-P5 was also reviewed. 

errors in Cycle 11 and the Cycle 12 input error were missed by 

the design analysis and verification process, and the addition of 

a preprocessor code to the CECOR input preparation process was 

not effective. 

Operations had not established an engineering procedure or 

guideline for the CECOR input preparation. The team concluded 

that Engineering Operations failure to (a) identify the five 

input errors in Calvert Cliffs Unit 1 Cycle 11 core reload and 

the Cycle 12 during the design analysis and verification process 

and (b) establish an engineering procedure or guideline for the 

CECOR input preparation for the Calvert Cliffs Unit 1 Cycle 11 

and Cycle 12 core reload CECOR coefficients constituted a 

nonconformance. As a result, instance (4) of Nonconformance 

94 31-01 was identified during this part of the Inspection. 

The verification and 

All five input 

The team also determined that Engineering 

4.3.4 Field Services 

The activities of ABB CENOs Field Services organization with 

respect to nuclear fuel are conducted by the following groups: 

a 

a 

a 

0 

Steam Generator & Reactor Services 

NDE (AMDATA) Products & Systems 

Engineering/Products 

Total Quality 

QAM-100, "Quality Assurance Manual, Nuclear Services," 

Revision 0, dated November 12, 1994 (QAM-100) and QAM-101, 

"Quality Assurance Procedures Manual," Revision 2, (QAM-101), 

used by the team in conducting its evaluation of Field Services, 

contained the quality assurance procedures (QAPs) that Field 

Services followed to implement the QA program. 

4.3.4.1 Fuel Examination and Inspection Equipment 

In reviewing the equipment used by Field Services to perform 

inspection and repair of fuel assemblies (e.g., failed fuel rod 

detection system, sipping, visual inspection, eddy current 

examinations, dimension measurements, oxide thickness 

measurements, crud sampling, gamma scanning, fuel channel 

insgection, control rod wear inspection, and fuel assembly 

reconstitution), the team found that, for the majority of fuel 

handling tools used by Field Services, the design control 

provisions of QAM-100 and QAM-101 were not applied. The design 

and construction of these tools were neither reviewed for their 

- 25 - 

34 
failure modes and effects, nor were the tools considered under a 

foreign material exclusion program. 

While the current design practices for fuel-handling tools 

appears to be adequate for handling ABB CENO-designed fuel 

assemblies because of their robust design and construction (i.e., 

ABB CENO-designed fuel reloads for ABB CENO-designed plants), the 

team concluded that, for future applications (e.g.! mixed cores, 

BWRs, and other PWR-designed plants), a programmatic design and 

configuration control effort that considers failure modes and 

effects and that provides for foreign material exclusion may be 

warranted to reduce the potential for’ fuel-handling tool failures 

that may inadvertently cause the placement of loose parts in the 

reactor core. The team concluded that Field Services‘ lack of 

establish design and configuration control procediires was a 

weakness in the outage service activities of ABB CENO. 

4.3.4.2 Licensee-Performed Audits 

Although several licensees procure safety-related site fuel 

services from ABB CENO, only a few of the 1icer.sees that procure 

these services, in which fuel inspection and repafir tools are 

used, also perform audits of Field Services design and 

configuration control activities of these tools. 

4.4 October 17 throuph 21, 1994 - Hematite. Missouri 

During this portion of the inspection, the team evaluated CENFOs 

Nuclear Fuel Manufacturing-Hematite (CENFM) nuclear-fuel-related 

activities in Hematite, Missouri. The major groups and focused 

factories of CENFM include: 

e 

0 

e 

e 

Chemical and Ceramic Operations 

Assembly Operations 

Regulatory Compliance 

Production Support 

QCPM-500H, “Quality Control Procedures Manual, Nuc:lear Fuel 

Manufacturing-Hematite,” Revision 2, dated October 1, 1994 (QCPM500H), which CENFM followed to implement the QA program, were 

used by the team in conducting its evaluation of CENFM. 

4.4.1 Chemical and Ceramic Operations 

4.4.1.1 Powder Blending 

In April 1991, CENFM switched from producing UO, powders with the 

same enrichment as the supplied UF, to mixing UO, powders with an 

enrichment different than, but close to, that of the supplied UF, 

(known as custom enrichment blending). In the old method of 

mixing the powder, the conversion process started with a fixed 

- 26 - 

35 
UF, enrichment, and the resulting powder and pellets had the same 

enrichments as the initial enrichment of the UF,. However, with 

the new method, the enrichment of the manufactured pellets is 

different from that of the supplied UF,, and, therefore, the 

possibility existed to produce enrichment variations in powders 

that are not completely blended. Such an enrichment variation 

may occur when the process controls fail to detect a problems in 

the manufacturing process. 

By mixing powders of different enrichments, CENFM could 

manufacture pellets and fuel rods of a desired enrichment with 

shorter lead time. This process, therefore, provided greater 

flexibility for licensees to determine their final power 

requirements by being able to delay final selection of exact fuel 

enrichment in response to changes in planned outage schedules; 

this results in improved efficiency of core reload design. 

During this inspection, the team observed CENFMs response to an 

enrichment variation problem that started in September 1994. 

From its initial situation appraisal of the enrichment variation, 

accoruing to CENFM, it appeared that the process controls 

monitoring the accuracy of the enriched powder did not detect the 

off-specification enriched pellets until the rod scanner rejected 

completed fuel rods containing the off-specification pellets. 

Between September 16, 1994, and October 18, 1994, CENFMs quality 

coordinators in the Assembly Operations focused factory issued 

four problem/disposition notices (P/DNs) , identifying - 

off-specification enrichments in four different lots of powder 

and 

0 

0 

0 

0 

pellets, as follows: 

P/DN 315 was issued on September 16, 1994, on lot c635ya 

placing 1135 fuel rods on hold in matrix storage. The 

downloading of rods on hold was being accomplished per P/DN 

342, and travelers had been issued for 74 rods on lot 4280f, 

50 rods on lot 4287b, 50 rods on lot 4287c, and 46 rods on 

lot 4287d. 

P/DN 324 was issued on September 21, 1994, on lot c196sa and 

placed 1076 rods on hold. P/DNs 336, 337, 338, 339, 340, 

and 341 were issued to direct downloading of specific rods 

whose gamma profiles had been evaluated. 

P/DN 335 was issued on October 1, 1994, on lot c639sa and 

placed 510 rods on hold and directed that all rods be 

downloaded per P/DN 342 for return of the pellets to the 

ceramic factory to be recycled into the blending process. 

P/DN 357 was issued on October 18, 1994, on lot c642 and 

placed 510 rods on hold while the gamma profile on each rod 

was reviewed to determine disposition per travelers and 

P/DNs. 

- 27 - 

36 
The quality coordinators' initial investigation of the enrichment 

variation problem did not attribute the problem 'to the powder 

blending process and, without performing a complete problem 

analysis, CENFMs staff attributed the problem to an isolated 

manufacturing event. 

On October 17, 1994, over a month after the problem was initially 

observed, CARs H-1-94 and H-2-94 were issued. Because of CENFMs 

delay in generating the CARs and completing its problem analysis, 

four lots of pellets with enrichment variations were manufactured 

and loaded in fuel rods. At the time of the teain's interim exit 

meeting with CENFM on October 21, 1994, CENFM had identified and 

isolated over 500 completed fuel rods (each containing about 300 

pellets) with off-specification pellets. CENFMs final resolution 

was to reject all pellets produced from four suspected lots of 

pellet material. 

Evaluating CENFMs response to the powder blending problem, the 

team formed the following conclusions: 

(1) CENFMs P/DN 315, dated September 16, 1994, reported that 

fuel pellet isotopic deviations of varying degrees were 

identified by rod scanner profiles. However, CAR H-1-94 to 

address the corrective action for the isotopic deviations 

was not written until October 17, 1994. 

Considering the importance of a known and consistent 

enrichment to fuel operating characteristics, the team 

believes this delay in issuing a CAR is a weakness. 

(2) QCP-516.1 of QCPM-5OOH assigned responsibility to the 

cognizant quality coordinator in each of CENFMs focused 

factories for the issuance and control of CARs for traceable 

quality problems identified by P/DNs. However, QCP-516.1 

did not address the responsibilities of the cognizant 

quality coordinators in CENFMs focused factory environment 

where the corrective action analysis to determine assignable 

root causes involve interface activities between more than 

one focused factory. 

CENFMs failure to address the interface responsibilities 

between cognizant quality coordinators contributed to the 

delay in its issuance of CAR H-1-94, described in (1) above. 

As a result, instance (4) of Nonconformance 94-01-02 was 

identified during this part of the inspection. 

(3) In October 1991, CENFM identified a ma1funct:ioning 

pulse/pause timer as the assignable root cause of enrichment 

variations that resulted from non-homogeneous lots or 

batches of uranium oxide powder. CENFMs preliminary problem 

analysis of the fuel pellet isotopic deviations, described 

- 28 - 
(4) 

As 

above, had determined the assignable root cause was a 

malfunction in the powder blenders‘ pulse/pause timer; a 

reoccurrence of the 1991 event. 

Therefore, CENFMs corrective actions taken for the 1991 

event were not adequate to prevent recurrence as required by 

QCP-516.1 of QCPM-500H. As a result, instance (5) of 

Nonconformance 94-01-02 was identified during this part of 

the inspection. 

The latest revision (Revision 7, dated October 12, 1994) of 

PQ-lC, “Process Qualification Plan and Report for Enrichment 

Blending Unit Process (UPIC),” the procedure used to 

requalify the enrichment blenders, addressed (a) an increase 

in the qualified enrichment spread and (b) the sampling plan 

to requalify the blenders after CENFM recovered from the 

enrichment variation problem that caused the isotopic 

deviations described above. However, the PQ-1C did not 

provide guidance for a qualified procedure or establish 

controlled conditions to assure process qualification as 

required by QCP-509.2, “Control of Special Processes,” of 

QCPM-500H and QCH-2, “Procedure for Unit Process 

Qualification,” Revision 2, dated June 23, 1994. As a 

result, instance (1) of Nonconformance 94-01-03 was 

identified during this part of the inspection. 

of November 2, 1994, ABB CENO had neither completed its 

problem analysis’ of the enrichment variation event, as described 

above, to identify possible causes, evaluate the possible causes, 

and confirm the true cause, nor had it identified corrective 

actions to be taken. The team identified ABB CENOs evaluation of 

the enrichment variation event as an open item of concern and 

requested ABB CENO to notify the NRC when the corrective action 

has been completed. (Open Item 94-01-05) 

4.4.1.2 Pelletizing 

The team monitored the pelletizing process at CENFM. 

inspection, one press machine was out of service to perform a 

major maintenance; there were two pelletizing press machines, one 

in each production line. Normally, during return to service, 

CENFM operating staff would produce various pellets, examining their physical attributes (height, diameter, density) and observe 

their outer surface. However, there was no procedure to control 

the qualification of and the return of the equipment to full 

service. 

During the 

In addition, the team found that no procedures were established 

for increases in sampling frequency and evaluation of the pellets 

subsequent to a major modification to the pellet press. 

Normally, if there was an increase in sampling frequency, it was 

- 29 - 

38 
performed on an ir,formal basis, rather than proceduralized. The 

team was concerned that lack of a specific procedure to invoke 

increases in sampling frequency to a specific va.lue, could lead 

to inadequate sampling. 

4.4.1.3 Enrichment Verification 

The team monitored testing of the powders and pellets. A key 

device for determining accuracy of the enrichment was the 

Enrichment meter, (E-meter). This tool was an in-house detecting 

device used to check on the accuracy of the produced pellets. 

When the quality of the E-meter results were suspect, CENFM sent 

a representative sample to an outside laboratory for processing 

and evaluation. In addition, the outside laboratory was used for 

certification of each lot of pellets. 

The E-meter was used to evaluate the enrichment of the finished 

pellets. For certification of a lot of powder, sample pellets 

were sent to an outside laboratory for evaluation and 

verification of the accuracy of the enrichment. 

the external evaluation (which used pellets) was greater than the 

E-meter. Further, powder samples were taken for the E-meter 

after mixing the powder for each individual lot, whereas, pellets 

were formed after another mixing step performed batch by batch. 

This additional mixing further improved the uniformity of the 

pellets as compared with uniformity of the powder after blending. 

For this reason pellets had a more uniform enrichment (as 

determined by the external evaluation) than the powder sample 

evaluated by the E-meter. 

The accuracy of 

For the enrichment variation event described in Section 4.4.1.1 

above, a powder sample failed the E-meter enrichment criterion. 

A separate sample of pellets was sent to the outside laboratory 

for verification of the E-meter results. 

results indicated that the pellets were acceptable. 

laboratory's results invalidated the earlier E-meter evaluation 

of the powder sample. 

E-meter did not evaluate the same sample, the outside 

laboratory's identification of uniform enrichment of the pellets 

should not have been interpreted as validating the uniform 

enrichment of the lot or batch. It appears that the E-meter 

correctly identified the enrichment variation and that CENFM 

failed to correctly interpret the E-meter result:;. 

The team found the normal testing of the pellets to be 

acceptable. 

production were tested and the test equipment were properly 

calibrated. 

appropriate test criteria, it would be properly dispositioned. 

The laboratory's 

The 

Since the outside laboratory and the 

Frequent samples of powders and pe1:Lets under normal 

If a batch or a lot of powder or pe:Llets failed the 

- 30 - 
4.4.1.4 Traceability 

The team found that traceability of powders by lot numbers, 

pellets by boats (carriers), and scrap and press fines material 

met the requirements of QCPM-500H. 

batch numbers remained identifiable through the sintering 

process. Traceability was maintained by lot numbers through the 

Cardex system. Since this is a key facet of CENFMs product 

quality, various means and process check points were established 

to isolate suspected products by the smallest increments. 

Traceability of pellets by 

4.4.2 Assembly Operations 

4.4.2.1 Fuel Rod Fabrication 

The NRC inspectors observed the fabrication of fuel rods for BGCE 

and Southern California Edison. The operations were performed by 

trained operators in accordance with approved written procedures 

and travelers at work stations. Much of the process is 

controlled automatically by computer terminals vThich interface 

with rod line operators at each work station. ihe inspectors 

observed that the computer would not accept the badge number of 

persons who were untrained for the operation being performed at a 

specific work station. 

Discussions with rod line operators revealed that indications had 

been observed on the surface of Zr4 fuel clad tubing supplied by 

SSM for use by CENFM. Two of three tubes on hold at station 110 

exhibited apparent indications of pitting on the outside surface. 

Manual Advisory Notice 75, dated October 15, 1994, described the 

indications as an unknown substance. P/DN 358, dated October 18, 

1994, stated that “two fuel tubes have pitted areas on OD of 

tubes.” P/DN 358 refereed the problem to technical personnel at 

CENFO and SSM for evaluation. The tubes from SSM lot DHU51 will 

not be used because sections containing the indications will be 

removed for photographs and metallographic evaluation. 

As of November 2, 1994, ABB CENO had neither completed its 

problem analysis of the surface indications to identify possible 

causes, evaluate the possible causes, and confirm the true cause, 

nor had it identified corrective actions to be taken. The team 

identified this as an open item of concern and requested that NRC 

be notified when the corrective action has been completed. 

(Open Item 94-01-06) 

Discussions with the quality coordinator indicated that since 

July 1994 CENFM had experienced problems with SSY tubing 

involving blue, pink, black, gray, and brown stains and white 

deposits on various surfaces. There is evidence that the 

problems are being addressed as required by the CENFO QA 

programs, but the corrective action has not been successful in 

- 31 - 

40 
preventing the recurrence of new but similar indications of 

nonconforming tubing for fuel cladding. (Open Item 94-01-07) 

The rod scanner, nicknamed “Big Bertha,” using a Californium 

source, performs gamma scans of completed fuel rods as an inprocess quality check to ensure that the gamma profile of each 

fuel rod meets process control requirements. The rod scanner 

automatically rejects rods which exhibit unacceptable gamma 

profiles, and the computer control automatically issues an 

advisory notice (AN) on the nonconforming material. 

issued by the QC specialist to put the material on hold and 

provide documentation of corrective action to disposition the 

nonconforming material. 

each rod must be performed to determine whether the rod can be 

reworked or must be scrapped. 

A P/DN is 

An evaluation of the gamma profiles for 

As a result of the enrichment variation event described in 

Section 4.4.1.1 above, the team observed fuel pellets being 

unloaded from fuel rods in accordance with written procedures and 

travelers. 

manuLacturing for recycling irito the blending process, and some 

of the pellets were acceptable for loading in new fuel rods. 

4.4.2.2 Fuel Rod Prestacking 

Some of the pellets were returned to pellet 

The team observed a rod line operator performing rod prestacking 

operations. 

Batch N core reload at Calvert Cliffs. Fuel assembly C2N629 was 

prestacked according to Shop Order CC2NN6, Part AEK000. The 

prestacking plan required loading 108 fuel rods of Part FHA000 

(standard fuel rod with 4.48% enrichment) using template 25097 

and then loading 68 fuel rods of Part FHBOOO (burnable poison 

fuel rod with 4.48% enrichment) using template 25096. The team 

observed the rod line operator scan the bar codes on each fuel 

rod. The team asked the operator to scan a rod of a type 

inappropriate to the template in use and the computer gave a 

message that the rod was inappropriate. 

to enter his visitor‘s badge number to see if he could be 

authorized computer access. The computer denied access. The 

team verified that component identification was complete and 

accurate and that only properly released components were being 

used to make fuel assemblies. The team‘s review of the 

operator‘s training records showed that he was trained in 

accordance with documented procedures for prestacking and that 

his qualifications were current. 

The fuel assembly had been ordered by BG&E for the 

A team member also tried 

The team also verified that measures were established and 

implemented to control fuel assembly rod loading patterns. 

Measures had been established in QCP-505.1, “Manufacturing 

Instructions, Procedures and Drawings,” of QCPM-500H; Operating 

Sheet (OS) 3300, “Prestacking Rods for Fuel Bundle,” and OS 3320, 

- 32 - 

41 
"Rod Loading - Bundle Assembly," Revision 5, dated June 8, 1994. 

These procedures had been implemented through a computer system 

that provided operator guidance, limited operator actions to 

acceptable responses, and allowed access only to authorized 

personnel. 

The team identified the loading pattern as defined by engineering 

in Drawing E-FC2-EOOO-AO4. 

document was identified as Shop Order CC2NN6, Part AEK000, and 

was observed to be consistent with the engineering definition. 

However, the team noted that Shop Order CC2NN6, Part AEK000, did 

not show evidence of having been reviewed and approved by quality 

control personnel as required in OS 508, "Document Control," 

Revision 4, dated October 10, 1994. The team concluded that this 

was an instance of failing to follow internal procedures. 

result, instance (2) of Nonconformance 94-01-03 was identified 

during this part of the inspection. 

4.4.2.3 Cage Assembly Receiving Inspection 

The team observed an assembly operator and a QC technician 

performing receiving inspection on a cage assembly. 

operator was performing his duties in accordance with OS 3501, 

and the QC technician was performing his duties in accordance 

Traveler P/N ADYlOO was also at the workstation. 

assembly had been inspected and found acceptable, documentation 

so stating was forwarded to production control personnel for 

entry into the computer system. Having entered acceptable cage 

serial number information into the computer system, assembly 

operators were then authorized to tag the cage assembly as 

released. 

authorized for use in fuel assembly operations. 

training records of the assembly operator and the QC technician 

showed that they were trained in accordance with documented 

procedures for the inspection of cage assemblies and that their 

qualifications were current. 

4.4.2.4 Rod Pushing and Bundle Assembly Operations 

The team observed rod pushing/bundle assembly operations to 

verify conformance with requirements. OS 3320, paragraph 5.1.2, required cleaning of the pushing table at the beginning of each 

shift. On arriving, the team observed numerous loose particles 

on the pushing table and further observed that the table had not 

been cleaned for 4 days. 

The team concluded that this lack of cleanliness was an instance 

of failing to follow internal procedures. As a result, instance 

(3) of Nonconformance 94-01-03 was identified during this part of 

the inspection. 

The manufacturing and inspection 

As a 

The assembly 

with OS 3503. In addition to these operating sheets, Shop 

After the cage 

Only properly released and tagged cage assemblies were 

Review of the 

- 33 - 

42 
OS 3320, paragraph 5.2.1, required the recirculating water used 

to flush over the assembly during rod pushing to be replaced 

after approximately every 10 bundles. 

the workstation showed a recent interval of 14 bundles. The team 

concluded that this was an instance of failing to follow internal 

procedures. As a result, instance (4) of Nonconformance 94-01-03 

was identified during this part of the inspection. 

A log of water changes at 

In other respects, bundle assembly operations conformed with 

requirements. 

operators showed that they were trained in accordance with 

documented procedures for bundle assembly and that their 

qualifications were current. 

Review of the training records of the assembly 

4.4.2.5 Upper Erfd-Fitting Installation 

The team observed an assembler remove an upper end-fitting 

assembly from a wooden box and place it on the upper end of a 

fuel assembly. Four outer posts were inserted through the upper 

end-fitting assembly and threaded into the openings in the guide 

tubes. The assembler tightened the outer posts and torqued each 

one to 50 ft-lbs (+ 5 ft-lbs). The team observed calibration 

data on the torque wrench to be current. These activities were 

performed in accordance with a shop traveler for 14x14 fuel 

assembly. Installation was performed in accordance with OS 3340. 

The shop traveler and the OS were present at the workstation. 

All activities were performed in accordance with the specified 

procedures. The team observed crimping operations to secure the 

joints between the upper end-fitting and the bundle. Crimping 

activity was performed in accordance with OS 3340. Crimping unit 

pressure was observed to be 3500 psi (+ 50 psi) as required by 

procedure. Calibration of the pressure gage on the crimping unit 

was current. All assembly activity was performed according to 

the documented procedures. 

records showed that she was trained in accordance with documented 

procedures for upper end-fitting assembly and crimping and that 

her qualifications were current. 

Review of the operator's training 

- 34 - 
APPENDIX 

PERSONS CONTACTED 

The NRC staff participating in the inspections of CENFO, 

Engineering Operations, Field Services, and CENFM; ABB CENOs 

personnel contacted during both inspection periods; and the 

personnel attending the final exit meeting are listed below. A. 

bullet (0) indicates that person attended the entrance meetings 

and a dagger (t) indicates that person attended the interim exit 

meetings. 

October 3 through 7, 1994 - Windsor. Connecticut 

Total Quality, CENFO, Engineering Operations, and Field Services: 

0 

I 

t 

t 

t 

f 

t 

t 

f 

Albert, E.L. 

Allen, D.E. 

Bachman, W.H. 

Baum, J.E. 

Bennett, W. D. 

Blatter, A.J. 

Book, M.A. 

Bloomquist, G.S. 

BOSCO, R.N. 

Breckenridge, N.J. 

Brown, J.A. 

Burns, W.J. 

Case, L.R. 

Chase, M.J. 

Clifford, P.M. 

Colflesh, J.A. 

Coppersmith, W.C. 

Corsetti, L.V. 

Curtis, P.J. 

D'Amato, L.S. 

Davis, R. 

Donohue, J.P. 

Drumm, D.M. 

Ebeling-Koning, D. 

Fiero, I.B. 

Machinist 

Vice President, Total Quality 

Manager, Fuel Development 

Fuel Project Manager 

Manager of Training 

Consulting Engineer, Fuel 

Mechanical Design 

Supervisor, Set point Analysis 

Manager, Quality Assurance, Field 

Field Services Engineer 

Fuel Project Manager 

Supervisor, Neutronics Analysis 

Product Manager 

Assembler 

Records Controller 

Sr. Nuclear Engineer 

Manager, Reactor Services 

Director, Total Quality 

Supervisor of Core Materials 

Vice President, Engineering 

Quality Coordinator, Grid & Cage 

Welder 

Total Quality Engineer, Engineering 

Mechanical Design Engineer 

Manager, BWR Licensing and Safety 

Sr. Consultant 

Services 

Development 

Operations 

Factory 

Operations 

Analysis 

44 
t 

t 

t 

t 

1' 

t 

t 

t 

Fitzgerald, R.J. 

Freeburn, H.R. 

Freeman, R.S. 

French, R.N. 

Gavin, P.H. 

Glotzer, M.M. 

Harris, W. 

Johnson, G. P . Junkrans, S.B. 

Kersteen, G.C. 

Land, R.J. 

Leland, S.D. 

Matheny , R. M . 

Pati, S.R. 

Peeke, R.E. 

Perez, R. 

Rodack, T. 

Rohan, P.E. 

Rossi, J. R. 

Rotondo, P.L. 

Schon, J . 

Shapiro, N.L. 

Solury , J. J. 

Suidek, R.S. 

Toelle, S.A. 

Trapp, E.L. 

Vaslet, R.J. 

Wagner, S.G. 

Waller, K.E. 

Williamson, E.A. 

Wiencek, J. 

Wotus, D.J. 

Wyvill , J . R. 

U.S. Nuclear Regulatory Commission: 

t 

t 

Brewer, D.H. 

Carew, J.F. 

Project Manager, BWR Fuel 

Operations 

Sr. Consulting Engineer, Core 

Materials Development 

Manager, Grid & Cage Factory 

Assembler 

Core Analysis Engineer 

Consultant to ABB CENO 

Manager, Core Engineering, BWR 

Total Quality 

Vice President, CENFO 

Manager, Components Factory 

Manager, Supply Management 

Project Manager, Reload Engineering 

Project Manager, BWR Fuel 

Supervisor of Core Materials 

QC Engineer, Component Factory 

Core Analysis Engineer 

Supervisor, Mechanical & Thermal 

Hydraulics Development 

Total Quality Officer, Engineering 

Operations 

Mechanical Design :Engineer 

Manager, Fuel Mechanical Design 

Project Manager, BWR Fuel 

Operations 

Sr. Consultant 

Manager, Customer Relations 

President, ABB CENO 

Manager, Nuclear Licensing 

Director, Nuclear Fuel Projects 

Supervisor, Supplier Quality 

Supervisor, Reload Design 

Welder 

Principal Engheer 

Quality Records Coordinator 

Project Manager, (APS) Reload 

Director, BWR Operations 

Operations 

Dwelopment 

(PW 

Engineering 

Metallurgical Engineer, 

TVIS/SIB/DOTS/NRE\ 

Neutronics Specialist, Brookhaven 

National Laboratory 

45 
e 

e 

e 

e 

t 

t 

Cilimberg, R.L. 

Grow, R.L. 

Kendrick, E. D . Matthews, S.M. 

Pettis, R.L. 

Metallurgical and Fuel Specialist, 

Neutronics Specialist, 

Reactor Engineer, SRXD/DSSA/NRR 

Quality Assurance Specialist, 

TVIS/SIB/DOTS/NRR 

Sr. Reactor Engineer, 

TVIS/SIB/DOTS/NRR 

Parsmeter, Inc. 

ParGmeter, Inc. 

October 17 through 21. 1994 - Hematite. Missouri 

CENFM: 

e 

e 

e 

t 

t 

e 

e 

e 

e 

t 

t 

t 

t 

t 

t 

t 

Allen, D.E. 

Becquette, J.A. 

Bennett, W. D . Black, V.L. 

Borell, S.G. 

Brockmeyer, J.A. 

Butler, K.A. 

Coile, J. R. 

Coleman, J.L. 

Coppersmith, W.C. 

Corner, R.R. 

Fadler, R.L. 

From, R.C. 

Gazaway, T.L. 

Hess, P.M. 

Hubert, P.W. 

Junkrans, S.B. 

Kondrasiewicz, W.J. 

Luckenbach, J.W. 

Luckenbach, M.J. 

Moore, L.E. 

Nelson, H.W. 

Noacik, A.J. 

Noel, L.M. 

Page, G.J. 

Reando, W.R. 

Rode, J. A. 

Sharkey, R.W. 

Solury, J.J. 

Stokes, D.L. 

Theodoro, G.L. 

Vice President, Total Quality 

Rod Line Operator 

Manager, Training 

Trainee 

Manager, Ceramics 

Rod Line Operator 

Rod Line Operator 

QC Specialist 

Rod Line Operator 

Director, Total Quality 

Rod Line Operator 

Trainee 

Director, Quality Systems 

Quality Coordinator 

Process Engineering Specialist 

Process Engineer 

Vice President, CENFO 

QC Technician 

Quality Coordinator, Assembly 

Operations 

Supervisor, Manufacturing 

Information Systems 

Welder 

Quality Engineer 

Operations Superintendent 

Rod Line Operator 

Manager, Assembly Operations 

Warehouseman 

Consultant to ABB CENO 

Manager, Regulatory Compliance 

Manager, Customer Relations 

Manager, Production Support 

Welder 

46 
0 

0 

t 

t 

Tindall, F. J. 

Underwood, D.E. 

Uding, G.L. 

Weaver, P.E. 

U.S. Nuclear Regulatory Cornmiscion: 

0 

0 

0 

0 

t 

t 

t 

t 

t 

Brewer, D.H. 

Cilimberg, R.L. 

Cwalina, G.C. 

Matthews, S.M. 

Salehi, K. 

Rod Scanner Operator 

Manufacturing Systems Engineer 

Quality Coordinator 

Manager, Material Scheduling and 

Accountability 

Metallurgical Engineer, 

Metallurgical and Fuel Specialist, 

Section Chief, TVIS/SIB/DOTS/NRR 

Quality Assurance Specialist, 

Reactor Engineer, SRXD/DSSA/NRR 

TVIS/SIB/DOTS/NRR 

Parsmeter, Inc. 

TVIS/SIB/DOTS/NRR 

November 2. 1994 - Windsor. Connecticut 

ABB CENO - Final Exit Meeting: 

Allen, D.E. 

Bachman, W.H. 

Colf lesh, J. A. 

Coppersmith, W.C. 

Curtis, P.J. 

Junkrans, S.B. 

Kersteen, G.C. 

Land, R.J. 

Rohan, P.E. 

Solury , J . J. 

Suidek, R.S. 

Trapp, E.L. 

Wyvill , J . R. 

Vice President, Total Quality 

Manager, Fuel Development 

Manager, Reactor Services 

Director, Total Quality 

Vice President, Engineering 

Vice President, CENFO 

Components Factory Manager 

Manager, Supply Management 

Total Quality Officer, Engineering 

Manager, Customer Relations 

President, ABB CENO 

Director, Nuclear Fuel Projects (PWR) 

Director, BWR Fuel Operations 

Operations 

Operations 

U.S. Nuclear Regulatory Commission: 

Brewer, D.H. 

Matthews, S.M. 

Metallurgical Engineer, TVIS/SIB/DOTS/NRR 

Quality Assurance Specialist, 

TVIS/SIB/DOTS/NRR 

47 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASH IN GTO N, D. C. 205550001 

December 5, 1994 

Mr. Joseph M. Tate 

General Manager 

ABB Service Company 

Regional Service Center 

5311 Commerce Parkway West 

Cleveland, OH 44130 

SUBJECT: NRC INSPECTION NO. 99901281/94-01 

Dear Mr. Tate 

This letter refers to the inspection conducted by Mr. K. R. Naidu of this 

office on September 22-23, 1994. 

activities conducted at your ABB Service Center facility (ABB) in Cleveland, 

Ohio. At the conclusion of the inspection, the findings were discussed with 

you ana the members of your staff identified in the enclosed report. 

The insrection included a review of 

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

enclosed report. 

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

Federal Requlations (10 CFR Part 21) for reporting of defects and 

noncompliance, and Appendix B to 10 CFR Part 50. 

selected tests being performed on two 480 Volt circuit breakers from Cleveland 

Electric Illuminating Company's (CEI) Perry Nuclear Power Plant after being 

overhauled at your facility. The inspector also reviewed the actions taken by 

your staff to correct adverse findings identified in an audit that CEI had 

conducted. Within these areas, the inspection consisted of an examination of 

procedures and representative records, interviews with personnel, and 

observations by the inspector. 

The inspector evaluated the program that ABB establ i shed and 

The inspector also observed 

Based on the results of this inspection, certain of your activities appeared 

to be in violation of NRC requirements, as specified in the enclosed Notice of 

Violation (Notice). 

maintenance problems involving hardened grease in circuit breakers or bypassed 

disconnect switches in 5HK type circuit breakers. The violation is of concern 

because it is not possible to determine from records if similar defects or 

deviations had been identified which may generically affect the operation of 

safety-re1 ated circuit breakers instal 1 ed at other nuclear power plants. 

Specifically, you did not document an evaluation of 

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

specified in the enclosed Notice when preparing your response. 

response, you should document the specific actions taken and any additional 

actions you plan to prevent recurrence. 

In your 

48 
Mr. Joseph M. Tate -2- 

In addition, during this inspection the NRC determined that the implementation 

of your QA program failed to meet certain NRC requirements. Specifically, the 

inspector ident i fi ed two exampl es of inadequate procedures which are contrary 

to Criterion V of Appendix B to 10 CFR 50. The specific findings and 

references to the pertinent requirements are identified in the enclosed Notice 

of Nonconformance. 

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

statement in accordance with the instructions specified in the enclosed Notice 

of Nonconformance. 

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

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

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

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

by the Paperwork Reduction Act of 1980, Pub. L. No. 96.511. 

Si ncerel y , 

Speci a1 Inspection Branch 

Division of Technical Support 

Office of Nucl ear Reactor Regul at i on 

Docket No.: 99901281 

Enclosures: 1. Notice of Violation 

2. Notice of Nonconformance 

3. Inspection Report 99901281/94-01 

49 
ENCLOSURE 1 

NOTICE OF VIOLATION 

ABB Service Company 

C1 eve1 and, Ohio 

Docket No. 99901281 

Report No. 94 01 

During an NRC Inspection conducted on September 22-23, 1994, a violation of 

NRC requirements was identified. 

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

(1992), the violation is listed below: 

In accordance with the "General Statement of 

Section 21.51, "Maintenance and inspection of records" of Title 10 of the Code 

of Federal Requlations (10 CFR Part 21) requires (a) each individual, 

corporation, partnership, or other entity subject to the regulations in 10 CFR 

Part 21 to prepare and maintain records Recessary to accomplish the purposes 

of 10 CFR Part 21, specifically (1) to retain evaluations of all deviations 

and failures to comply for a minimum of five years after the date of 

eval uati on. 

Contrary to Section 21.51, as of September 23, 1994, ABB did not document its 

evaluations of deviations and failures to comply and maintain them to allow 

representatives of the commission to inspect those records to ensure 

compl i ance. 

potential deviations regarding hardened grease in circuit breakers and 

bypassing of the "On-Off" charging motor. switch in 5HK type circuit breakers. 

Specifically, there were no records of eval uati ons for two 

(94-01-04) 

Pursuant to the provisions of 10 CFR 2.201, ABB Service Company is hereby 

required to submit a written statement of explanation to the U.S. Nuclear 

Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555 

with a copy to the Chief, Special Inspection Branch, Division of Technical 

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

the letter transmitting this Notice of Violation. 

clearly marked as a "Reply to a Notice of Violation" and should include for 

each violation: (1) the reason for the violation, or, if contested, the basis 

for disputing the violation, (2) the corrective steps that have been taken and 

the results achieved, (3) the corrective steps that will be taken to avoid 

further violations, and (4) the date when full compliance will be achieved. 

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

response time. Under the authority of Section 182 of the Act, 42 U.S.C. 2232, 

this response shall be submitted under oath or affirmation. 

This reply should be 

Dated at Rockvi 11 e, Mary1 and 

this 5 day of December, 1994 

50 
ENCLOSURE 2 

NOTICE OF NONCONFORMANCE 

AB6 Service Company 

Cleveland, Ohio 

Docket No. 99901281 

Report No. 94 01 

Based on the results of a U.S. Nuclear Regulatory Commission (NRC) inspection 

conducted at ABB Service Company (ABB) on September 22-23, 1994, it appears 

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

requirements as described below. 

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

Part 50 of Title 10 of the Code of Federal Resulations (10 CFR Part 50, 

Appendix B), requires, in part, that activities affecting quality be 

prescribed by documented instructions , procedures and draw.ings , of a type 

appropriate to the circumstances. The instructions, procedures, or drawings 

shall include appropriate qualitative and quantitative acceptance c r i t e r i a f o r 

determining that important activities have been satisfactorily accomplished. 

Contrary to the above, the attachment to Revision 2 of ABB Procedure 15.1, 

"Control of Nonconforming Items" dated May 20, 1991, did not have appropriate 

qualitative criteria to document evaluations for determining if the defect or 

deficiency identified in the nonconformance needs a 10 CFR Part 21 report to 

be filed. (99901281/94-01-03A) 

Contrary to the above, the ABB test instruction used for testing low voltage 

metal-cl ad circuit breakers for safety-re1 ated applications did not specify 

the maximum acceptable contact resistance value, or a1 ternatively, the maximum 

acceptable voltage drop at full load for the millivolt drop test on metal-clad 

low voltage circuit breakers. (99901281/94-01-03B) 

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

Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555, 

with a copy to the Chief, Special Inspection Branch, Division of Technical 

Support, Office of Nuclear Reactor Regulatim, within 30 days of the date of 

the letter transmitting this Notice of Nonconformance. This reply should be 

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

for each nonconformance: (1) a description of steps that have been or will be 

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

will be taken to prevent recurrence; and (3) the dates your corrective actions 

and preventive measures were or will be completed. 

Dated at Rockvi 11 e, Mary1 and 

This 5 day of December, 1994. 
INSPECTION REPORT 

ORGANIZATION: 

REPORT NO.: 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY: 

CORRESPONDENCE 

ADDRESS : 

INSPECTION CONDUCTED: 

INSPECTOR: 

APPROVED : 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF TECHNICAL SUPPORT 

ABB Service Company 

C1 evel and , Ohio 

99901281/94-01 

Mr. E. Link, Quality Assurance Manager 

(216) 267 2882 

ABB Service Company services swi tchgear 

manufactured by I.T.E/BBC 

ABB Service Company 

Regional Service Company 

5311 Commerce Parkway West 

C1 evel and, Ohio 44130 

September 22-23, 1994, 

A 

QLihA?.*L /2/(/4y 

I 

Kamalakar R. Naidu Date 

Vendor Inspection Section 

Special Inspection Branch 

Division of Technical Support 

Office of Nucl ear Reactor Regul ati on 

c. 

wal ina, Chief Date 

ction Section 

Speci a1 Inspecti on Branch 

Division of Technical Support 

Office of Nuclear Reactor Regul at! on 

INSPECTION BASES: 10 CFR Part 50, Appendix B; 10 CFR Part 21 

INSPECTION SCOPE: Tests on K-Line breakers, actions to correct 

audit findings, 10 CFR Part 21 evaluations, and 

implementation of 10 CFR 50, Appendix B in 

selected areas. 

PLANT SITE APPLICABILITY: Perry Nuclear Power Plant, and other plants with 

I.T.E/Gould/Brown Boveri 480 Volt (V), 4160 V, 

and 15kV swi tchgear. 

1 

52 
1. INSPECTION SUMMARY 

ABB Service Company 1 ocated in C1 evel and, Ohio performed overhaul i ng services 

on electrical circuit breakers installed at Cleveland Electric Illuminating 

Company's (CEI ' s) Perry Nucl ear Power P1 ant (PNPP) . The circuit breakers were 

originally designed and manufactured by I.T.E, which subsequently changed 

ownerships and became known as 1.T.E-Imperial , 1.T.E-Gould, Gould-Brown 

Boveri, Brown Boveri Electric and finally Asea Brown Boveri (ABB). ABB 

currently manufactures metal-clad low voltage and medium voltage circuit 

breakers at Florence, South Carolina. ABB has established service centers to 

repair and service switchgear at several locations in the U.S. including C1 evel and. 

acceptance tests on two K-Line low voltage metal-clad circuit breakers that 

had been refurbished, reviewed actions taken by ABB to correct audit findings, 

and evaluated the program that ABB established and executed to implement the 

provisions of 10 CFR 50, Appendix By and 10 CFR Part 21 in selected areas. 

On September 22-23, 1994, the NRC inspector witnessed selected 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

Th:p is the first NRC inspection of this facility. 

3 INSPECTION DETAILS 

3.1 Entrance and Exit Meetinas 

During the entrance meeting on September 22, 1994, the NRC inspector discussed 

the scope of the inspection and the areas to be reviewed arid established the 

persons to contact within ABB Service Company management arid staff. 

the exit meeting on September 23, 1994, the NRC inspector discussed his 

findings and concerns with the ABB Service Company staff identified in 

Section 4. 

During 

3.2 Review of ABB Quality Assurance (QA) Manual 

The inspector reviewed the ABB QA manual, Revision 4, dated January 22, 1993, 

which outlines the program to be adopted by the various ABE service facilities 

in the U.S. The ABB Cleveland facility developed QA procedures to implement 

the corporate QA manual. The inspector reviewed the QA manual and the 

implementing procedures and determined the following: 

A current organization chart for the Cleveland facility was not 

available. The corporate QA manager said that one was being prepared. 

Pending review of the organization chart, this matter is considered an 

unresolved item. (94-01-01) 

The job descriptions and qualifications of persons performing safetyrelated activities were not available. The Corporate QA manager said 

that this documentation is being prepared. Pending review of the job 

descriptions and qualifications of personnel , this matter is considered 

an unresolved item. (94-01-02) 

2 

53 
Qual i ty Assurance Procedure (QAP) 15.1, "Control of Nonconforming 

Items," Revision 2, dated December 18, 1991, states, in part, "Any 

significant defect in the Nuclear safety-related equipment shall be 

reported to the Corporate QA Manager in order for an evaluation to be 

made if a 10 CFR Part 21 Report needs to be filed." The procedure did 

not elaborate how this provision was to be implemented. For instance, 

the procedure did not attach a checklist or a form in which personnel 

could document the significant defect observed and forward it to 

designated individuals within the organization for evaluating the 

generic applicability. 

evaluate a nonconforming condition is considered contrary to 10 CFR 50, 

Appendix B, Criterion V. (94-01-03A) 

Lack of qualitative acceptance criteria to 

3.3 Review of Actions Taken on a Part 21 Reoort 

In a letter dated March 3, 1989, Indiana Michigan Power Company informed the 

NRC pursuant to 10 CFR Part 21 that their personnel observed failures of 

closing mechanisms in two BBC Brown Boveri , Inc. type 5HK250 circuit breakers 

installed in its D.C. Cook Nuclear Plant Unit 1. The cause of failure was 

determinc:! to be aging and dirt contamination. 

the closing mechanism, the breaker functioned properly. 

After cyedning and lubricating 

The plant personnel had noticed that on February 27, 1989, a similar breaker 

had failed close. After cleaning and lubricating the closing mechanism, the 

breaker functioned properly. 

the breakers, plant personnel determined that the breaker manufacturer's 

instructions contained in 1.T.E Instruction No. 1B-8.2.7-2 pertaining to the 

lubrication of ABB circuit breakers were unclear. Specifically, the problem 

related to the following lubrication instruction contained in the instruction 

book for 5HK250 circuit breakers: 

During their investigations of the failure of 

All other mechanism parts, bearings, pins, etc. have been 

7 ubri cated with ANDEROL L757 manufactured by Tenneco Chemical, 

Inc. Intermediate Division. The circuit breaker requires no 

lubrication during its normal service life. However, if the 

grease should become contaminated or if parts are replaced, any lubrication should be done with NO-OX-ID or ANDEROL grease as 

appl i cab1 e. 

In a letter to the NRC dated March 21, 1989, ABB Power Distribution, Inc., 

responded to the 10 CFR Part 21 Report filed by the D.C. Cook Nuclear Power 

Plant concerning the contaminated lubricant on the operating mechanisms of two 

5HK250 ABB circuit breakers resulting in their failure to close on command. 

In this letter, ABB stated that breakers should be periodically checked for 

cleanliness and if the lubricant is found to be contaminated and dry (hardened 

and discolored), the grease should be removed with an approved solvent and 

relubricated. 

has to be disassembled. 

breaker has to be completely disassembled to observe if the grease had been 

contaminated. 

ABB stated that to perform this activity, -the circuit breaker 

The ABB letter did not expressly state that the 

3 

54 
In a letter to the NRC dated October 3, 1994, ABB provided the following 

cl ari f i cation: 

In performing refurbishing activities on circuit breakers during 

the past few months, ABB has found that the condition of the 

lubricant (i .e. the extent of hardening and contamination) cannot 

be easily determined without complete disassembly of the breaker 

mechanism. 

components requiring lubrication are not visible or a.ccessible for 

inspection without completely disassembling the mechanism of the 

breaker. 

This is due to the fact that some of the mechanism 

Even though it is not clear from the ABB letter, the ABB personnel confirmed 

that ANDEROL type grease should be used on the mechanisms. 

grease should be used on the primary and secondary electrical mating contacts, 

i.e., the contacts on the circuit breaker which mate with the stationary 

contacts in the breaker cubicle. The bridge pivot points should also be 

lubricated with NO-OX-ID type grease. 

contacts should not be lubricated with any type of grease under any 

ci rcumstances. The inspector enquired i f ABB personnel advised PNPP to clean 

and grease the contacts in the stationary cubicle and found that they had not 

done so in the past, but would advise them in the future. 

NO-OX-ID type 

'The main and interrupting electrical 

ABB did not have any documents indicating that it evaluated this condition 

pursuant to 10 CFR 50 Part 21 to determine if such conditions were prevalent 

in ABB breakers installed at other operating plants on which they perform 

service at other locations. ABB's failure to document its evaluation of this 

potential deviation is considered a violation of 10 CFR part 21.51. (94-01-04) 

3.4 Evaluation of a Potential Part 21 Item 

The inspector reviewed the as-found conditions documented by ABB service 

personnel during the 10-year refurbishment of PNPP circuit breakers and 

observed that PNPP personnel had removed the motor disconnect switch on a 5HK 

350 MVA, 1200 ampere circuit breaker, Serial No. 51958B-319064, bo1 ted 

together the connections, and applied heat shrink tubing to the joint. ABB 

personnel noticed that perhaps due to non-uniform heat, the insulation on the 

conductor underneath the heat shrink tubing had melted and the bare conductors 

had been exposed. 

The purpose of the "On-Off" motor disconnect switch is to control power to the 

motor which charges the closing strings of the breaker mechanism. 

closing springs are charged, the breaker becomes ready to close on demand. 

With the motor disconnect switch jn the "Off" position, the charging motor 

does not charge the closing springs. 

placed in the ''Off" position before the breaker is racked into the test 

position or into the cubicle. 

the breaker is in either of the positions. However, when the circuit breaker 

is racked into the "test" position, with the switch bypassed, auxiliary power 

is made available to the motor as soon as the secondary disconnects of the 

breaker and cell connect. The motor will start and charge the closing 

springs, enabling the breaker to be closed if any of the interlocks 

When the 

For operational safety, the switch is 

The switch is placed in the "On" posiiion after 

4 

55 
inadvertently fail while the breaker is being racked into the "connect" 

position. 

the breaker meet the live contacts in the cell, endangering the operating 

personnel and possibly damaging the switchgear. ABB had no documents 

indicating that this condition was evaluated pursuant to the requirements of 

10 CFR Part 21.51. The inspector observed that by removing and bypassing the 

motor disconnect switch, PNPP could not only endanger the safety of the 

personnel operating the swi tchgear, but a1 so the original 1E qual if i cati on of 

the circuit breaker. The ABB QA manager informed the inspector that he would 

document this matter. ABB's failure to document its evaluation of this 

deviation is considered another example of a violation of 10 CFR part 21.51. 

This may result in a severe flash-over when the primary contacts of 

(94-01-04) 

3.5 Review of Problems With Inadeauate Service 

During March 10-April 21 and April 22-May 27, 1994, inspectors from NRC Region 

I11 conducted inspections at PNPP and documented their findings in Inspection 

Reports (IRs) 50-440/94006 (DRP) and 50-440/94009 (DRP) respectively. IR 50- 

440/94006(DRP) stated that, in early 1993, PNPP sent one 4.1 kV and two 480 V 

circuit breakers to ABBY Cleveland, for refurbishment. ABB reported to PNPP 

that the "grease hardening process had begun" in the breakers. In March 1993, 

PNPP sent four 13.8 kV breakers for refurbishment. ABB reported that the noncritical parts of the breaker mechanism had indications of hardened grease; 

however, the critical portions of the breaker assemblies had sufficient 

lubrication. 

breaker maintenance and were concerned about the operability of the breakers. 

PNPP initiated Condition Report (CR) 94-511 to correct this matter. 

stated that PNPP issued purchase order (PO) No. 132717 to ABB to refurbish 

five breakers. When ABB returned the refurbished breakers, PNPP did not 

remove the arc chutes and observe whether ABB lubricated the internal 

mechanisms and, therefore, did not detect that ABB had not lubricated them. 

On April 18, 1994, NRC Region I11 inspectors accompanied PNPP personnel during 

a PNPP audit of the ABBY Cleveland service shop. PNPP personnel identified 

deficiencies in the imp1 ementati on of the qual i ty assurance program 

established by ABB and documented adverse findings in the following four 

"Vendor Action Requests" (VARs) . 

The NRC Region I11 inspectors reviewed the PNPP reports on the 

CR 94-511 

VAR 1 identified that ABB had not satisfied the requirements of CEI's 

PO # 132717 because the requirements to perform general lubrication inspection 

and overhaul were not forwarded to the shop. 

procedure to define the method of translating and transmitting customer 

requirements into work documents. 

revise Procedure QAP 4.4, "Sales Order Processing," to provide direction 

regarding the method of trans1 ating customer requirements from PO into 

internal ABB service work documents. The inspector reviewed QAP 4.4, 

Revi si on 1 , and determined that Paragraph 3.2.2 requires resol ving any 

inconsistencies between the purchase order requirements defining the scope of 

work and the ABB quote; any inconsistencies must be resolved before commencing 

work. 

CEI required ABB to establish a 

Corrective action proposed by ABB was to 

5 

56 
VAR 2 identified that there was no objective evidence to verify that persons 

performing qual ity-related activities had received indoctrination and 

training. ABB proposed to review existing records and compile the records in 

a central location. The ABB quality assurance manager informed the inspector that the documents related to the indoctrination and training had been 

retained by different managers in the past. This has now been corrected and 

now all the documents are being maintained in a central location. 

VAR 3 identified that ABB used checklists that had not been reviewed and 

approved. ABB proposed to revise Procedure QAP 10.1 to require review and 

approval of checklists. The inspector reviewed Revision 1 to QAP 10.1 and 

determined that paragraph 3.4 now states, "Inspection shall be performed with 

checkl ists, drawings, instruction books or specifications as applicable. 

Locally prepared checklists (i .e., blank forms) shall be assigned revision 

numbers and/or issue dates. The service center QA manager shall review and 

approve such locally prepared checkl ists/forms, as well as ,any revisions, 

prior to use. Approval for use shall be indicated by his/her signature and 

date on a local document review/approval sheet which will be maintained on 

file at the service center." 

VAR 4 identified that ABB failed to document corrective action taken to 

prevent recurrence in response to nonconforming conditions identified by CEI 

personnel on their five safety-related breakers. ABB issued a notice 

providing clarifications to the definitions used for service. 

No adverse findings were identified in this area. 

3.6 Observation of AcceDtance Tests 

The inspector observed ABB test personnel perform final acceptance tests on 

PNPP K-6005 circuit breakers Serial Nos. 812545A-010188 and 51817D-22-1135 in 

the presence of an PNPP engineer. 

documented in a checklist that had been prepared and approved. The test 

equipment being used to perform the acceptance test contained stickers which 

indicated that the calibration was current. 

The results of the tests were being 

The test results were acceptable. 

The inspector observed that ABB did not specify the maximum acceptable contact 

resistance value, or alternatively, the maximum acceptable voltage drop at 

full load for the millivolt drop test value on low voltage metal-clad 

breakers. The inspector informed ABB personnel that the test procedure is 

incomplete and contrary to Criterion V, of 10 CFR 50, Appendix B because it 

did not contain acceptance criteria to assure that all important activities 

have been accompl i shed. (94-01-03B) 

6 
4. PERSONS CONTACTED 

ABB Service ComDany 

* 

* + J M. Tate, General Manager, North Central Region 

* + J 0. Webb, Quality Manager * + E. Link, Manager, Cleveland Service 

M. Rice , Product Development Manager 

T N. Kelly, Vice President, Total Quality/Supply Management 

* Individuals present at the entrance meeting on September 22, 1994. 

+ Individuals present at the exit meeting on September 23, 1994. 

7 

58 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-0001 

December 30, 1994 

Mr. Scott Akers Jr., Chief Executive Officer 

Cardinal Industrial Products, 

Limited Partnership 

3873 West Oquendo 

Las Vegas, Nevada 89118 

SUBJECT: NRC INSPECTION NO. 99901076/94-01 

Dear Mr. Akers: 

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

of your facility at Las Vegas, Nevada-, conducted by Messrs. U. Potapovs and 

L. L. Campbell of this office November 29 through December 1, 1994, and the 

discus~;~ns of their findings with members of your staff at the conclusion of 

the in3pection. 

assurance program and its implementation in selected areas such as (1) control 

of purchased material and services, (2) material and traceability control , (3) supplier audits, (4) commercial grade item dedication, and (5) a review of 

your program for implementing Part 21, “Reporting Defects and Noncompl iance,” 

of Title 10 of the Code of Federal Requlations (10 CFR). 

The inspection was conducted to evaluate Cardinal’s quality 

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

enclosed inspection report. 

procedures and representative records, discussion and interviews with 

personnel , and observations by the inspectors. 

This inspection consisted of an examination of 

While our review of your procurement and commercial grade item dedication 

activities indicated that, in several areas, your quality assurance program 

had established good controls for assuring product compl iavlce to applicable 

specifications, Lhe inspectors identified two areas where your program failed 

to meet certain NRC requirements. The first area relates ‘to the selection of 

an adequate sampling plan for verification of critical characteristics. 

Specifically, your program does not contain a basis or objective evidence to 

show that the destructive sampling plan for verifying critical characteristics 

provides reasonable assurance that the dedicated items meet the applicable 

procurement document requirements. The second area concerns the adequacy of 

measures for evaluation and disposition of nonconforming conditions identified 

during the ASME Code material upgrade and commercial grade item dedication 

process. In at least one instance Cardinal’s mechanical testing of material 

from a qualified supplier invalidated the supplier’s certification, but no 

evaluatia was conducted to determine the root cause of this situation and 

Cardinal continued to supply materi a1 for appl i cat i ons where at 1 east part i a1 

reliance was placed on the supplier’s certification. 

and references to the pertinent requirements are identified in the enclosures 

of this letter. 

The specific findings 

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

statement in accordance with the instructions specified in the enclosed Notice 

59 
Mr. Scott Akers Jr. -2- 

of Nonconformance. 

show good cause for us to do so. 

Me will consider extending the response time if you can 

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

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

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

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

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

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

discuss them with you. 

Sincerely, 

Robert M. Gallo, Chief 

Special Inspection Branch 

Division of Technical Support 

Off i ce of Nucl ear Reactor Regul at i on 

Docket No. 99901076 

Encl osures : 

1. Notice of Nonconformance 

2. Inspection Report 99901076/94-01 

60 
Enclosure 1 

NOTICE OF NONCONFORMANCE 

Cardinal Industri a1 Products 

Limited Partnership 

Las Vegas, Nevada 

Docket No. : 99901076/94-01 

Based on the results of an NRC inspection conducted on November 29 through 

December 1, 1994, it appears that certain of your activities were not 

conducted in accordance with NRC requirements. 

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

of Appendix B to Title 10 of the Code of Federal Requlations (10 CFR) 

Part 50, requires, in part, that measures shall be established to assure 

that purchased materi a1 conforms to procurement documents. 

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

to substantiate that its destructive testing sampling plan for verifying 

critical characteristics provides reasonable assurance that dedicated 

commerci a1 grade i tems (CGIs) suppl ied met appl i cab1 e procurement document requirements. (Nonconformance 99901076/94-01-03) 

B. Criterion XV, "Nonconforming Materials, Parts, or Components, I' of 

Appendix B to Title 10 of the Code of Federal Requlations (10 CFR) Part 

50, requires , in part, that measures shall be established to assure 

that nonconforming items are reviewed and accepted, rejected, repaired, 

or reworked in accordance with documented procedures. 

Contrary to the above, Cardinal had not implemented measures for the 

review, evaluation, and disposition of material accepted based, in part, 

on the supplier's certification when Cardinal's tests on this material 

show this certification to be invalid. (In the example reviewed, the 

manufacturer's certified impact test results were found to be 

incorrect). (Nonconformance 99901076/94-01-04) 

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

Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555, 

with a copy to the Chief, Special Inspection Branch, Division of Technical 

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

the letter transmitting this Notice of Nonconformance. This reply should be 

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

for each nonconformance: (1) 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 30th day of December, 1994 
Enclosure 2 

ORGANIZATION: 

REPORT NO. : 

CORRESPONDENCE 

ADDRESS : 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY: 

U. S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF TECHNICAL SUPPORT 

VENDOR INSPECTION SECTION INSPECTION REPORT 

INSPECT ION CONDUCTED : 

INSPECTOR: 

OTHER INSPECTORS: 

APPROVED : 

INSPECTION BASIS: 

INSPECTION SCOPE: 

PLANT SITE 

APPLICABILITY: 

Cardinal Industri a1 Products 

Limited Partnership 

99901076/94-01 

Mr. Scott Akers Jr., Chief Executive Officer 

Cardinal Industri a1 Products 

Limited Partnership 

3873 West Oquendo 

Las Vegas, Nevada 89118-3098 

Mr. David 2. Hathcock, Vice President 

Qual i ty Assurance 

Supplies fastener products and other metal1 ic 

materials for use at nuclear power plants 

h December 1, 1994 

12. - 2.5 - 94 

U1 di s Potapovs , Senior Engineer Date 

Vendor Inspection Section 

Speci a1 Inspecti on Branch 

Larry L. Campbell, Reactor Engineer 

/J7/L9/V 

GregorgCfCwal ina, Chief Date 

Vendor InYpect i on Sect i on 

Special Inspection Branch 

10 CFR Part 21 and Appendix B to 10 CFR Part 50 

To review and evaluate Cardinal Industrial Products 

(Cardinal ’ s) qual i ty assurance program and its 

implementation in selected areas such as (I) control 

of purchased material and services, (2) material and 

traceabi 1 i ty control , and (4) commerci a1 grade i tern 

dedication. 

All plants using fastener products and other material 

suppl ied by Cardinal . 

62 
1 INSPECT I ON SUMMARY 

Inspector Foll ow Up Items 99901076/94-01-01 and 99901076/94-01-02 were 

identified and are discussed in Sections 3.4.4.6 and 3.4.5.3 of this report. 

Nonconformances 99901076/94-01-03 and 99901076/94-01-04 were identified and 

are discussed in Sections 3.4.6 and 3.6 of this report. 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

This was the first inspection at Cardinal Industrial Products Limited 

Partnership (Cardinal). 

3 INSPECTION FINDINGS AND OBSERVATIONS 

3.1 En+-ance and Exit Meetinss 

In the entrance meeting on November 29, 1994, the NRC inspectors discussed the 

scope of the inspection and established interfaces with Cardinal management. 

During the exit meeting on December 1, 1994, the NRC inspectors discussed 

their findings and observations with Cardinal management and staff. 

3.2 Historical Summarv and Description of Facilities 

According to Cardinal management, Cardinal was established in May 1993, when 

Scott Akers, Jr., the current CEO and owner, purchased the company from Hub, 

Inc. 

organization (Cardinal Industri a1 Products Corporation {CIPC}) had declared 

bankruptcy. 

Industri a1 Products , Inc. The 1 ocati on of the manufacturing and warehouse 

facilities has not changed as a result of these transactions. 

Hub had previously acquired the company in 1987 after the previous 

During Hub's ownership, the company operated as Cardinal 

The principal items manufactured on site are ferrous fastener products (nuts, 

bo1 ts, threaded rod) however, Cardinal s scope of supply a1 so includes ferrous 

and nonferrous bars, castings, forgings, plates, flanges, tubular products, 

structural shapes, and similar items. Cardinal has been accredited by the 

American Society of Mechanical Engineers (ASME) as a materi ill manufacturer and 

a material supplier and holds current Quality System Certificates (QSC) in 

these areas. 

Cardinal has in-house destructive and nondestructive testing capabilities 

including chemical analyses, tensile, hardness and impact testing as well as 

magnetic particle, 1 iquid penetrant, and ultrasonic examination capabilities. 

Dimensional inspections, including thread measurement are also performed at 

the plant. 

2 

63 
3.3 10 CFR Part 21 PROGRAM 

3.3.1 Procedures and Imp1 ementati on 

The latest issue of 10 CFR Part 21 along with Section 206 of the Energy 

Reorganization Act of 1974 and Cardinal's implementing procedure were observed 

to be posted in several conspicuous locations as required by the regulation. 

Review of the implementing procedure identified several weaknesses. 

Spec i f i call y : 

0 In several paragraphs of the 10 CFR Part 21 implementing procedure, the 

term 

specifically defined in the regulation and can not be interchanged. 

defect" was used in place of "deviation". These terms are 

0 The procedure did not contain provisions for informing Cardinal's 

customers of a deviation when Cardinal can not determine whether this 

deviation is also a defect. 

0 The procedure did not require documentation of eval uati ons to determi ne 

whether a deviation is also a defect. 

0 The procedure did not specifically require that the NRC be notified when 

a defect is determined to exist. 

conditions he would follow the 10 CFR Part 21 requirements. 

The QA manager stated that under these 

These weaknesses were discussed with the QA Manager who indicated that the 

imp1 ementing procedure would be revised to address these concerns. 

3.3.2 Follow-up on 10 CFR 21 report by Pennsylvania Power & Light Company 

(PPL) concerning cracked ASME Class 1 bolts at the Susquehanna Steam 

El ectri c Stat i on. 

On April 19, 1994, PPL reported the discovery of several cracked 1 inch ASME 

Section 111, Class 1 hex bolts at the Susquehanna plant. 

manufactured to ASME SA 193 GR.B7 requirements by CIPC from material heat no. 

111890 and were obtained from Capitol Pipe & Steel Products Inc. (Capitol) in 

1985. 

bolt condition was attributed to "quench cracking" that had occurred during 

the original manufacturing cycle. 

These bolts were 

The bolts were used in various safety and non-safety applications. The 

PPL reported that they had been in contact with Cardinal regarding the cracked 

bolts but had not been able to obtain any procurement or quality records 

because of the company's new ownership and the absence of records from the 

time when these bolts were manufactured. 

function of each safety related bolt from the suspect heat and conducted a 

failure mode and effects analysis. No substantia7 safety hazards were 

identified. 

PPL determined the location and 

PPL indicated that bolt replacements would be made as required. 

As noted in Section 3.2, two changes in company ownership have occurred since 

these bolts were supplied to PPL. According to Cardinal management, many 

3 

64 
records were lost before Hub acquired CIPC after its bankruptcy. During this 

inspection, Cardinal staff searched their computer files for stock inventory 

records related to material heat 111890 and failed to locate any information 

associated with this materi a1 . 

3.3.3 Fol 1 ow-up on nonconforming materi a1 received by Southern Cal i forni a 

Edison (SCE) 

In 1993, SCE returned several orders of fasteners supplied by Nova Machine 

Products Corporation (Nova) to the manufacturer after their chemical overcheck 

analyses identified that this material was outside the applicable 

specification limits. The material was in the form of ASME SA 194, 1 1/4-7 

heavy hex nuts and 2 inch threaded rod, which Nova manufactured from 2 1/4 

inch diameter bar (Republic Engineered Steel heat 8094865) supplied by Hub. 

Nova's tests on the returned material indicated chemical composition within 

acceptable limits, however, some of the mechanical test specimens failed to 

meet the required yield strength minimums suggesting a nonliomogeneous heat of 

material. Because of Hub's prior ownership of CIPC, the inspectors requested 

a record search to determine if any material from Republic Engineered Steel 

heat 8094865 was in Cardinal's material inventory. None was found. 

3.4 Cardinal Commercial Grade Dedication Proqram 

3.4.1 Methodology 

The requirements for Cardinal's commercial grade item (CGI) dedication process 

are prescribed in Addenda No. 1 , Sect i on 8.0 , "Control of Purchased Materi a1 

and Services, I' to the Cardinal Quality System Manual , "ASME Quality System 

Certificate Holders," First Edition, Revision 8, dated January 8, 1994. 

Additionally, Cardinal Standard Practices (CSPs) provide requirements for 

performing and documenting various dedication activities. The following CPSs 

were reviewed during the NRC inspection: 

e 

e 

e 

e 

CSP No. 4.001, "Purchase Order Review," Revision 6, dated 

December 29, 1994, 

CSP No. 9,001, "Material Receiving Inspection," Revision 11, dated 

April 20, 1994, 

CSP No. 10.010, "Testing and Examination of Ferrous aind Non-Ferrous 

Materials," Revision 9, dated July 8, 1994, 

CSP No. 11.001, "Final dimensional Inspection," Revision 8, dated 

October 10, 1994, and 

CSP No. 11.002, "Final Visual Inspection of Fasteners," Revision 8, 

dated December 29, 1992. 

Cardinal selects critical characteristics for CGIs based on the testing 

requirements specified by the applicable ASME Boiler and Pressure Vessel Code 

4 

65 
(Code) , Section 11, "Materials," (Section 11) or American Society for Testing 

and Materials (ASTM) material specifications and any other testing 

requirements specified by the customer. Critical characteristics, in general, 

include dimensions, workmanship, and, when required by the applicable material 

specification, chemistry and mechanical properties and other specified tests 

such as an oxalic acid etch test. The verification of critical 

characteristics is performed using sampling plans as discussed in 

Section 3.4.6 of this inspection report. 

inspections performed are discussed in Section 3.4.7 of this report. 

Additionally, the dimensional 

The identification and verification of critical characteristics are controlled 

by the Cardinal Procurement Review Sheet (PRS) and a shop traveller, 

identified as the Customer Manufacturing Record (CMR). The PRS, CMR, along 

with visual, dimensional, and testing records are collected and reviewed, and 

serve as the basis for Cardinal to issue a certification statement that the 

dedicated CGIs meet the applicable material and customer purchase order (PO) 

requirements. 

Cardinal also uses its CGI dedication process to control til2 supply of 

products in accordance with the small products exclusion provisions of NX-2610 

and the 3/4 inch and less material provisions of NCA-3800 of the ASME Code 

(see Section 3.4.8 of this report for additional discussion). 

3.4.2 Dedication Program Strengths 

The NRC inspectors considered the following to be strengths in Cardinal's 

dedication program: 

0 Cardinal selects critical characteristics based on the testing and 

materi a1 requirements of the appl icabl e materi a1 specification and its 

customers. 

0 Cardinal personnel performing testing , inspection , and document review 

activities were knowledgeable about their work and had a positive 

attitude. 

0 Cardinal 

of items supplied from the manufacturer (qualified and non-qual ified 

manufacturers) and reviews the certification, when received, for 

conformance with the applicable material specification. 

testing capability. 

attempts to obtain material certification for each item or lot 

0 Cardinal has good in-house chemical, mechanical, and nondestructive 

0 Cardinal performs comprehensive dimensional inspections of fasteners. 

5 

66 
3.4.3 Dedication Program Weaknesses 

The NRC inspectors concluded that Cardinal's CGI dedication program addresses 

the essential elements of the dedication process and that sufficient guidance 

for performing activities such as inspection and testing are provided. 

However, the NRC inspectors considered the following to be weaknesses in 

Cardinal's CGI dedication program: 

Cardinal's procedures do not contain provisions for verifying critical 

characteristics such as material strength by methods other than 

destructive testing. survey/audit to accept stress/strain charts in lieu of testing (see 

Section 3.4.4.7 of this report). Also, some procedures are not 

prescriptive in defining the various types of tests to be controlled by 

Cardinal's sampling plan (see Section 3.4.4.6 of this report). 

For example, Cardinal used an undocumented 

Cardinal has no documented bases, (qualitative or quantitative) for- its 

destructive sample plan used !n verifying critical characteristics (see 

Section 3.4.6 of this report). 

When material overchecks such as chemistry and tensile testing are 

performed to verify critical characteristics, there are no procedural 

requirements or guidance provided in Cardinal's dedication program for 

determining the amount that these test results may deviate from those 

listed on the manufacturer's certification or to requ'ire that such 

devi at i ons be eval uated. 

The NRC inspectors questioned Cardinal's practice of including 

nonqual ified supplier material certifications (stamped "QA Accepted") in 

documentation packages supplied to customers. 

certi f i cati ons are only reviewed for compl i ance to the appl i cab1 e 

materi a1 specification requirements as part of the i n-i ti a1 screening 

process for the CGIs received. 

stamp, che customer could assume that Cardinal had verified the test 

results reported in these certifications. 

These material 

However, based on the "QA Accepted" 

3.4.4 Cardinal CGI Dedication Program Implementation 

The NRC inspectors reviewed several compl eted CGI dedi cati oris packages to 

determine if the critical characteristics for the items had been properly 

identified and verified, and if adequate procedural controls were in place. 

The NRC inspectors a1 so observed i n-process inspection and testing activities 

and the processing of CMRs. The NRC inspectors reviewed the following 

completed CGI dedication packages. In addition to the identified technical 

requirements, all customer POs invoked the requirements of l\ppendix B to 

10 CFR Part 50 and 10 CFR Part 21. 

3.4.4.1 CMR No. 54602-8, dated August 24, 1994, was for the supply of 100 

heavy hex head nuts, 3/4-10, SA 194, Grade 6, in accordance with Tennessee 

Valley Authority (TVA) Contract No. P-92NMB-45217B/PO No. 1030568. Cardinal 

6 
purchased 125 of these nuts from a nonqualified supplier, B & G Manufacturing 

Company, Inc. , Hatfield, Pennsylvania. Cardinal verified that markings and 

selected dimensions were correct, and reviewed the nonqualified manufacturer’s 

material test report for conformance with the material specification 

requirements, and performed spectrochemical analyses on 2 of the 125 nuts, a 

proof load test on 2 of the 125 nuts, and a hardness test on 2 of the 125 

nuts. The test results met applicable material specification requirements and 

were, in general , consistent with the nonqual ified manufacturer’s material 

certi f i cati on. 

Based on the above dedication activities, Cardinal certified that the nuts 

were in accordance with the requirements of ASME Section I1 and Section 111, 

Class 2, 1989 Edition, that NCA-3867.4 and NX-2610 applies (see Section 3.4.7 

of this report for additional discussion), and that the nuts were quenched and 

tempered (1100 degrees F minimum). 

statement appears to be based on the nonqual i f i ed manufacturer’ s cert i f i cat i on 

and results of hardness testing. 

The quenched and tempered condition 

3.4.4.2 CMR No. 51670-19, dated May 24, 1993, was for the supply of 5000 hex 

head capscrews, 1/2-13 by 2 inch long, A 307, Grade A, in accordance with TVA 

Contract No. P-92NMB-45217B/PO No. RD348130. Cardinal purchased 5400 of these 

capscrews from a nonqualified supplier, L & J Fasteners, Hillard, Ohio. 

Cardinal verified that markings and selected dimensions were correct and 

reviewed the nonqualified supplier’s documentation for these capscrews. 

supplier provided a test report generated by INFASCO (Quebec, Canada), the 

nonqualified manufacturer of the capscrews. Cardinal performed a 

spectrochemical analyses on 13 of the 5400 capscrews, tensile (wedge test) 

tests on 13 of the 5400 capscrews, and hardness tests on 13 of the 5400 

capscrews. 

The 

The test results met applicable material specification requirements and were, 

in general , consistent with the nonqualified manufacturer’s material 

certification. Cardinal also contracted A-1 Plating, Henderson, Nevada, to 

cadmium plate the capscrews. 

Cardinal certified that the capscrews were in accordance with its customer’s 

PO requ i rement s . 

Based on the above dedication activities, 

3.4.4.3 CMR No. 54511-1, dated August 6, 1994, was for the supply of 10 hex 

nuts 7/8-9, A 194, Grade 2H, in accordance with Commonwealth Edison Contract 

No. 503085/PO No. XX176. Cardinal purchased 125 of these nuts from a 

nonqualified supplier, B & G Manufacturing Company, Inc., Hatfield, 

Pennsylvania. Cardinal verified that markings and selected dimensions were 

correct , and reviewed the nonqual if i ed manufacturer’s materi a1 test report for 

conformance with the material specification requirements, and performed a 

spectrochemical analyses on 2 of the 125 nuts, a proof load test on 2 of the 

125 nuts, and a hardness test on 2 of the 125 nuts. Cardinal also performed 4 

additional hardness tests on nuts subjected to an additional heat treatment at 

1000 degrees F for 24 hours in accordance with A 194 requirements. 

7 

68 
The test results met ’applicable material specification requirements and were, 

in general , consistent with the nonqualified manufacturer’s material 

certification. Based on the above dedication activities, Cardinal certified 

that the nuts were in accordance with its customer’s PO requirements. 

3.4.4.4 

feet of 1 1/8-inch rod to be supplied as American Iron and Steel Institute 

(AISI) Grade 316 stainless steel. Castle Metals supplied this material in 

five pieces with Mill Certification from Roldan, SA (unqualified Spanish 

company); Cardinal tested samples from two of the bars (tensile, chemical, 

hardness) and accepted all five bars based on satisfactory test results. This 

material was subsequently used to make threaded rod. Nine three-foot lengths 

of the threaded rod were later supplied to GPU Nuclear Corp. certified as 

complying with ASTM 193 B8M. 

PO 552058 was issued to Castle Metals (unqualified vendor) for 77 

3.4.4.5 CMR (CIPI Stock) Requisition No. 100584, dated November 14, 1994, 

was for the supply of 12 heavy hex bolts, 3/4-10, by 3 inch ‘long, SA 325, Type 

1, in accordance with TVA Contract No. P-92NMB-45217B/PO ;io. 1045143. 

Cardinal purchased 500 of these bolts from a nonqualified supplier, Cordova 

Bolt, Inc. , Buena Park, California. 

selected dimensions were correct, and reviewed the nonqualified manufacturer’s 

material test report for conformance with the material specification 

requirements. The supplier provided a test report generated by NUCOR 

Fastener, Saint Joe, Indiana, the nonqual ified manufacturer of the bolts. 

Cardinal performed a spectrochemical analyses on 3 of the 500 capscrews, 

tensile (wedge test) tests on 3 of the 500 bolts, and hardness tests on 3 of 

the 500 bo1 ts. The test results met appl icable material specification 

requirements and were, in general , consistent with the nonqual ified 

manufacturer’s material certification. 

Cardinal verified that markings and 

Based on the above dedication activities, Cardinal certified that the bolts 

were in accordance with the requirements of ASME Section I1 and Section 111, 

Class 2, 1989 Edition, that NCA-3867.4 and NX-2610 applied (see Section 3.3.7 

of this report for additional discussion), and that the bolts were quenched 

and tempered (minimum 800 degrees F). 

statement appeared to be based on the nonqual ified manufacturer’s 

certi f i cation. 

The quenched and tempered condition 

3.4.4.6 

stainless steel hex nuts, 3/8 inch-16 (2B), SA F 594, Grade 304, in accordance 

with Energy PO No. 94-6-74057, Revision 0. 

CMR 54438-1RAY dated October 4, 1994, was for the supply of 100 

Cardinal purchased 300 of these nuts from a nonqualified supplier, Bell 

Fasteners, Santa Fe Springs, California. Cardinal verified that markings and 

selected dimensions were correct, and reviewed the nonquaiified manufacturer’s 

materi a1 test report for conformance with the materi a1 specification 

requirements. 

Fastener, Saint Joe, Indiana, the nonqualified manufacturer of the nuts. 

Cardinal performed a spectrochemical analyses on 3 of the 300 nuts, tensile 

(proof load) tests on 3 of the 300 nuts, and hardness tests on 3 of the 300 

The supplier provided a test report generated by NUCOR 

8 
nuts. The test results met applicable material specification requirements and 

were, in general , consistent with the nonqual ified manufacturer’s material 

certification. Cardinal also performed an intergranular corrosion test 

(oxalic acid etch test) on one nut. The results of this test indicated an 

acceptable microstructure of the nut tested. 

activities, Cardinal certified that the nuts were in accordance with its 

customer’s PO requirements. 

Based on the above dedication 

The NRC inspectors questioned why Cardinal had not verified the microstructure 

of 3 nuts as required by Section 8.0 of Addenda 1 to the Cardinal Quality 

System Manual. Cardinal informed the NRC inspectors that error was the result 

of its procedures not being prescriptive in the area of destructive testing 

and that the intergranular corrosion test should have been identified as an 

example of a test controlled by Cardinal’s destructive sampling plan. 

Cardinal also informed the NRC inspectors that it was in process of revising 

its procedure to clarify which destructive tests are within the scope of its 

sample plans for destructive testing (NRC Inspector Follow Up Item 

3.4.4.7 CMR No. 53694-1, dated April 27, 1994, was for the supply of 20 

special tube clamps, three directional , Girard Part No. 3/8-1/2 T500SS3DY in 

accordance with Radnor A1 1 oys , Inc. (Radnor) PO No. RA-26732-NWY with Change 

Order No. 1, dated March 24, 1994. Radnor’s PO, in part, invoked the 

following quality and technical requirements: 

9990107“,‘94-01-01). 

0 Clamps are to be stainless steel and meet the requirements of ASME SA351, Grade CF8, in accordance with ASME Section 111, Subsection NF, 

C1 ass 1 , 1977 Edition through Winter 1979 Addenda requirements , i ncl udi ng NCA-3800. 

Appendix B to 10 CFR Part 50, 10 CFR Part 21, and Baltimore Gas and 

Electric Specification 10280, Revision 11, dated February 25, 1994, 

Radnor’ s PO a1 so invoked the requirements of 

0 Radnor’s Change Order No. 1 authorized the clamps to be provided in 

accordance with the provisions of NX-2610. 

Cardinal issued PO No. 8710901, dated April 4, 1994, to Girard Development, 

Inc. (Girard) , a nonqual ified suppl ier, for 25 of the clamps. Cardinal’s PO 

to Girard did not invoke any specific quality controls nor did it invoke 10 

CFR Part 21. However, the PO did require, in part, the following: 

0 Material must be from the same heat lot. 

0 Vendor must supply dimensional drawings or cut sheets and CMTRs for each 

lot. 

0 Trace codes must be stamped on each piece and traceable to the CMTR. 

0 Material must comply chemically to SA 351, Grade CF8. 

9 

70 
0 Suppl ier must provide stress/strain charts for mechanical testing, 

traceable to the heat code. 

0 Certification must accompany materi a1 and must i ncl ucle Cardinal ’ s order 

number, trace code, material description, QSC number and expiration date 

or statement that material has been purchased in accordance with the 

above referenced manual/procedure, including revision and date, as 

appl i cabl e. 

Cardinal performed the following activities to dedicate the commercially 

procured clamps and to supply them as ASME Code, Section I11 material. 

0 

Cardinal verified that markings were correct and reviewed the 

nonqual if ied manufacturer’s materi a1 test report for conformance with 

the materi a1 specification requirements. 

Cardinal performed the following dimensional checks on each clamp: a) 

53lt hole; b) 1/2 inch and 3/8 inch holes; c) 3/8 inch and 1/2 inch 

spacing; d) thickness; and e) squareness. Cardinal a1 so performed 

spectrochemical analyses on 2 of 25 clamps and the results met 

appl i cabl e materi a1 speci f i cation requirements and were, in general , consistent with the nonqual ified manufacturer’s material certification. 

Addenda No. 1 to Cardinal’s QA manual requires verification of the physical 

properties of a dedicated CGI when the applicable material specification 

identifies such physical properties. The NRC inspectors questioned why 

Cardinal had not verified the physical properties of the clamps as specified 

by the material specification for the clamps. Cardinal informed the NRC 

inspectors that the physical testing was not performed because the clamps 

received from Girard were too small for machining a tensile test specimen. 

The NRC inspectors also questioned why a sub-size specimen was not machined 

and requested Cardinal’s justification for not verifying the physical 

propert i es of the cl amps. 

Cardinal informed the NRC inspectors that as an alternative to verifying the 

clamp’s critical characteristic, tensile strength, Girard had been requested 

to provide stress/strain charts for mechanical testing that were traceable to 

the heat code of the clamps supplied to Cardinal. The NRC inspectors then 

questioned Cardinal on accepting the stress/strain charts from Girard, a 

nonqualified supplier. Cardinal informed the NRC inspectors that it had 

audited Girard on April 8, 1994, but was unable to fully qualify them. 

However, Cardinal stated that based on its audit, Girard appeared to have 

sufficient controls in the area of generating the stress/strain charts and 

Girard’s subcontractor, Southern Tool, Inc., Anniston, Alabama, had adequate 

controls for material traceabil ity. The NRC inspectors reviewed the results 

of Cardinal’s audit at Girard and further discussed the use of the Girard 

audit for accepting the stress/strain charts instead of performing testing to 

verify the physical properties of the clamps. 

10 
The NRC inspectors found that there are no provisions in Cardinal’s QA manual 

or procedures to use an audit, commercial grade survey, or other methods as an 

alternative to performing testing to verify critical characteristics such as 

chemical and physical properties. Although the NRC inspectors found that the 

use of the stress/strain charts appeared to be acceptable, Cardinal had not 

generated any documentation to justify the use of the stress/strain charts as 

alternative to performing testing. 

weakness in Cardinal’s dedication program. 

The NRC inspectors considered this to be a 

3.4.5 Nuclear Uti1 ities Procurement Issues Committee (NUPIC) 

3.4.5.1 NUPIC Audit of Cardinal 

During the NRC inspection, it was found that NUPIC had performed an audit of 

Cardinal on April 26 through 29, 1994. The NUPIC lead audit utility was the 

Yankee Atomic Electric Company (YAEC) with auditors from Northern States Power 

and Washington Public Power Supply System. A YAEC technlcal specialist 

evaluated Cardinal’s CGI dedication activities by reviewing 4 CMRs (CMRs 

reviewed included products such as bolts, nuts, cotter pins, and round steel 

tubing). The technical specialist determined that all products on the CMRs 

were being upgraded/dedicated by performing vi sua1 and dimensional inspections 

and chemical and mechanical tests. The YAEC technical specialist also 

determined that the critical characteristics were adequate for the 4 CMRs 

reviewed and that CGI dedication activities were being carried out in 

accordance with Cardinal’s procedures. 

that the Cardinal QA manual should reference the procedure control1 ing 

dimensional and visual inspections performed as part of receiving inspection. 

One minor observation by NUPIC was 

3.4.5.2 

The NUPIC audit report states, in part, that Cardinal’s dedication program is 

performed in accordance with Cardinal ’s approved and accepted procedures and 

that the critical characteristics for the 4 CMRs reviewed were determined to 

be adequate. Information, such as the following, was not provided in the 

NUPIC audit report and may be beneficial to NUPIC members (NRC licensees) in 

evaluating the use of items, dedicated by Cardinal, in safety significant/critical plant applications. 

NRC Inspector Comments on NUPIC’s Evaluation of Cardinal’s CGI 

Dedication Program 

0 Cardinal’s dedication program uses sampling for verifying critical 

characteristics such as chemistry, hardness, strength, microstructure, 

and dimensional checks. The NUPIC audit report does not discuss the 

acceptability of the sample size selected for verifying certain critical 

characteristics using destructive testing methods. 

report does not address the use of sampling by Cardinal (see Section 

3.4.6 of this report). 

The NUPIC audit 

0 The NUPIC audit report briefly addresses the upgrading of material by 

using the CGI dedication process, but does not clearly state that 

Cardinal supplies and certifies certain size material and products as 

11 

72 
meeting NX-2610 and NCA-3800 ASME Section I I I Code requirements using 

its CGI dedication program (see Section 3.4.8 of this report). This 

information may be beneficial to licensees using this size material in 

critical appl ications. 

3.4.5.3 Correspondence Between Cardinal and NUPIC 

During the review of Cardinal's correspondence to Girard, the NRC inspectors 

reviewed correspondence between Cardinal and NUPIC that addressed NUPIC and 

Cardinal audits at Girard. The correspondence reviewed indicated that NUPIC 

had been requested to provide additional information to Cardinal relative to 

their inspection of Girard. The NRC inspectors informed Cardinal that they 

had reviewed the correspondence between Cardinal and NUPIC and would follow up 

on the status of this correspondence (NRC Inspector Follow Up Item 

3.4.6 

Finished fasteners, purchased commerci a1 grade from nonqua 

dedicated as basic components, are inspected and tested in 

foll owing sampl e pl ans unl ess a customer specifies another 

1. Visual inspection is performed on 100 percent of the 

99901076/94-01-02). 

Use of Sampling in the CGI Dedication Process 

CPS No. 4.001). 

ified suppliers and 

accordance with the 

sampl i ng pl an. 

materi a1 (Reference 

2. Dimensional inspection is performed as follows and is based on Electric 

Power Research Institute (EPRI) guide1 ines (Reference CPS No. 4.001). 

It is noted that Cardinal considers the lot size to be the number of 

fasteners received from a nonqual ified supplier in ii single container, 

bundle, or batch. 

LOT SIZE SAMPLE SIZE REJECT* 

1-40 Entire Lot 1 

41-49 45 1 

50-55 47 1 

56-60 53 1 

61-75 59 1 

76-100 68 1 

101- 150 73 1 

201-300 78 1 

301-500 83 1 

501-1200 90 1 

1201-3200 125 1 

etc. etc. 1 

*Reject lot on one defect per sample. 

12 
3. According to Cardinal, destructive testing to verify critical 

characteristics such as hardness, strength, microstructure, and 

chemistry are performed on fasteners (bolts and nuts) and other CGIs 

being dedicated in accordance with the following sample plan (reference: 

CPS No. 4.001). 

LOT SIZE SAMPLE SIZE 

1-280 

281-500 

501-1 200 

1,201-3,200 

3,201-10,000 

10,001 and greater 

2 

3 

5 

8 

13 

20 

REJECT* 

"Reject lot on one defect per sample. 

The NRC discussed the use of Cardinal's sampling methodology in detail during the inspection and asked Cardinal to provide the bases for its destructive 

sampling plan as providing reasonable assurance that the t'asteners not tested 

had the required material chemistry and strength and other properties required 

by the applicable material specification. According to Cardinal, this sample 

plan is based on the ASTM A 325 sample size for the Shipping Lot Sample. 

Cardinal uses its destructive sampl ing plan for dedicating CGIs regardless of 

the size or type of product, whether the CGIs were purchased directly from the 

manufacturer or through a distributor, the use of single or multiple 

production lots, or the historical performance of the manufacturer or 

distributor. 

The NRC inspectors expressed a concern to Cardinal that the Shipping Lot Size 

sample plan provided in A 325 is for the manufacturer of the fasteners and is 

defined by A 325 as "that quantity of bolts of the same nominal size and same 

nominal length necessary to fill the requirements of a single purchase order." 

Also, the NRC inspectors identified that the quality controls in A 325 do not 

appear to meet the quality requirements of Appendix B to 10 CFR Part 50 and 

that such a sample size could include more than one production lot or heat of 

materi a1 . 

Cardinal informed the NRC inspectors that it had not established a 

quantitative acceptable quality level or reject level for its sampling plan 

nor an associated confidence 1 eve1 . appears to be based on undocumented qualitative factors such as engineering 

judgement and experience of its personnel in inspecting and testing fasteners, 

and not on using quantitative statistics. (Nonconformance 99901076/94-01-03). 

Cardinal 's destructive sampl e pl an 

3.4.7 Fastener Dimensional Inspections 

Cardinal typically performs the following dimensional and visual inspections 

on bolts: a) threads; b) pitch diameter; c) major diameter; d) thread length; 

e) grip gaging length; f) body diameter; g) width across flats; h) width 

across corners; 1) head height; j) fillet radius; k) runout of bearing 

surface; 1) bearing surface diameter; in) bearing surface height; n) 

straightness; 0) length; p) markings, and q) plating. 

13 

74 
Cardinal typically performs the following dimensional and visual inspections 

on hex nuts: a) threads; b) minor diameter; c) width across flats; 

e) thickness; f) runout of bearing face; g) position of tapped holes; 

h) degree of chamfer; i) washer pad diameter; j) washer pad thickness; 

k) countersink diameter; 1) marking, and m) plating. 

The NRC inspectors found that the visual and dimensional inspections, 

including the sample size for performing dimensional inspections for the CMRs 

reviewed, provided reasonable assurance that the fastener dimensions were 

acceptable for the dedicated CGIs. 

3.4.8 Cardinal's Use of CGI Dedication for ASME NCA-3800 (3/4 Inch and Less 

Excl uded Materi a1 ) and NX-2610 (Small Products) 

Cardinal uti1 izes their CGI dedication program when supplying ASME Code 

material exempted from the requirements of NCA 3800 by Section I11 paragraph 

NX-2610 unless specifically prohibited by the customer's PO. This approach is 

des-*-fbed in their ASME Quality Systems Manual. 

Paragraph NX-2610 of the ASME Code, Section I11 permits certain small 

products to be furnished as ASME Code, Section 111, material with a 

Certificate of Compliance certifying that the material is furnished in 

accordance with the applicable material specification and the requirements of 

Section 111. However, NX-2610 further requires that for these small products, 

measures to assure that the applicable material specification requirements and 

Section I11 requirements are met shall be provided by a Certificate Holder. 

' 

NCA-9000 of the ASME Code, Section 111, defines Certificate Holder as an 

organization holding a valid N, NPT, or NA Certificate of Authorization issued 

by the ASME. Therefore, where the customer is not an ASME Certificate Holder, 

Cardinal's application of the NX-2610 exemptions for small products did not 

appear to fully comply with the requirements of NX-2610. 

3.4.9 Observation of In-Process Inspection and Testing 

During the NRC inspection, the NRC inspectors observed Cardinal personnel 

performing inspection, testing, and manufacturing activities. Personnel 

performing these activities demonstrated proficiency in accomplishing their 

activities and there were no abnormalities observed. 

3.5 Cardinal 's Proqram for Sum1 vi nq ASME Code Materi a1 

3.5.1 Program Scope 

CardSnal has been accredited by the ASME as a Material Manufacturer of ferrous 

and nonferrous bars, threaded fasteners, seamless fittings and flanges, and as 

a Material Suppl ier of these items as well as of castings, forgings, plates, 

tubular products, structural shapes, sheet, and hot rolled wire and rod. The 

accredited scope of supply also includes the qualification of other material 

manufacturers and suppl iers and performance and certification of materi a1 s 

14 

75 
testing, nondestructive examination, heat treating, and upgrading of stock 

material. According to Cardinal's management, fastener products compose 

approximately 99% of their sales. 

not inventoried but are purchased as required by special orders. 

Items such as flanges, pipe, and plate are 

Cardinal purchases material from qual ified sources or upgrades stock material 

purchased from nonqual i f i ed suppl i ers by performing and/or subcontracting a1 1 

testing necessary to upgrade such material. Programmatic controls of these 

activities are described in Cardinal 's ASME approved Quality Systems Manual 

and several referenced imp1 ementation procedures. Their materi a1 upgrading 

process is described in CSP No. 12.001, Revision 3, dated February 27, 1991, 

"Materi a1 Upgrading Process Non-qual i f i ed Vendors". Thi s procedure i s 

applicable to processing ASME Section I11 orders for products exceeding 1 inch 

nominal diameter and adequately addresses the Code requirements for this class 

of material. 

Cardinal's ASME Quality Systems Manual appears to be inconsistent with Section 

I11 of the Code in defining the conditions for utilization of small parts 

ex--$ion (NX-2610) by organizations in the material SUPF~,- chain. 

As indicated in paragraph 3.4.8, the inspectors identified that 

3.5.2 Program Implementation 

The inspectors reviewed selected POs and related documents for material that 

Cardinal had purchased and upgraded under their Qual i ty System Certificates 

(QSC) . The documentation reviewed i ncl uded materi a1 purchased from suppl i ers 

on Cardinal's approved vendor list as well as material procured from 

nonqual'ified suppliers. 

In reviewing the upgrading of stock material purchased from Castle Metals, the 

inspectors noted an example where Cardinal apparently did not assure full 

compl iance with their procedure requirements in accepting materi a1 

certification from unqualified suppliers. Specifically, in response to 

Cardinal's PO requirement for certification from the stock material supplier 

that no welding had been performed on the material, such certification was 

provided by a third party rather than the manufacturer and accepted by 

Cardinal as meeting the PO requirements. In this case, the material test 

report was issued by Acindar (foreign company) who manufactured and suppl ied 

the material to B.S. Livingston Co. 

form of a letter from B.S. Livingston Co. to Hy Alloy Steel Co. and was signed 

by a sales manager. 

hexagonal bar (heat 70759) which Cardinal later used to manufacture nuts and 

certified them as meeting as meeting the 1983 Edition of the ASME Code, 

Section 111, Subsection NC. 

The non-welding certification was in the 

The material in question was 1 1/4 inch AIS1 4140 

With the exception of the apparent procedural deviations noted above, based on 

the sample of documents reviewed, the Code procurement activities appeared to 

be generally conducted in accordance with the applicable Quality Systems 

Manual and procedural requirements. 

15 

76 
3.6 Transfer and Utilization of Stock Material from CIPC 

Cardinal's material stock includes material and finished fasteners that were 

acquired during the last exchange of the company's ownership. Some of this 

material dates back to the early 1980'~~ before the original company (CIPC) 

declared bankruptcy. In July, 1987, after CIPC's acquisition by Hub, Inc., 

CIPI (the new company) issued CSP No. 22.001, "Transfer of Material and 

Associated Documentation Between CIPC and CIPI. I' This procedure is currently 

referenced in Cardinal 's Qual ity Systems Manual and contains a check1 ist of 

items which need to be verified before the material can be accepted for use in 

ASME or non-Code safety re1 ated appl i cati ons. 

Review of selected documentation packages of this material indicated that 

several heats or lots of .the material were marked "rejectetl" on the document 

review checklist. 

traceability to the manufacturer or the lack of specific documentation. 

was noted that these rejections had never been dispositioned although material 

from these heats/lots is still .being used to fill custoc':;- orders. According 

to Cardinal's staff, a CMR is prepared for each order. The CMR identifies all 

steps required to assure compliance with the customers PO. This CMR assures 

that all applicable documents and/or tests required to certify this material 

are obtained or performed. Review of several customer orders indicated that 

this approach was generally foll owed. 

Most of the rejections were based on inadequate material 

It 

One lot of such material involved 2 1/4 - 8 heavy hexagonal nuts (heat 08876) 

which had been procured on PO 001912 from Fast-ER-Tech Trading, Osaka, Japan. 

The mill test report for this material was issued by Aichi Steel Works while 

the finished nuts were certified by Hamanaka Nut Coy Ltd. 

certification included impact tests showing acceptable results at 0 degrees F 

and also at -150 degrees F. 

Hamanaka's 

This material was reviewed in September 1987 in accordance with CSP No. 22.001 

and rejected for inadequate traceability and lack of QA signatures. 

material had been used, however, to fill customer orders on several occasions 

after this review, thb latest being June 1994. Review of CMRs associated with 

these orders indicat4d that the documentation required by -these orders had 

been obtained, however the "reject" indicated on the CSP checklist had not 

been dispositioned. 

The 

It was noted that CIPI had subcontracted Eagle Mountain Testing to perform 

additional impact tests on this material in October 1988 apparently as a part 

of their sampling procedure to fill an order to ASME Code requirements. The 

tests failed with the impact test results being significantly lower than 

certified by Hamanaka Nut Coy Ltd. Although these nuts were not used to fill 

the particular order that prompted these tests, the materi.al was not rejected 

for failure to verify the manufacturer's certification, and material from this 

lot was used to fill subsequent orders not requiring impact testing or, in one 

case, where impact testing was specified at 72 degrees F based on performing a 

successful retest at that temperature. The inspector expressed concern about 

16 
continued use of a heat/lot of material without adequate justification when 

overcheck of the certified test values produces unacceptable results. 

Failure to initiate a nonconforming condition report and to evaluate the basis 

for continued use of this material was identified as a nonconformance with 

Criterion XV of 10 CFR 50, Appendix B (Nonconformance 99901076/94-01-04). 

3.7 Control of SPecial Processes 

The inspector examined Cardinal’s heat treatment facilities and reviewed CSP 

No. 18.001, Rev. 5, which controls this activity. 

this procedure did not provide specific parameters (time/temperature) for 

several of the heat treating operations routinely performed by Cardinal. 

According to Cardinal staff , the necessary heat treating parameters are 

available in uncontrolled notes and tables which are used to program the 

controls of their heat treatment furnace. In response to the inspector’s 

questions concerning the use of uncontroll ed documents, Cardinal ’s management 

stated ‘kat the critical parameters would be incorporated in CSP No. 18.001. 

A revision of this procedure containing these parameters was accomplished 

before the completion of this inspection. 

The inspector noted that 

4. PERSONNEL CONTACTED 

*+ 

*+ 

*+ *+ *+ 

*+ 

*+ 

*+ 

+ 

+ 

Scott Akers, Chief Executive Officer 

David Hathcock, Vice President, Qual i ty Assurance 

Vincent Cortez, Vice President, Operations 

Chuck Mapes , Materi a1 s Manager 

Norman Henderson , Laboratory Manager 

Earnest Roberts, Chief Qual ity Control Inspector 

Greg Kell er , Technical Services Manager 

Mike Ti beri 0, Sales 

Russ Ni ermayer , Sal es 

J. Lee Conrad, Quality Assur-ance Technician 

Steven Gautlier, Laboratory Technician 

* Attended the Entrance Meeting + Attended the Exit Meeting 

17 

78 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-0001 

October 11, 1994 

Docket Nos. 52-004 and 99900403 

Mr. Patrick W. Marriott, Manager 

Advanced P1 ant Techno1 ogies 

GE Nuclear Energy 

175 Curtner Avenue 

San Jose, California 95125 

Dear Mr. Marriott: 

SUBJECT: NRC INSPECTION REPORT NO. 99900403/94-02 

This letter addresses the inspection of your facility at San Jose, California, 

conducted by Richard P. McIntyre and Billy H. Rogers of the Nuclear Regulatory 

Commission's (NRC's) Vendor Inspection Branch, Alan E. Levin of the Reactor 

Systems Branch, Joseph L. Staudenmeier of the Analytical Support Group, 

Robert A. Gramm and Frederick R. Allenspach of the Performance and Quality 

Evaluation Branch, and Tim M. Lee of the Office of Nuclear Regulatory Research 

on June 21 through 23, 1994. The details of the inspection were discussed 

with you and your staff during the inspection and at the exit meeting on 

June 23, 1994. 

The purpose of the inspection was to determine if activities performed to 

suprJrt the design of the Simplified Boiling Water Reactor (SBWR) and 

spccifically, the Gravity Driven lntegml rull-Height Test for Passive Heat 

Removal (GIRAFFE) test program performed by Toshiba in Kawasaki City, Japan 

were conducted under the appropriate provisions of the GE Nuclear Energy 

(GE-NE) 10 CFR Part 50, Appendix By quality assurance (QA) program as 

imp1 emented by NED0 1 1209-04AY Revi sion 8, tne most recent Qual i ty As'surance 

Program Description that has been approved by the NRC, NEDG-31831, 

"SBWR Design and Cert i f i cat i on Program Qual i ty Assurance P1 an, I' and 

NEDG-31836, "SBWR Team Organization and Procedures Manual," prepared for 

Department of Energy (DOE) Contract No. DE-AC03-90SF18494.. The inspection 

also reviewed GE-NE corrective actions for certain nonconformances identified 

in NRC Inspection Report No. 99900403/93-01. 

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

enclosed inspection report. 

procedures and representative records, interviews with personnel , and 

observations by the inspectors. 

records and documents that are maintained by Toshiba at the GIRAFFE test 

facility in Japan. 

The inspection consisted of an examination of 

The inspection did not include a review of 

The results of the inspection indicate that GE-NE failed to ensure that an 

adequate quality assurance program was in place and properly implemented by 

Toshiba for the GIRAFFE Phase 1 test activities as would be required for 

design basis tests by the GE-NE Quality Assurance Program Description and the 

79 
Mr. Patrick W. Marriott -2- October 11, 1994 

GE-NE implementing procedures. The team identified that GE-NE had not imposed 

NQA-1 or any other specific quality requirements in their technical agreement 

with Toshiba for the GIRAFFE tests. 

had concluded in a December 1990 review that Toshiba did not have a QA program 

implemented for SBWR Phase 1 test activities that met NQA-1-1983. 

addition, the team determined that GE-NE had not audited or reviewed the 

Toshiba Phase 1 testing activities during that time period and that Toshiba 

did not formally document a SBWR QA program until December 16, 1993. Based on 

the above, it appears that GIRAFFE tests w2re conducted as developmental tests 

and that GIRAFFE Phase 1 data should not be used, in other than a confirmatory 

mode for data generated by other tests, for TRACG code qualification as part 

of SBWR design certification. 

Further, GE-NE quality assurance staff 

In 

In response to the above findings, GE-NE submitted letter MFN No. 087-94, 

dated July 1, 1994, to Docket No. STN 52-004, acknowledging the fact that 

GIRAFFE was a developmental test conducted in a disciplined, professional 

manner, but not explicitly under the requirements of NQA-1. 

stated that PANDA and PANTHERS are the principal tests to be used as design 

basis tests for the SBWR design certification program and it was GE-NE'S 

intention to use GIRAFFE data to substantiate the results of PANDA and 

PANTHERS at another scale. Since GE-NE does not intend to use the GIRAFFE 

test data for code qualification as part 0.f design certification, no 

nonconformances or unresolved items have been identified in this inspection 

report. 

GE-NE further 

In addition, the team determined that GE-NE had verified that appropriate PA 

programs were in place or being developed for the PANTHERS (SIET) and PANDA 

(Paul Scherrer Institute) test facil ties. 

In accordance with 10 CFR Part 2.790 

of this letter and its enclosures wi 

Room. 

of the NRC's "Rules of Practice," a copy 

1 be placed in the NRC's Pub1 ic Document 

Should you have any questions concerning this inspection, we will be pleased 

to discuss them with you. 

R. W. Borchardt, Director 

Standardization Project Directorate 

Associate Directorate for Advanced Reactors 

Office of Nuclear Reactor Regulation 

and License Renewal 

Enclosure: 

1. In3oection Report No. 99900403/94-02 

FOR DISTRIBUTION AND CONCURRENCE SEE NEXT PAGE 

ao 
Mr. Patrick W. Marriott 

GE Nuclear Energy 

-3- 

Docket Nos. 52-004 

99900403 

cc: Mr. Laurence S. Gifford 

GE Nuclear Energy 

12300 Twinbrook Parkway 

Suite 315 

Rockvi 11 e , Maryl and 20852 

Director, Criteria & Standards Division 

Office of Radiation Programs 

U.S. Environmental Protection Agency 

401 M Street, S.W. 

Washington, D.C. 20460 

Mr. Sterling Franks 

U.S. Department of Energy 

Washington, D.C. 20585 

NE-42 

Mr. John E. Leatherman 

SBWR Licensing Manager 

GE Nuclear Energy 175 Curtner Avenue, M/C-781 

San Jose, California 95125 

Mr. Frank A. Ross 

Program Manager, ALWR 

Office of LWR Safety & Technology 

U.S. Department of Energy 

19901 Germantown Road 

Germantown, Maryl and 20874 

NE-42 

Mr. Victor G. Snell 

Safety and Li censi ng 

AECL Techno1 ogi es 

9210 Corporate Boulevard 

Suite 410 

Rockvi 11 e Maryl and , 20805 
Enclosure 1 

ORGANIZATION: GE Nuclear Energy 

San Jose, Cal i forni a 

REPORT NO.: 

CORRESPONDENCE 

ADDRESS : 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY: 

INSPECTION 

CONDUCTED: 

TEAM LEADER: 

99900403/94-02 

Mr. Patrick W. Marriott, Manager 

Advanced P1 ant Techno1 ogi es 

GE Nuclear Energy 

175 Curtner Avenue 

San Jose, California 95125 

Mr. Kenneth W. Brayman, Manager 

Qual i ty Assurance Systems 

GE Nucl ear Energy (GE-NE) suppl i es Advanced Boi 1 i ng 

Water Reactor (ABWR) and Simplified Boiling Water 

Reactor (SBWR) designs worldwide. 

furnishes engineering services, nuclear rep1 acement 

parts, and dedication services for commercial grade 

electrical and mechanical equipment. 

(408) 925-6587 

GE-NE also 

June 21 through 23, 1994 

Richard P. Mantvre Date 

React i ve Inspect: on Sect i on No. '1 

Vendor Inspection Branch (RVIB) 

OTHER INSPECTORS: Billy H. Rogers, RVIB 

Robert A. Gramm, RPEB 

Frederick R. A1 1 enspach, RPEB 

Alan E. Levin, SRXB 

Joseph L. Staudenmeier, ASG 

Tim M. LeeeRES 

APPROVED : - Z%-T4 

Uldis Potapovs, Chie Date 

Reactive Inspection iection NO. 1 

Vendor Inspection Branch (VIB) 

INSPECTION BASES: 10 CFR Part 50, Appendix B and 10 CFR Part 21 

-INSPECTION SCOPE: To determine if activities performed to support the 

design of the SBWR and specifically, the Gravity 

Driven Integral rull-Height Test for Passive Heat 

Removal (GIRAFFE) test program performed by Toshiba in 

Kawasaki City, Japan were conducted under the 

appropriate provisions of NED0 11209-04A, Revision 8, 

the most recent GE-NE Quality Assurance Program 

Description that has been approved by the NRC. 

PLANT SITE 

APPLICABILITY: None 

82 
1 INSPECTION SUMMARY 

1.1 Violations 

No violations were identified during this inspection. 

1.2 Nonconformances . 

No nonconformances were identified during this inspection. 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

2.1 (Open) Nonconformance 99900403/(93-01-01) 

Nonconformance 93-01-01 stated that contrary to Criterion 111 of Appendix B to 

10 CFR Part 50 and Section 4.4.1 of Engineering Operating Procedure (EOP) 

40-3.00, "Engineering Computer Programs" (ECPs), (1) the TRACG input decks 

used to model the gravity-driven cooling system integrated systems test (GIST) 

facility were not independently verified to be correct, and (2) the GE-NE Code 

Qual ification Document (CQD) , Licensing Topical Report NEDE-32177P, "TRACG 

Qualification, I' dated February 1993, which provides a description of the 

qualification of TRACG against various activities including the gravity-driven 

cooling system (GDCS) integrated systems test, was submitted to the NRC for 

review and approval for referencing in licensing actions for the Simplified 

Boiling Water Reactor (SBWR) without receiving independent design verification 

or design review as required for a level 1 code used to support design basis 

analyses. 

GE-NE committed to take certain corrective actions in response to this 

nonconformance. One commitment was to independently verify the TRACG input 

decks for GIST. A GE-NE internal audit of the corrective actions identified 

that the subsequent independent verification was not complete and therefore 

did not meetl paragraph 4.2.1 of EOP 42-6.00. 

the independent verification by September 30, 1994. Therefore, Nonconformance 

93-01-01 remains open. 

GE-NE has committed to complete 

2.2 

Nonconformance 93-01-02 stated that contrary to Criterion XVII of Appendix B 

to 10 CFR Part 50, EOP 42-10.00, "Design Record Files" (DRF), EOP 35-3.00, 

"Engineering Tests," and the GIST Program Test Plan and Procedure (TP&P) 

521.1322, Revision 2, dated November 29, 1988, certain documentation required 

to be contained or referenced in the DRF was not included therein. 

documents that should have been part of the DRF were: the Final Test Report 

(NEDO-31680) for the GIST Program; instrument calibration records, which were 

located in a desk drawer in another building; and final design drawings for 

the facility. Some drawings were found in a cabinet at the facility itself. 

This set of drawings did not include final numbered, approved, as-built desigr, 

drawings, which are required by the QA Plan to be retained for. the lifetime of 

the item. Also, data tapes for the GIST tests, which are part of the test 

records specified for inclusion in the DRF by TP&P 521.1322, were not 

referenced therein. 

(C1 osed) Nonconformance 99900403/ (93-01-02) 

Specific 

GE-NE has taken corrective actions that include the generation of as-built 

drawings and inclusion of those drawings in DRF A00-02917-1, the inclusion of 

certain post-test instrument calibration records in the DRF, the inclusion of 

83 
the Final Test Report (NEDO-31680) in the DRF, and the inclusion of 

documentat ion regarding GIST reactor pressure vessel (RPV) heat 1 osses. 

heat loss calculations were reviewed and some of the assumptions made in the 

calculations are open to question, such as the temperature distribution from 

the vessel surface to the outside of the insulation. However, the results of 

the calculation appear to be reasonable and come within about 50 percent of an 

indirect experimental measurement of heat losses. 

The "as-built" drawings included in the DRF are not "blueprint"-type drawings, 

but rather sketches of the facility dated May 1994. These drawings do, 

however, contain as-built dimensions and are indicated as having been verified 

and checked. Post-test instrument calibration is included for the GIST 

wattmeter, performed November 1993, approximately four years after completion 

of the test. 

35-3.00, and EOP 42-10.00. Nonconformance 93-01-02 is closed. 

These 

GE-NE'S corrective actions satisfy the requirements of EOP 

2.3 (Open) Nonconformance 99900403/(93-01-03) 

Not reviewed during this inspection. 

2.4 (Closed) Nonconformance 99900403/(93-01-04) 

Nonconformance 93-01-04 stated that contrary to Criterion XI1 of Appendix B to 

10 CFR Part 50, Section 2.2 of EOP 35-3.20, "Calibration Control," and Section 

4.1.2 of the GIST Program TP&P 521.1322, GE-NE purchased flow meters used in 

the GIST tests from a commercial grade supplier, not on GE-NE'S approved 

supplier list, and accepted and used the instruments as calibrated by the 

supplier without further verification of the quality or traceability of those 

cal i brati ons. 

The team verified that GE-NE had obtained certificates of calibration from the 

original supplier and had filed these in DRF A00-02917-1. In addition, since 

the supplier had not been on the GE-NE approved supplier list, the 

measurements obtained from the flow meter had been verified by calculation and 

documented in the DRF. The team reviewed Revision 7 of EOP 35-3.20, dated 

July 27, 1994, and determined that it had been revised to define measurement 

and test equipment to include all devices or systems (from the sensing element 

through the output or recording device) which could affect the accuracy of 

test qeasurements. Nonconformance 93-01-04 is closed. 

2.5 (C1 osed) Nonconformance 99900403/ (93-01-05) 

Nonconformance 93-01-05 stated that contrary to Criterion XI of Appendix B to 

10 CFR Part 50, EOP 35-3.00, "Engineering Tests," EOP 42-10.00, "Design Record 

Files," and Section 4.2.4 of the GIST Program TP&P 521.1322, GE-NE failed to 

document in the DRF the review and disposition of anomalies in three tests, 

CO1, Dol, and D03. These tests were considered to be "invalid" as a result of 

incorrect valve alignment (CO1) or incorrect power input to the test section 

(DO1 and D03). For one of the tests (COl), a note was found on the folder in 

the DRF in which hard-copy data plots were stored, indicating that a problem 

existed for the test; however, the problem indicated on the folder (incorrect 

-2- 

84 
power input) was not consistent with the actual reason given in NEDO-31680 for 

the test's invalidation (incorrect valve alignment). 

The team determined that the documentation and disposition of the anomalies 

had been included in DRF A00-02917-1. 

11 of EOP 35-3.00, dated July 27, 1994, which clarified the requirement to 

document the review and disposition of anomalies in the DRF. 

93-01-05 is closed. 

In addition, the t.eam reviewed Revision 

Nonconformance 

3 INSPECTION FINDINGS AND OTHER COMMENTS 

3.1 Quality Assurance Proqram 

The Quality Assurance (QA) program for the SBWR program is described in the 

"SBWR Design and Cert i f i cat i on Program Qual i ty Assurance 1'1 an, I' NEDG-31831 , dated May 1990. 

No. DE-AC03-90SF18494. This QA plan commits to meeting the requirements of 

ANSI/ASME NQA-1-1983 and its Addenda (NQA-la-1983). This plan is based 

primarily on the standard GE-NE "Quality Assurance Program Description, I' 

NEDO-11209-04AY Revision 8, dated March 31, 1989. The GE-NE topical QA 

program has been previously reviewed by the NRC and found to meet Appendix B 

of 10 CFR Part 50. 

The plan was prepared for the Department of Energy, Contract 

The principle difference between NEDO-11209-04A and NEDG-31831 is that the QA 

topical program conforms to ANSI N45.2 and the associated ANSI standards, 

while the SBWR QA plan conforms to NQA-l/la-1983. 

The QA plan described in NEDG-31831 contains a work element/implementing 

procedure matrix that contains 18 major subdivisions which correlate with the 

18 criteria of Appendix B to 10 CFR Part 50; and the supplemental requirements 

of NQA-la-1983. The matrix establishes the correlation between the criteria 

and four categories of GE-NE procedures and/or pol icies. These four types of 

GE-NE procedures/policies describe the implementation of the QA plan. These 

are GE Pol icy, GE-NE Pol iries and Procedures (PAPS) , BWR Engineering Operating 

Procedures (EOPs) , and Advanced Nucl ear Techno1 ogy (ANT) Operations Pol i ci es 

and Instructions (OP&Is). 

The GE-NE Policies and procedures are high level GE policies that establish 

overall policies and responsibilities for GE-NE. GE-NE nuclear activities are 

currently conducted under the responsibility of the Vice President of GE 

Nuclear Energy. 

policies and the QA program. ANT Operations Policies and Procedures deal with 

subjects such as schedules, budgeting, contract award, and business 

management. 

The EOP's are a series of procedures that implement GE-NE 

The SBWR QA organization reports to the Manager, Services and Projects 

Qual i ty. 

Quality,and a QA engineer provides QA overview of the SBWR project. 

A dedicated QA group of a Qual i ty Project Manager, Advanced Reactor 

The SBWR QA plan states that design work performed by the SBWR team will be 

performed in accordance with QA programs evaluated and accepted by GE-NE. The 

-3- 

85 . 
plan additionally states that design and testing work performed by 

International Technical Associates (ITAs) will be performed to their internal 

QA programs acceptable to the regulatory authorities of their *-espective 

countries. The plan stated that by December 31, 1990, GE-NE will perform an 

initial review to establish that the ITA'S QA programs comply to NQA-1 and its 

addenda. If def i ci enci es are found , speci al, QA procedures wi 11 be devel oped 

and applied to attain equivalence. After initial acceptance, GE-NE plans to 

perform QA reviews approximately on an annual basis to assure the ITA programs 

are effectively implemented. 

During the course of the inspection, cognizant GE-NE personnel stated that 

SBWR activities associated with Japanese Atomic Power Company (JAPC) were a 

special case that is not covered by the provisions stated in the SBWR QA plan. 

The justification for treating some ITAs in a manner not described within the 

SBWR QA Plan was not clearly evident to the team. 

requirements imposed on ITAs is further discussed in Sections 3.4 and 3.5. 

The aspect of QA 

The team additionally revkewed NEDG-31836 "SBWR Team Organization and 

Procedures Manual . I' 

team communications and as an operational guideline to produce consistent 

actions and documents by team members. 

project organization, procedures for control of correspondence, and reference 

to applying the SBWR QA plan. 

The %anual was devel oped to provide for e-ffect i ve SBWR 

The manual provides an overview of the 

The team found that a QA program has been established for the SBWR activities. 

In some cases, less than effective measures had been implemented to ensure 

that all organizations involved in the SBWR project had appropriate QA 

programs. 

3.2 Instruction. Procedures and Drawings 

The quality requirements utilized for the SBWR are contained in the following 

GE-NE documents that were reviewed by the inspection team. 

GE Policy 

GE QP 20.1, "Corporate Commitment" 

GE-NE Pol icies and Procedures 

70-1, "Nucl ear Energy Qual i ty Pol icy" 

70-11, "GE-NE Quality System Requirements" 

70-30, "Personnel Proficiency in Quality Related Activities" 

70-50, "GE-NE Handling and Storage of Quality Assurance 

Record s I' 

Enqineerinq ODeratins Procedures 

30-3.40, "Engineering Information System" 

30-5.00, "Engineering Records Documentation Supplied by 

Extern a 1 Sources 'I 

30-7.00, "Technical Design Procedures" 

-4- 

86 
35-3.00, "Engineering Tests" 

40-3 .OO, "Engineering Computer Programs" 

40-7.00, "Design Reviews" 

42-5.00, "Engineering Requirements Document Release" 

42-6.00, "Independent Design Verification" 

42-8.00, "Document Issue and Application by ERM" 

42-10.00, "Design Record File" 

45-4.00 , "Suppl i er Servi ces/Equi pment Document Review" 

55-2.00, "Engineering Change Control" 

60-3.10, "Engineering Records" 

66-2.10, "Safety Related Classification" 

EOP 10-2.00 states that EOPs provide uniform and required practices for 

control of technical and or quality related activities within GE-NE 

components. Engineers and Project Managers performing activities governed by 

the requirements of the EOPs are responsible for ensuring that activities are 

performed in accordance with the EOPs, or if impractical, the manager 

initiates appropriate EOP revisions. The EOPs allow for the generation of 

administrative guide1 ines that provide detailed how-to instructions for 

engineers . gui del i nes . Responsi bl e managers are a1 1 owed , but not required , to i ssue such 

ANT Operations Policies and Instructions 

BR 4-117, "Project Directive" 

BR 11-100, "Material Request and Cost and Lead Time" 

BR 11-104, "Supplier Evaluations" 

The OP&Is were written to provide specific instructions for the SBWR project. 

The instructions require that quality requirements for suppliers are to be 

designated on the Material Requests (MRs), either by direct incorporation or 

reference to other documents. 

subjected to NQA-1 for quality verification, access and need for QA approved 

suppliers. 

planned check1 ists. 

QA staff are to review each IIR with items 

QA provides for, and evaluates,' supplier quality system using preThe intent of the OP&Is appears to be to ensure appropriate level of QA 

involvement in material requests for supplied services to ensure translation 

of QA requirements into the documentation. 

circumvented in practice by technical services agreements between GE-NE and 

other team members and ITAs as further discussed in Sections 3.4 and 3.5. 

However, this has been 

The various classes of procedures and instructions were found to provide an 

appropriate framework to translate the upper tier quality provisions to the 

working level staff. 

3.3 Document Control 

Document control is prescribed by P&P 70-11 , "Qual i ty Systems Requirements" 

and numerous EOP's such as 15-2.00, "EOP Application," 30-5.00, "Engineering 

Records Documentation Supplied by External Sources, 'I 40-7 .OO, "Design 

Reviews, 'I 42-5.00 , "Engineering Requi rements Document Re1 ease, I' 42-6.00 , 

-5- 

87 
"Independent Design Verification," 42-8.00, "Document Issue and Application by 

ERM," 55-2.00, "Engineering Change Control, I' and 60-6.00, "Drafting Manual 

Control .I1 

document control provi si ons. 

These procedures were found to provide an acceptable framework of 

3.4 Control of Purchased Material, EauiDment. and Services 

The team reviewed the GE-NE quality assurance oversight activities related to 

Phase 1 of the GIRAFFE tests performed in 1989-1990 at the Toshiba test 

facility in Kawasaki, Japan, which GE-NE had used to support the certification 

of the SBWR design. GE-NE had not placed a purchase order with Toshiba, an 

International Technical Associate (ITA) for performance of the test, however, 

the testing agreement was documented in two contracts. 

The first contract, "Joint Study Contract for Feature Technology of Simp1 ified 

BWR (Phase 1) between the Japan Atomic Power Company (JAPC) and General 

Electric Company," dated March 30, 1989, included items such as use of the 

results, industrial property rights, 1 iabil i ty and warranty. The document 

includes a general warranty provision that "GE warrants to JAPC that the study 

will be performed in conformance with generally accepted professional 

standards prevailing in the nuclear industry at the time the study is 

performed" but the study contract imposes no specific QA requirements such as 

NQA-1-1983 and NQA-la-1983 to be imposed during the performance of the test. 

The Phase 1 study was a preliminary study of the containment vessel 

configuration, basic tests on the PCCS heat removal performance, and study of 

the GDCS and accumulators. 

The second contract, "Agreement on Execution of Joint Study Contract for 

Feature Technology of Simplified BWR (Phase 1) among Hitachi, Ltd. and Toshiba 

Corporation and General Electric Company, I' dated March 30, 1989, included 

items such as definitions, cost share, protection and use of information, 

workscope description, objective of the study (test), method of the study, 

contents of the study (test specifics), study period, work split, study 

schedule, study organization, and study reports. The document includes a 

general warranty provision that each "partner warrants to JAPC that its 

portion of the study will be performed in conformance with generally accepted 

professional standards prevailing in the nuclear industry at the time the 

Study is performed," but the agreement also contained no explicit QA 

requirements. 

GE-NE Advance Reactor Quality (SBWR/QA) requested the Toshiba SBWR QA Plan on 

November 29, 1989. SBWR/QA did not receive the QA Plan but did receive a list 

of standards and procedures on July 6, 1990, which Toshiba said applied to the 

SBWR work. The list of standards and procedures was presented in an informal 

matrix (handwritten) as opposed to a formal QA Plan. 

SBWR/QA visited Toshiba on October 29-30, 1990, (the period that the GIRAFFE 

Phase 1 test was being performed) in order to meet the commitments of the SBWR 

Quality Assurance Plan NEDG-31831 which required GE-NE to review and evaluate 

the 4.4 programs of the International Technical Associates. SBWR/QA "concluded 

that Toshiba considers its current work on the SBWR to be preliminary and has 

not applied a QA program that meets the applicable requirements of NQA-1-1983 

-6- 

88 
and NQA-la-1983 at this time. 

the direct application of Toshiba's final design work on the SBWR requires 

evidence of the application of a QA program that meets the appropriate parts 

of NQA-1-1983 and NQA-la-1983. 

applied to SBWR Design and Certification program in 12 to 18 months." These 

conclusions were documented in a letter to Toshiba, SBWF! Reference No. 

GETO-039, dated December 10, 1990. 

issued by SBWR/QA to ensure management attention and timely resolution. 

Qual i ty Project Manager for Advanced Reactor Qual i ty (SEWR QA Manager) 

indicated to the team that he had notified the responsible people in GE-NE of 

his results and requested to be notified if Toshiba's activities were to be 

used for final design, as opposed to preliminary uses, so that he could take 

the required actions to verify that an adequate QA program was in place. In 

addition, he indicated that he had inquired about testing during the review at 

Toshiba but had not been advised of the ongoing activities related to the 

GIRAFFE test. 

This will not affect preliminary work. However 

GE plans to review the [Toshiba] program 

No Corrective Action Request (CAR) was 

The 

In the Spring of 1993, a GE-NE engineer was on site at Toshiba during the end 

of Phase 2 of the GIRAFFE tests. 

between the engineer and the SBWR QA Manager. Two telefaxes, dated March 17, 

1993 and June 1, 1993, documented the engineer's attempt to determine whether 

Toshiba was applying a QA program to the Phase 2 activities. 

engineer was initially unable to determine whether a QA program was in place, 

the QA Manager suggested in a telefax dated June 3, 1993, that Toshiba NED 

standard procedures may have been used. 

1993, telefax that the Toshiba NED standard procedures were being used and 

provided documentation of implementation in several areas such as document 

control and inspection and test Control. The engineer's conclusions were 

formally documented in NEDC-32215PY "GIRAFFE Passive Heat Removal Testing 

Program," Revision 0, dated July 1, 1993. 

quality assurance program was applied to GIRAFFE testing and provided examples 

of the requirements related to Document Control and Inspection and Test 

Control. 

The team reviewed a series of correspondence 

After the 

The engineer responded in a July 2, 

The report stated the Toshiba 

On July 19, 1993, the SBWR QA Manager issued a letter to the engineer, also 

distributed to SBWR management, which stated that the engineer's documented 

verification of Toshiba Nuclear Engineering Laboratories's compliance to 

NQA-1-1989 applicable to the GIRAFFE test provided adequate evidence that work 

performed under the GIRAFFE program could be used directly for SBWR final 

design work. The basis for making this conclusion was not evident to the team 

as GE-NE was aware of several facets of the Toshiba QA program that did not 

address NQA-1 aspects such as: evidence of design verification, development of 

test procedures, and conduct of audits. 

On October 18, 1993, GE-NE issued a memorandum identifying that the three 

party agreement for the Phase I1 testing performed by Toshiba shall be under a 

QA program meeting the intent of JAEG-4101-1985. The team also reviewed an 

October 21, 1993, GE-NE document outlining the elements of an NQA-1 program 

that were not implemented during the GIRAFFE testing. 

verification of design activities, no formal test procedures, identification 

and control of items was not addressed, and no formal audit program. 

These included no 

-7- 
On December 16, 1993, Toshiba issued Document No. AS-50092, "Quality Assurance 

Program for Simplified Boiling Water Reactor," Revision 0. AS-50092 stated 

that prior to October of 1990 Toshiba was performing preliminary engineering 

and testing work on the SBWR program and that no QA program specific to the 

preliminary work was developed or documented. 

SBWR work progressed to support the final design the appropriate quality 

assurance procedures were appl ied and that AS-50092 documented the QA program 

that was applied to all final design work performed by Toshiba on the SBWR 

program since October 1990, and was intended to meet NQA-1-1983 and 

NQA-la-1983. 

AS-50092 further stated that as 

On March 14, 1994, the SBWR/QA visited Toshiba to review the QA program 

applied to the GIRAFFE tests conducted by Toshiba during the previous four 

years. This activity was documented in a trip report, dated April 28, 1994, 

which listed the comparison of GIRAFFE test activities to AS-50092. 

report documented the SBWR/QA conclusion of generally adequate implementation 

and an example in which the flowmeters in the GIRAFFE program were used based 

upon the manufacturer's certification without additional verification. In an 

April 15, 1994, internal GE-NE memorandum? SBWR/QA documented an additional 

comparison of the SBWR QA Manager's review, the GE-NE engineer's review, 

AS-50092, and the Japan Electric Association JEAG 4101-1990, "Guide for 

Qual i ty Assurance of Nuclear Power P1 ants, 'I with concl usi ons simi 1 ar to the 

April 28, 1994, SBWR/QA trip report. The comparison identified that initial 

calibration of procured instruments was not done, items were not procured from 

suppliers on an approved suppliers 1 ist, that while test anomaly documentation 

was weak, the tests were re-performed as necessary? and no test procedures 

were used as the system responded naturally after establishing the initial 

conditions. 

The 

In summary, the NRC determined that the appropriate quality requirements had 

not been included in the contracts with Toshiba, which established the 

agreement for the GIRAFFE tests, and GE-NE QA had been onsite at Toshiba in 

1990 and had not audited or reviewed that ongoing Phase 1 testing activities. 

In addition, GE-NE QA had concluded that the SBWR activities reviewed in 1990 

were preliminary, that Toshiba was not implementing a QA program which net the 

requirements of NQA-1-1983 and NQA-la-1983, and further QA actions would be 

required to accept the SBWR activities for final design work; Toshiba 

documented in AS-50092 that prior to October of 1990 no QA program which met 

the requirements of NQA-1 was being applied to the SBWR design and testing 

work being performed; and that the oversight provided by the GE-NE engineer 

was for Phase 2 tests being performed at a time three years later than the 

Phase 1 test supporting SBWR certification. 

The team concluded, based on a review of the documents and discussions with 

GE-NE personnel , that no quality requirements applicable to the GIRAFFE Phase 

1 tests had been contractually imposed on Toshiba and no GE-NE verification of 

quality activities had been made during performance of the GIRAFFE Phase 1 

tests. 

-8- 

90 
3.5 GE-NE RelationshiDs With Additional ITAs 

In terms of arrangements governing other SBWR test facilities, the team 

reviewed the quality activities related to PANTHERS and PANDA test programs, 

(neither of which had yet been performed) which are intended to support the 

SBWR design and re1 ated computer code ' qual i f i cat i on. These programs were the 

PANTHERS, to be performed by Societa' Informazioni Esperienze Termoidraul iche 

(SIET) in Italy, and PANDA, to be performed by Paul Scherrer Institute (PSI) 

in Switzerland. 

The team reviewed the contract (GE-NE had not used a purchase order) which 

established the agreement with SIET, for performance of the PANTHERS test, 

"Agreement among ENEL s. p. a, ENEA, ANSALDO, and GE-NE for SBWR Development , I' 

dated May 29, 1992, and determined that it contained requirements that 

specified the work be performed in accordance the ANSI/ASME NQA-1-1983 and 

NQA-la-1983. In addition, in September 1993 GE-NE had performed a review and 

evaluation of the QA program in place at SIET as documented in a report dated 

October 29, 1993. The report concluded that the SIET Quality Plan which would 

be used for the SBWR test activities met the applicable requirements of NQA-1- 

1983 and NQA-la-1983 and the work performed under this program could be used 

directly for the SBWR final design. 

specified the required quality assurance requirements to SIET and had taken 

adequate actions to verify that an acceptable quality assurance program was in 

pl ace. 

The team concluded that GE-NE had 

The team reviewed the contract (GE-NE had not used a purchase order) which 

established the agreement with PSI for performance of the PANDA tests, 

"Agreement between the Swiss Confederation, represented by the Paul Scherrer 

Institute and the General Electric Company and the Electric Power Research 

Institute, Inc. (EPRI) on Passive Decay Heat Removal and f:ission Product 

Retention Tests and Model Qualification," dated April 1, 1.991, and determined 

that the contract did not specify any quality requirements. 

performed a review and evaluation of the QA program in place at PSI with 

respect to the 18 criteria of NQA-1 as documented in a report dated October 

29, 1993. 

the SBWR test activities met the applicable requirements of NQA-1-1983 and 

NQA-la-1983 and the work performed under this program could be used directly 

for the SBWR final design. However, PSI was not contractually committed to 

using the Quality Plan which GE-NE had reviewed and accepted. GE-NE indicated 

to the inspection team that the required quality requirements, NQA-1-1983 and 

NQA-la-1983, would be included in the Test Specification that GE-NE was 

currently preparing for issuance to PSI. 

GE-NE had 

The report concluded that PSI Quality Plan whic:h would be used for 

Further evidence of the fact that QA programs are implemented by the SBWR team 

members was obtained during the team's review of GE-NE eva.luations of the QA 

programs. An October 15, 1990, GE-NE memorandum "Review a.nd Evaluation of QA 

Programs Applied to the SBWR by the European Technical Associates - Trip 

Report" concluded that ANSALDO, NUCON, and KEMA meet NQA-1. The ENEL QA 

manual was found to meet NQA-1 but it had not been implemented. ENEA and ECN 

programs were undefined . 

-9- 

91 
A similar GE-NE memorandum dated November 16, 1990, "Review and Evaluation of 

QA Programs Applied to the SBWR by the Japanese Technical Associates - Trip 

Report," documented that the Hitachi QA program was found satisfactory. An 

October 29, 1993, GE-NE document "Review and Evaluation of QA Programs Applied 

to the SBWR by the European Technical Associates - Trip Report" concluded that 

all of the ITAs (UTE, ENSA, KEMA, NUCON, ECN, ANSALDO, ANSALDO ACO, ENEA/S ET, 

and ENEL) were meeting the requirements of NQA-1-1983. The only exception was 

that CIEMAT was not. With the exception of the fact that NQA-1 provisions had 

not been explicitly invoked by GE-NE for the SBWR work performed by severa 

Team members and ITAs, the team noted no other concerns. 

3.6 Oual i tv Assurance Records 

Qual i ty Assurance record requi rements are speci f i ed by P&P 70- 11 , "Qual i ty 

Systems Requirements", and numerous EOP's such as 35-3.00, "Engineering 

Tests", 40-7.00 , "Design Reviews" , 40-9.00, "ASME Code Design Veri f i cat i an" , 42-6.00, "Independent Design Verification", 42-10.00, "Design Record Files", 

and 60-3.10, "Engineering Records Retention". 

EOP 42-10.00 describes the Design Record Files (DRF's) as formal, organized 

accumulations of information, which provide a controlled system for retention 

of documented engineering activities, necessary to substantiate significant 

design decisions. The DRF provides a mechanism for controlling and archiving 

important design records, such as design verification, studies and analyses. 

It does not include documents, such as drawings and specifications, which are 

maintained under separate corporate design controls. 

states that the DRF should Drovide for desiqn notes, calculations. records and 

The procedure a1 so 

other supporting information , and cross-reference to re1 ated or supporting 

DRF s . 

The team was informed by the cognizant GE-NE manager responsible for SBWR 

document retention that the SBWR team members and ITAs are currently holding 

their records rather than turning them over for GE-NE retention as outlined in 

NEDG-31836 section 2.6.2, that states "GE-NE is responsible for maintaining a 

complete physical file of ali relevant design input documents received frw 

participating organizations." 

The team reviewed selected portions of DRF T15-00004, Volume 6 that pertained 

to QA aspects of the GIRAFFE test. This information was previously discussed 

in Section 3.4. 

inspector was not able to draw any conclusions about the QA records review 

conformance with applicable GE-NE requirements. 

Based on the limited sample of records reviewed, the 

3.7 Desiqn Control 

GE-NE is responsible 

for SBWR design cert 

information and test 

code qual i f i cat i on. 

T15-00004. GIRAFFE 

for maintaining records of all testing information used 

fication. 

data was the GIRAFFE Experiment and Analysis DRF 

acility information and test data was also used for TRACG 

GE-NE'S main archive for GIRAFFE facility 

-10- 

92 
The team reviewed DRF T15-00004 for the GIRAFFE test program. GE-NE stated 

that it was not complete, in that key information, such as test procedures, 

instrument calibration records, test acceptance criteria, and as-built 

drawings, reside with Toshiba in Japan. However, some important design 

information and as-built sketches are included in the DRF. Therefore, the 

team review focused primarily on the technical content of the San Jose DRF. 

Since the conclusion of the inspection, GE-NE has submitted a letter to the 

NRC stating that GIRAFFE was a developmental test conducted in a disciplined, 

professional manner, but not explicitly under the requirements of NQA-1. 

GE-NE further stated that PANDA and PANTHERS are the principal tests to be 

used as design basis tests for the SBWR design certification program and it 

was GE-NE'S intention to use GIRAFFE data to substantiate the results of PANDA 

and PANTHERS at another scale. 

The information in the DRF indicated in general that GIRAFFE was a competently 

performed developmental engineering test program. 

carefully designed, and evidence of considerable dialogue between GE-NE and 

Toshiba was found regarding design characteristics to be included in GIRAFFE. 

In regards to the conduct of the test program,' it appears due attention was 

paid to the acquisition of good data, including characterization of the 

facility. Heat and mass balances were checked for the system during shakedown 

tests. There is some question about measurement of total power to the 

facility during some tests, due to the use of "microheaters" to minimize 

facility heat losses and the uncertainty in the heat input from these heaters. 

The test facil i ty was 

Extensive data from the GIRAFFE test program, both in the form of plots and in 

test reports, are included in the DRF. 

any apparent anomalous behavior. Reports from a resident GE-NE engineer onsite at Toshiba during part of the GIRAFFE program did not indicate any 

problems with the test data (although these reports did raise some questions 

concerning the implementation of a formal QA program to cover the testing). 

There is also substantial evidence in the DRF of information exchange between 

GE-NE and Toshiba during GE-NE'S efforts to analyze the data, which would seem 

to indicate that Toshiba had the technical basis from which to address 

testing-related questions. There appeared to be a clear reluctance on the 

part of Toshiba to release to GE-NE, any information beyond which there was a 

clear "need tu know" by GE-NE. 

analyze the data, and also made it more difficult to review the information in 

the DRF, since some gaps still appear to exist. While any data used in a 

confirmatory fashion would have to be reviewed for applicability by the NRC, 

it appears that the data from this developmental test can provide a useful 

basis for additional verification of computer code models beyond that afforded 

by the design certification data acquired from the PANDA and PANTHERS test 

facilities. 

Examination of this data did not show 

This made it more difficult for GE-NE to 

3.8 TRACG Oualification of GIRAFFE 

The team interviewed the GE-NE engineer in charge of TRACG code qualification 

on how GIRAFFE data was being used in TRACG code qualification. 

discovered that the GIRAFFE input deck used in the TRACG .internal design 

review and NRC code qualification document had been suppl-ied by Toshiba and 

had not been checked by GE-NE. GE-NE did not have the calculation notebook 

The inspector 

-1 1- 

93 
which contains the supporting information for the development of the input 

deck. GE-NE stated that such a check is not required by their engineering 

procedures even if GE-NE does not have the supporting information for the 

development of the input deck. GE-NE performed an independent verification of 

the GIRAFFE input deck because of the findings that were previously identified 

during the GIST inspection at GE-NE in August 1993. 

the independent verification came from the GE-NE GIRAFFE test report 

NEDC-32215P and information gathered from a trip to Toshiba by GE-NE and 

Brookhaven National Laboratory in February 1994. 

The information used in 

The independent veri f i cation reveal ed di screpanci es i n the original input 

deck. 

to the chairman of the GE-NE design review team. 

TRACG DRF. 

conclusions of the adequacy of TRACG in predicting GIRAFFE data. 

also discussed TRACG condensation modelling with GE-NE since this is the most 

important process for long term decay heat removal in the GE-NE SBWR design. 

GE-NE confirmed that the only testing of the TRACG condensation model was the 

GIRAFFE data. The integral GIRAFFE data is not adequate to test a local film 

condensation model that has strong interactions with other code models such as 

interfaci a1 drag, wall drag, and the transport of noncondensi bl e gases. 

Questions about the adequacy of Toshiba modelling and the adequacy of the 

TRACG condensation model were raised upon review of the GIRAFFE DRF T15-00004. 

The GIRAFFE DRF T15-00004 included a well documented historical record of the 

communications that took place between GE-NE and Toshiba on the GIRAFFE test 

program and TRACG modelling. Documentation in the DRF made it clear that the 

GIRAFFE experiment should not be looked at as an SBWR simulator or proof that 

the long term decay heat removal system will work. 

reason the facility pressure started decreasing as early as it did was due to 

environmental heat losses that were greater in the test facility compared to 

what they would be in a perfectly scaled SBWR simulator. The GIRAFFE Phase 1 

testing was also done in a facility that is nonprototypical compared to the 

present SBWR design. 

The GIRAFFE experiment i!r,ed the May-Witt decay heat curve for its power input. 

This is higher than the ANS-79 standard and is the Japanese regulatory 

licensing standard. 

The code qualification test cases were rerun and the results were shown 

Changes in the results were not significant and did not change the 

This was documented in the 

The team 

It stated that the only 

It was designed to represent an earlier SBWR design. 

The DRF also documented the difficulties in modelling GIRAFFE and calculating 

the experimental results. An internal memorandum in the DRF recommends 

against model1 ing the GIRAFFE facility with one TRACG VESSEL component because 

heat losses to the environment could not be modelled properly because of code 

limitations. In this modelling choice the heat losses to the environment were 

modelled by reducing the power input to the reactor core. 

the steam generation rate in the core and reduce the steam flow to the 

containment and the load on the PCCS. 

this modelling practice, this is the model used in TRACG code qualification. 

A letter in the DRF from GE-NE to Toshiba stated that GE-NE could not use the 

Toshiba TRACG GIRAFFE model for TRACG code qualification because it would not 

meet GE-NE’S QA requirements. 

qualification without any explanation. 

This will reduce 

In spite of recommendations against 

This model was later used for TRACG code 

-12- 

94 
The TRACG film condensation model used in the PCCS is the University of 

California at Berkley (UCB) correlation which is the Nusselt correlation with 

multipliers that take into account the degradation due to noncondensible gases 

and enhancement due to shear effects. Initial evaluations of the GIRAFFE data 

showed that Nussel t underpredicted the heat transfer by approximately a factor 

of two. The TRACG code qual ification document states that the UCB correlation 

slightly underpredicts the GIRAFFE PCCS heat transfer. A letter from GE-NE to 

Toshiba as late as January 1992 indicated that TRACG predicted inadequate PCCS 

performance. 

correlation is used in the Toshiba TRAC, the heat removal is overpredicted. 

There is no clear resolution of why and how the UCB correlation was able to 

give acceptable predictions of the GIRAFFE data and why it was accepted for 

use in the code. 

The above questions and concerns in Section 3.8 were discussed during the 

inspection with GE-NE staff and represent the type of issues that would be 

raised by the NRC staff during the review of the TRACG code qualification for 

long term containment heat removal. 

Contradictory to this Toshiba states that when the UCB 

3.9 Corrective Action 

The team reviewed EOP 75-3.00, "Corrective Action and Audits." Corrective 

Action Requests (CARs) are used to document qual ity system problems, to 

request timely response to these problems, and to record objective evidence of 

action tzken as a result. The CAR mechanism is used to assure that root 

causes and preventive actions are identified in response to quality problems. 

EOP 75-6.0, "Quality Assurance Records" specifies that audit reports and CARS 

are kept for 3 years. 

The team reviewed several Corrective Action Requests (CARs) that had been 

generated either during an audit or individually. For example, on June 20, 

1994, a CAR was issued as committed actions related to verification of GIST 

input decks had not been totally completed. Another example involved CAR ARP 

94-3-1 that was issued when a number of DRFS were found non-conforming to the 

pertinent requirements during the course of an internal GE-NE audit. 

committed corrective actions were found appropriate, they included correcting 

the DRFs, re-training personnel on guiding requirements, and review of other 

DRFs for problems. 

The 

In one case however, when QA had identified that the Toshiba quality program 

could not be demonstrated to meet NQA-1 in accordance with the SBWR quality 

program in 1990, no documented action was taken until the 1993 time frame to 

resolve the issue. 

to ensure it was rectified. 

This problem should have been a candidate for using a CAR 

The CAR mechanism was found to assure that appropriate levels of management 

were aware of quality issues and that timely corrective measures were 

instituted and verified. One weakness was identified where an issue was not 

captured in the CAR system to ensure resolution. 

-13- 
3.10 Audits 

The team reviewed EOP 70-11, section 3.13 related to audits. 

required to be performed annually by NQA of each staff level component 

reporting to the Vice President or General Manager. The audits shall comply 

with NQA-1 and ANSI N45.2.12 as applicable. 

suppliers are required every three years. The audits are to be staffed with 

personnel independent of the work being audited. The program shall ensure 

timely corrective action for audit findings. 

Audits are 

External GE-NE audits of 

The following GE-NE internal audits were reviewed: 

Services Oualitv Assurance 

Audit ABQ-88, "Quality Assurance Audit of ABWR Program by Material Services 

Qual i ty of Product Qual i ty Assurance. I' The audit was conducted from September 

21, 1988 to April 19, 1989. The content of DRFs, one of which involved the 

SBWR Gravity Driven Cooling System test program, was reviewed. 

Audit ABQ-90, I'PQA Audit ABQ-90 of the ABWR Programs K6/K7 and SBWR." The 

audit was conducted from October 10, 1990 to November 30, 1990. The audit 

scope examined compliance with the governing SBWR QA program, control of US 

Team member interfaces, indoctrination of SBWR staff, design control , and test 

control. An audit finding was issued for training of SBWR personnel on the 

quality program requirements. 

and GDCS. The corrective act i on measures i ncl uded documented evidence that 

training had been received by the appropriate GE-NE staff. 

The audit also reviewed design files on TRACG 

Audit ABQ-91-3, I'PQA Audit of the SBWR Program and the K6/K7 Program." The 

audit was conducted from December 9, 1991 to March 2, 1992. The audit scope 

included interface controls, design control , .test control , and corrective 

actions from previous audits. An audit finding had been generated related to 

handling of Engineering Review Memoranda (ERMs). 

Advanced Reactor Proqrams 

Audit 493-02, "SBWR Design and Certification." The audit was conducted from 

June 22, 1993, to August 8, 1993. The audit scope included design interfaces, 

implementation of SBWR Team and Organization manual requirements, DRF file 

generation, design reviews, procurement control, and document control. Some 

audi t observat i ons i ncl uded one i nvol vi ng admi ni strati ve control of DRFs . 

Audit 492-03, "ABWR Certification.and SBWR Design and Certification." The 

audit scope included interface controls, design procedure implementation, 

corrective action on prior findings, and SBWR procurement. 

Audit 94-3, "SBWR Design and Certification." The audit was conducted from 

March 29, 1994, to May 19, 1994. The audit scope included DRFs, training 

procurement control, and prior audit finding corrective action. A audit 

finding was generated as multiple DRFs were found nonconforming to 

requirements pertaining to: organization incomplete, missing indices, des 

verification is incomplete, open verifications, corrections made without 

-14- 

96 
traceability. 

considered a positive approach. 

The inclusion of a technical specialist on this audit team is 

The team reviewed related documentation such as: lead auditor qualification 

and cert i f i cat i on records , audit pl ans , audit check1 i sts , and audit exit 

meeting notes. 

The team concluded that GE-NE was implementing an appropriate internal audit 

program. The conduct of GE-NE audits external to the GE-NE organization was 

not examined . 

4 PERSONNEL CONTACTED 

GE Nuclear Enercrv 

Robert H. Buchholz, Manager, SBWR - U.S. Programs Patrick W. Marriott, Manager, Advanced Plant Technologies Ken Brayman, Qual ity Systems Manager, Nuclear Qual ity Assurance (NQA) 

Forrest Hatch, Manager, Nuclear Services & Projects Quality 

Phil ip Novak, Qual i ty Assurance Manager, Advanced Reactor Programs (ARP) Don Kaye, Qual i ty Assurance, SBWR Project 

Norman E. Barclay, Nuclear Service & Projects Quality 

Jay Murray, QA Auaits Manager, NQA 

Terry McIntyre, Project Manager, SBWR Test Operations and Analysis 

Paul F. Billig, Senior Engineer, ARP 

John Leatherman, Manager, SBWR Certification 

Sandra Devl in , Project Manager , SBWR Design Documentation C1 osure 

Gordon Wingate, Engineer, SBWR Design Documentation Closure 

Jim Shaug, Senior Engineer, Safety & Thermal Hydraulic Methods 

Maryann Herzog, Senior Engineer, SBWR 

David Foreman , SBWR Li cens i ng 

Doug Hashi , Project Manager, International Agreements 

Ted Bush, Information Management Systems 

Nuclear Requlatorv Commission 

Richard McIntyre, Team Leader, Vendor Inspection Branch (VIB) 

Billy Rogers, VIB 

Joseph Staudenmeier, Analytical Support Group 

Robert Gramm, Performance & Qual i ty Eva1 uation Branch (PQEB) 

Frederick A1 1 enspach, PQEB 

A1 an Levin, Reactor Systems Branch 

Tim M. Lee, Office of Nuclear Regulatory Research 

Andre Drozd, Containment Systems and Severe Accident Branch 

Son Ninh, Standardization Project Directorate 

-15- 
Toshi ba 

Koichi Taira, SBWR Resident Engineer, 

Ken j i Arai , Nucl ear Engineering Laboratory 

DeDartment of Enerqv 

Kashmira Vi jaiyan, Oakland Field Office 

- EPRI 

Richard Burke, Manager, SBWR Design Certification 

Hans Zimmer, 

Paul Scherrer Institute 

Paul Coddington, PANDA Test Facility 

-16- 

98 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 205550001 

December 14, 1494 

Tennessee Val 1 ey Authority 

ATTN: 

6A Lookout Place 

1101 Market Street 

Chattanooga, TN 37402-2801 

Mr. Oliver D. Kingsley, Jr. 

President, TVA Nuclear and 

Chief Executive Officer 

SUBJECT: INSPECTION OF THE REPLACEMENT ITEMS PROGRAM (RIP) CORRECTIVE 

ACTION PROGRAM (CAP) PLAN AT THE WATTS BAR NUCLEAR PLANT, UNIT 1 

(NRC INSPECTION REPORT NO. 50-390/94-201) 

Dear Mr. Kingsley: 

This letter transmits the report of the U.S. Nuclear Regulatory Commission 

(NRC) inspection, conducted by Robert Pettis, Jr., Larry Campbell , Stephen 

Alexander, of the Special Inspection Branch, and Juan Peralta of the Quality 

Assurance and Maintenance Branch, of the Office of Nuclear Reactor Regulation 

(NRR), and Ronald Gibbs and William Bearden of NRC Region 11. The inspection 

was conducted to review the implementation of the Tenness1.e Valley Authority 

(TVA) Replacement Items Program (RIP) Corrective Action Program (CAP) for the 

Watts Bar Nuclear Plant (WBN), Unit 1. 

The inspection, conducted June 20 through 24, July 5 through 8, and July 18 

through 20, 1994, was performed at the 75-percent completion stage of the RIP 

CAP Plan and was related to activities at the plant site authorized by NRC 

Construction Permit CPPR-91. At the conclusion of the inspection, the 

findings were discussed with those members of your staff identified in the 

enclosed inspection report. 

i dent i f i ed in the report. 

of procedures and representative records, interviews with personnel , and 

observation of activities in progress. 

Areas examined during the inspection are 

The inspection consi sted of se'l ecti ve examinations 

Subsequent to the onsite portion of the inspection, TVA continued to provide 

information for the inspectors to review to help close out several issues that 

remained unresolved at the time of the exit meeting. The:;e reviews occasioned 

several telephone conversations with TVA and the submission of additional 

information in response to questions raised during review of the supplementary 

submittals. 

November 7, 1994. 

appropriate report sections revised accordingly. 

The last of this information was received by the inspectors on 

The review of this material was completed and the 

In general, the inspection team determined that the RIP CAP activities were 

effective and that significant improvements had been made, 

deficiencies regarding program implementation were identified. For example, 

during the review of the Material Improvement Project, numerous instances were 

However, 
Mr. Oliver D. Kingsley, Jr. -2- 

identified where material had been installed without performing a review for 

suitability, since the affected items were not tracked by the RIP data base as 

an outstanding work item. During the inspection, TVA initiated an evaluation 

which determined that 84 of a total of 158 sanitization packages (the 

sanitization process involved reviewing the adequacy of an item's 

documentation), identified for further evaluation under the RIP, had not been 

entered into the RIP data base and therefore, had not been evaluated for their 

intended safety-related applications. 

The inspection team also identified several instances where inadequate 

commercial grade dedication was performed for installed material (i .e., 

circuit breakers and related components). The deficiencies appeared to 

reflect a less than full understanding among some procurement engineers of the 

design basis requirements of the various Watts Bar electrical distribution 

systems. The inspection team also identified the failure to perform adequate 

corrective action for equipment which failed in service and the failure to 

follow TVA procedures. 

During the review of Revision 5 to the RIP CAP Plan, the inspection team 

determined that many elements of it were conducive to general improvement of 

the procurement and material control programs at WBN. However, Revision 5 

itself (as had been the case with Revision 4), as a stand alone document, 

expresses positions (e.g., in Note 2 of the Introduction) that are not 

consistent with the requirements of Appendix B to 10 CFR Part 50 and the staff 

positions stated in NRC Generic Letters 89-02 and 91-05, where the staff 

positions differ from Electric Power Research Institute (EPRI) EPRI NP-5652. 

As written, Revision 5 to the RIP CAP would not ensure that all material, 

particularly commercial grade items, would be adequately rendered suitable or 

verified to be suitable for specific safety-related plant applications under 

all design basis conditions. However, these differences from staff positions 

were not reflected, in general, in the implementing procedures that were 

actually in use for procurement and dedication activities; although che 

procedures did contain some less significant deficiencies as noted n this 

report. 

Plan to reflect the actual procurement practices in-place at WBN. 

Therefore, the staff requests that you revise and docket a RIP CAP 

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

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

No response to this letter is required. 

any enforcement actions resulting from this inspection. 

The NRC Region I1 Office will issue 

100 
Mr. Oliver D. Kingsley, 3r. -3- 

Should you have any questions concerning this inspection, please contact the 

NRR Project Manager Peter Tam at (301) 504-1451 or the inspection team leader 

Robert Pettis at (301) 504-3214. 

S i nhe re 1 y , 

' b4rhctor Division of Reacto rojects . 1/11 

Off ice of Nuclear Reactor Regul ati on 

Enclosure: 

Inspection Report 50-390/94-201 

cf': See next page 
Tennessee Val 1 ey Authority 

cc w/encl s : 

Mr. Craven Crowel 1 , Chai rman 

Tennessee Val 1 ey Authority 

ET 12A 

400 West Summit Hill Drive 

Knoxvi 11 e, TN 37902 

Mr. W. H. Kennoy, Director 

Tennessee Val 1 ey Authority 

ET 12A 

400 West Summit Hill Drive 

Knoxville, TN 37902 

Mr. Johnny H. Hayes, Director 

Tennessee Val 1 ey Authority 

ET 12A 

400 West Summit Hill Drive 

Knoxville, TN 37902 

Dr. Mark 0. Medford, Vice President 

Engineering & Technical Services 

Tennessee Val 1 ey Authority 

3B Lookout Place 

1101 Market Street 

Chattanooga, TN 37402-2801 

Mr. D. E. Nunn, Vice President 

New P1 ant Compl et i on 

Tennessee Val 1 ey Authority 

3B Lookout Place 

1101 Market Street 

Chattanooga, TN 37402-2801 

Mr. J. A. Scalice, Site Vice President 

Watts Bar Nuclear P1 ant 

Tennessee Val 1 ey Authority 

P.3. Box 2000 

Spring City, TN 37381 

General Counsel 

Tennessee Val 1 ey Authority 

ET 11H 

400 West Summit Hill Drive 

Knoxville, TN 37902 

Mr. 2. W. Huston, Manager 

Nucl ear Licensing and 

Regulatory Affairs 

Tennessee Val 1 ey Authority 

46 Blue Ridge 

1101 Market Street 

Chattanooga, TN 37402-2801 

Watts Bar Nuclear Plant, Unit 1 

Mr. B. S. Schofield 

Site Licensing Manager 

Watts Bar Nuclear Plant 

Tennessee Valley Authority 

P.O. Box 2000 

Spring City, TN 37381 

TVA Representative 

Tennessee Val 1 ey Authority 

11921 Rockville Pike 

Suite 402 

Rockville, MD 20852 

Regional Administrator 

U. S. Nuclear Regul atory Commi ssion 

Region I1 

101 Marietta Street, NW., Suite 2900 

Atlanta, GA .iU323 

Senior Resident Inspector 

Watts Bar Nuclear Plant 

U.S. Nuclear Regulatory Commission 

Route 2, Box 700 

Spring City, TN 37381 

The Honorable Robert Ai kman 

County Execut i ve 

Rhea County Courthouse 

Dayton, TN 37321 

The Honorable Gar1 and Lanksford 

County Executive 

Meigs County Courthouse 

Decatur, TN 37322 

Mr. Michael H. Mobley, Director 

Division of Radiological Health 

3rd Floor, L and C Annex 

401 Church Street 

Nashville, TN 37243-1532 

Ms. Danielle Droitsch 

Energy Project 

The Foundation for 

P.O. Box 1101 

Knoxville, TN 37901 

G1 obal Sustainabi 1 i ty 

Mr. Bill Harris 

Route 1, Box 26 

Ten Mile, TN 37880 

102 
Enclosure 

U. S . NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF TECHNICAL SUPPORT 

INSPECT ION 50-390/94-201 

REPORT NO.: 

CONSTRUCTION CPPR-91 

PERMIT NO.: 

FACILITY NAME: Watts Bar Nuclear Plant, Unit 1 

INSPECTION June 20 through 24, 1994 

CONDUCTED : July 5 through 8, 1994 

July 18 through 20, 1994 

INSPECTION TEAM: Robert L. Pettis, Jr., P.E., NRR 

Stephen A1 exander, NRR 

Larry Campbell , NRR 

Juan Peralta, NRR 

Ronald Gibbs, Region I1 

William Bearden, Region I1 

PREPARED BY: 

Date 

Vendor Inspecti on Section 

Special Inspection Branch 

Division of Technical Support 

Office of Nucl ear Reactor Regul ati on 

APPROVED BY: 

Robert Gallo, Chief Date 

Special Inspection Branch 

Division of Technical Support 

Office of Nuclear Reactor Regul ati on 
TABLE OF CONTENTS 

EXECUTIVE SUMMARY ......................................................... i 

1 INTRODUCTION ........................................................... 1 

2 INSPECTION OBJECTIVES AND SCOPE ........................................ 2 

3 PROGRAM REVIEW ......................................................... 3 

4 CURRENT AND FUTURE PROCUREMENTS ........................................ 4 

4.1 Procurement Package Review ......................................... 5 

5 PLANT INSTALLED ITEMS VIA PREVIOUS MAINTENANCE ACTIVITIES 

REPLACEMENT ITEMS PROGRAM (RIP) ........................................ 17 

5.1 QA Level I PPSP Review ............................................ 18 

5.2 QA Level 11 PPSP Review ........................................... 18 

5.3 QA Level I11 PPSP Review .......................................... 23 

5.4 Review of Generic Dedication Packages ............................. 28 

5.5 Control of Product Contaminants .................................... 28 

5.6 Replacement Items Installed by Previous Maintenance Activities .... 29 

5.6.1 Maintenance History Data Base .............................. 29 

5.6.2 Problem TIICs .............................................. 30 

5.7 RIP Hold Items .................................................... 31 

5.8 IMPELL Seismically Sensitive Electrically Active Items Review ..... 34 

5.9 RIP QA Level I11 Item Review ...................................... 36 

6 REPLACEMENT ITEMS INSTALLED BY PREVIOUS CONSTRUCTION ACTIVITIES 

REPLACEMENT ITEMS PROGRAM (RIP) ........................................ 37 

7 CURRENT WAREHOUSE INVENTORY-MATERIAL IMPROVEMENT PROJECT (MIP) ......... 38 

7.1 Implementation Review of the MIP .................................. 38 

7.1.1 Acceptable Sanitization Packages ........................... 39 

7.1.2 Unacceptable Sanitization Packages ......................... 40 

8 EXIT MEETING ........................................................... 42 

APPENDIX A . LIST OF DEFICIENCIES ......................................... A-1 

APPENDIX B . LIST OF UNRESOLVED ITEMS ..................................... B-1 

APPENDIX C . LIST OF INSPECTOR FOLLOWUP ITEMS ............................. C-1 

APPENDIX D . GLOSSARY OF SPECIFIC TERMS ................................... D-1 

104 
EXECUTIVE SUMMARY 

The Nuclear Regulatory Commission (NRC) inspection of the Tennessee Valley 

Authority's (TVA's) Rep1 acement Items Program (RIP) piece parts Corrective 

Action Program (CAP) Plan for the Watts Bar Nuclear Plant (WBN) inspection, 

led by the Vendor Inspection Branch (Now the Vendor Inspection Section of the 

Special Inspection Branch) of the Office of Nuclear Reactor Regulation (NRR) , was conducted from June 20-24; July 5-8; and July 18-20, 1994. The inspection 

team consisted of staff members from NRR and Region 11. 

The objective of the inspection was to review the implementation of Revision 5 

to the TVA RIP CAP Program to determine its compliance with Appendix B to 

10 CFR Part 50 (Appendix B), and related TVA commitments. The inspection team 

determined that many elements of Revision 5 of the RIP CAP were conducive to 

general improvement of the procurement and material control programs at WBN. 

However, Revision 5 itself (as had been the case with Revision 4), as a stand 

alone document, expresses positions (e.g., in Note 2 of the Introduction) that 

are not consistent with the requirements of Appendix By and that take 

exception to certain key provisions of NRC GLs 89-02 and 91-05, where they 

differ from Electric Power Research Institute (EPRI) EPRl NP-5652," Guideline 

fr- the Utilization of Commercial Grade Items in Nuclear Safety Related 

Applications (NCIG-07) .I' 

As written, Revision 5 to the RIP CAP would not ensure that all material, 

part i cul arly commerci a1 grade i terns, would be adequately rendered sui table or 

verified to be suitable for specific safety-related plant applications under 

all design basis conditions. However, the positions expressed.in Revision 5 

were not reflected in the implementing procedures that were in use for 

procurement and dedication activities; although the procedures did contain 

some relatively less significant deficiencies as noted in the report. 

In addition, the team identified some deficiencies in the implementation of 

these program procedures. The findings i ncl uded inadequate commerci a1 grade 

dedication of items installed in safety-related applications, such as General 

Electric (GE) 125-Vdc trip coils and spring charging motors for GE Magne-Blast 

circuit breakers used in Class 1E 6900V shutdown boards, Westinghouse direct 

trip actuators used in 480-Vac air circuit breakers and the lack of 

verification of the procurement requirements of Telemechanique starters. 

Other deficiencies included implementation weaknesses in TVA's corrective 

action program, the lack of evaluation of previously installed material not 

contained in the RIP data base, and procedures used which were not appropriate 

to the circumstances. 

Many of the deficiencies identified by the inspection team were attributed'to 

incomplete or ineffective program implementation or to various procedural 

weaknesses. In the exit meeting conducted on July 20, 1994, the inspection 

team discussed the inspection findings with TVA's senior management. 

the rzsults of this inspection, it was determined that, in general, the RIP 

CAP plan activities appeared to be effective, however continued management 

attention is required to assure proper program implementation. 

Based on 

i 

105 
1 INTRODUCTION 

The Tennessee Valley Authority (TVA) began the Watts Bar Nuclear Plant (WBN) 

Replacement Items Program (RIP) Corrective Action Program (CAP) in 1988 to 

resol ve deficiencies with the procurement of safety-re1 ated repl acement i tems 

procured without TVA engineering review and to address TVA employee concerns 

and NRC audit findings from the Sequoyah Nuclear Plant. 

objective of the RIP CAP is to ensure that replacement piece parts, especially 

commerci a1 grade i tems (CGIs) instal 1 ed in safety-re1 ated basic components, 

are consistent with previously validated environmental or seismic 

qualifications, and with the component's ability to perform its intended 

safety function. 

TVA's primary 

The scope of the RIP CAP covers commercial grade replacement item procurements 

for safety-re1 ated applications and procurements of repl acement items for 

environmentally qualified (10 CFR 50.49) components. 

CAP plan inspection are Appendix B to 10 CFR Part 50 (Appendix B), 10 CFR 

50.49, Revision 5 to TVA RIP CAP of March 1, 1994, and NRC Inspection 

Procedure (IP) 38703, "Commercial Grade Dedication." 

The bases for this RIP 

The RIP CAP addressed a lack of engineering involvement in the procurement 

process, which resulted in unreviewed and potenti a1 ly unacceptable spare or 

replacement parts being installed in previously qualified equipment in the 

plant. Revision 5 of the RIP CAP addresses the following four distinct areas: 

0 Current and Future Procurements: These procurements are made through 

TVA's current program, which requires engineering involvement in all new 

procurements. TVA's Procurement Engineering Group (PEG) does an 

engineering review of the purchases of all safety-related equipment. 

The PEG reports directly to Nuclear Engineering and is connected to 

Materials and Procurement in the organizational matrix. 

0 Plant Installed Items from Previous Maintenance Activities: This 

activity of the RIP CAP involves a historical review (for technical 

adequacy) of spare or replacement parts installed from the power stores 

inventory for work controlled by maintenance work requests or work 

packages. Installed material is identified by a TVA Item Identification 

Code (TIIC) number and is controlled by TVA Nuclear Stores requisition 

Form 575N. 

0 ReDlacement Items Installed bv Previous Construction Activities: This 

activity of the RIP CAP involves a historical review (for technical 

adequacy) of the replacement of parts installed from the construction 

warehouse inventory by personnel from the construction group. 

0 Current Warehouse Inventorv: This activity of the RIP CAP is commonly 

referred to as the Material Improvement Project (MIP). 

initiated as a result of a 1991 TVA quality assurance (QA) assessment 

which identified weaknesses in commercial grade dedication. This 

activity is designed to strengthen the Quality Release Program by 

"sanitizing" all old inventory prior to release for installation to the 

plant. 

The MIP was 

The sanitization process includes an engineering review for 

1 

106 
technical adequacy, verification of the item, QA receipt inspection, and 

storage. 

TVA implemented the MIP on June 5, 1991, after which replacement items, in inventory as of that date, must have a MIP sanitization package 

prepared and approved before the item can be issued to the plant. 

issued prior to June 5, 1991, are to be evaluated and resolved by the 

historical reviews discussed in the RIP portion of the inspection. 

Items 

In accordance with the NRC letters to TVA regarding the RIP CAP (particularly 

Revision 5), the team reserved its judgement of the RIP CAP itself until it 

had completed a substantial portion of the review of the program procedures 

established under the RIP CAP and their implementation. As a result of that 

review conducted during this inspection, the team's assessment of the RIP CAP 

as written is consistent with the reservations expressed in the NRC letters to 

TVA regarding the RIP CAP. The team concluded that as a stand-alone document, 

presumed to describe and generally prescribe the practices employed by TVA in 

procurement and commercial grade dedication, RIP CAP, Revision 5, expresses 

positions that are not consistent with the requirements of Appendix By and 

that take exception to certain key provisions of GL 89-02 and GL 91-05 where 

they differ from EPRI NP-5652. 

However, the team's assessment of the actual WBN program, procedures and 

implementation, was that with the few exceptions noted, and contrary to some 

of the positions stated in the RIP CAP and associated correspondence, the 

program procedures appeared on the basis of the review within the scope of 

this inspection, to be more consistent with the applicable requirements of 

Appendix By and generally consistent with the staff positions as promulgated 

in Generic Letters 89-02 and 91-05. 

of certain programmatic deficiencies discussed below and the identified 

classification and dedication deficiencies. 

The exceptions mentioned above consisted 

2 INSPECTION OBJECTIVES AND SCOPE 

The NRC conducted this inspection in accordance with NRC IP 38703, Appendix By 

and Revision 5 to the TVA RIP CAP. The objective of this inspection was to 

review TVA's implementation of the RIP CAP to determine its compliance with 

Appendix B and related TVA commitments. 

The inspection team has characterized its findings within this report as 

deficiencies, unresolved items, or inspector followup items. Deficiencies are 

either a) the apparent failure of the licensee to comply with a requirement or 

b) the apparent failure of the licensee to satisfy a written commitment or to 

conform to the provisions of applicable codes, standards, guides or accepted 

industry practices when the commitment has not been made a legally binding 

requirement. Unresolved items involve a concern about which more information 

is required to ascertain whether it is acceptable or deficient. 

deficiency identified during the inspection, and designated in the body of the 

report, is ljsted in Appendix A. Unresolved items and inspector followup 

items are listed in Appendix B and Appendix Cy respectively. 

TVA-WBN specific terms is included in Appendix D. 

Each 

A glossary of 

2 

107 
3 PROGRAM REVIEW (TI 2512/27, 35065, 35746, 35747) 

Before and during the inspection, the inspection team conducted a review of 

the WBN site procedures which define material procurement, receipt inspection, 

storage and handling, and material issue and control. This review included 

procedures which define the current program, as well as the RIP and MIP, and 

included procurement, receipt inspection, storage and handling. The following 

procedures were included in this review: 

0 

0 

0 

m 

a 

a 

a 

0 

0 

a 

m 

0 

Site Standard Practice (SSP) SSP-10.01, "Procurement of Materials and 

Services," Revision 10 

SSP-10.02, "Material Receipt and Inspection," Revision 18 

SSP-10.03, "Material Handling, Storing, and Shipping," Revision 16 

SSP-10.04, "Material Issue, Control , and Return," Revision 20 

SSP-10.05, "Technical Evaluation for Procurement of Materials and 

Services," Revision 10 

SSP-10.06, "Materials Marking and Identification," Revision 2 

SSP- 10. By "Materi a1 s Improvement Project , I' Revi si on 5 

SSP-lO.C, "Evaluation of Installed Safety-Related Replacement Items," 

Revision 1 

Site Administrative Instruction (SAI) SAI-lO.B, "Haterials Improvement 

Project Sanitization process," Revision 1 

Quality Assurance Instruction (QAI) QAI-10.03, "Material 

Sanitization QA Program," Revision 4 

QAI-10.04, "Material Receipt Inspecxion," Revision 11 

QAI-10.05, "PPSP-Review and Inspection," Revision 1 

QAI-10.06, "Inspection of Storage Facilities," Revision 3 

The team verified that procedures were established and implemented for key 

parts of the material control program. 

attributes of the requirements: 

The team verified the following 

inclusion of technical and QA requirements in procurement documents, 

including environmental qualification and seismic requirements, testing 

and inspection requirements, documentation requirements, TVA access to 

vendors facilities and records for audit purposes, and the basis for 

quality classification of materials. 

control of procurement documents to ensure incorporation of engineering 

design requirements, as well as, engineering review and approval of the 

purchases and changes during the procurement process. 

inclusion of 10 CFR Part 21 requirements, including applicant dedication 

of CGIs where appropriate. 

establ ishment, maintenance, and use of an approved suppliers list (ASL) 

including the basis for qualification of suppliers such as by audit and 

source inspection. 

request for receipt and review of vendor documentation, including 

Certificates of Compliance/Conformance (COCs) and Certified Material 

Test Reports (CMTRs). 

3 

108 
e 

e 

e 

performance of recei pt inspection of materi a1 s i ncl udi ng veri f i cati on of 

the adequacy of vendor documentation, vendor compl iance with purchase 

order requirements , and performance of speci a1 inspections specified by 

engineering. The team also reviewed the receipt inspection facilities, 

tools and equipment used by quality control (QC) inspectors, and receipt 

inspection records which include programs for identification and 

resol uti on of recei pt inspection deficiencies , and the control of 

conditional re1 eases for materi a1 s not yet i nsperted . 

means to separate acceptable material from material not yet accepted, 

and from nonconforming material (i .e. , tagging, identification, 

segregat i on , and . storage) . 

identification and implementation of storage requirements and storage 

1 eve1 s , i ncl udi ng segregation of non-conformi ng materi a1 s , and 

requirements for inspection of storage areas by TVA QC personnel. 

The inspection team inspected the storage areas for proper storage of 

materi a1 s and verified the adequacy of storage conditions, access controls, 

identification and retrievabil ity of stored materials, and segregation of 

nor-xforming materials. Except for the procedural deficiencies identified 

wi in SSP-10. C (Section 5.1 of this report) , the procedures reviewed appeared 

satisfactory in implementing the RIP CAP. 

4 CURRENT AND FUTURE PROCUREMENTS (TI 2512/27, 35065, 35746, 35747) 

The requirements for the preparation of TVA's current procurement practice and 

processes are prescribed in SSP-10.01, Revision 10, "Procurement of Materials 

and Services" and in SSP-10.05, Revision 10, "Technical Evaluation for 

Procurement of Materials and Services." These procurements are made to TVA's 

current program which requires engineering involvement by 'TVA's PEG for all 

new safety-related procurements. 

In a letter of April 5, 1994, the NRC staff documented its review of 

Revision 5 of the WBN RIP CAP plan of March 15, 1994, and !stated that it could 

not substantiate the acceptability of the commercial grade dedication portion 

of the program. Accordingly, the NRC stated that an inspection would be 

performed to determine the acceptabil i ty of control s and procedures governing 

commercial grade dedication activities at WBN. TVA's decision to take 

exception to the guidance in Generic Letter (GL) 89-02 (in Note 2 of the 

Introduction) was one of the reasons for the staff's decision not to reach a 

conclusion on the acceptability of the RIP CAP program unt?l an inspection of 

its implementation was performed. In a letter to TVA on July 27, 1992, the 

staff stated that since TVA takes exception to the GL and does not state an 

acceptable alternate position, the staff cannot conclude that the RIP CAP plan 

is acceptable for commercial grade dedication. 

staff recommends that TVA revise the RIP CAP plan to reflect the actual 

procurement practices in-place at WBN. 

As stated previously, the 

4 

109 
4.1 Procurement Packaqe Review 

Prior to the on-site inspection, the NRC inspection team requested that TVA 

provide lists of PEG packages addressing all TVA QA Level I materials 

(Appendix By safety-related); all TVA QA Level I1 materials (dedicated CGIs); 

and all TVA QA Level I11 materials (materials used in safety-related systems 

that do not have a safety function), which have been purchased since June 5, 

1991, the inception of TVA's new procurement program. 

The team selected a sample of PEG packages from these lists for review and 

organized the packages by discipline into electrical and instrumentation, 

mechanical , and materi a1 s (incl udi ng 1 ubri cants). The PEG packages selected 

were to contain the appropriate purchasing, engineering, and receipt 

inspection documentation. TVA provided associated commerci a1 grade survey 

reports in those instances when it used acceptance method 2, "Commerci al-Grade 

Survey of Supplier," of Electric Power Research Institute (EPRI) NP-5652, 

"Guideline for the Utilization of Commercial-Grade Items in Nuclear SafetyRelated Applications (NCIG-07) ." 

PEG packages for review: 

The inspection team selected the following 

QA Lwel I 

PEG Package 9400023342 (RIMS No. T49940401828), Revision 0, April 1, 

1994, for an ITT Barton transmitter, part number 764. 

PEG Package 9400018340 (RIMS No. T49940311858) , Revision 0, March 11 , 1994, for Class 1E motor inspection/repair services. 

PEG Package 4730-1-01, Revision 8 (purchase order (PO) 93N3F-41936D-01, 

RD 363172 and RD 1025437), for metallic pipe. 

PEG Package 9300002498 (RIMS No. T4992113835) , November 23, 1993, for a 

90 degree, 1. 5-inchY 3000 pound, pipe el bow suppl ied in accordance with 

the requirements of Section 111, Class 2, of the American Society of 

Mechanical Engineers (ASME) Boi 1 er and Pressure Vessel Code. 

PEG Package 9300006087 (RIMS No.49930120800), for TVA PO No. P-93PJX36732H-000, April 22, 1993, for various shapes and sizes of structural 

steel products purchased from Mid-South Nuclear, Incorporated. 

PEG Package 9400020408 (RIMS T49940317808), March 17, 1994, for ASCO 

sol enoid Val ves. 

PEG Package 9400018339 (RIMS T49940312803) , March 11, 1994, for 

General Electric (GE) power supply. 

PEG Package 9308024 (RIMS No. T49930924807) , September 23, 1993, for 

various WBN System 31 instruments. 

PEG Package 9400002182, March 31, 1994, for various types and 

quantities of molded-case circuit breakers, manufactured by United 

Control s , Incorporated, used for both safety-re1 ated and nonsafety5 

110 
re1 ated appl i cati ons . 

QA Level I1 

a PEG Package 9300016028 (RIMS No. T49940312824) , Revision 0, March 12, 

1994, for an ITT Barton flow switch, part number 288A. 

a PEG Package 9400018459 (RIMS No. T49940309810), Revision 0, March 9, 

1994 for Amphenol connector, part number MS 3106A-14S-7S. 

a PEG Package 922236 (RIMS No. T49930415858), Revision 0, April 14, 

1994, for a Sorrento Electronics, Incorporated, flow meter part number 

02815067-001. 

PEG Package 9400012490 (RIMS No. T49940203922) , Revision 0, February 3, 

1994, for Mal 1 ory capacitors. 

a 

a PEG Package 9300009828 (RIMS No. T49930318808), March 4, 1993, for a 

safety re1 i ef val ve. 

a PEG Package AND037P (RIMS B26910820640), August 1990, for a 

rotary oil pump for circulating oil through the control room chiller 

compressors. 

a PEG Package 9400004823 (RIMS T49940202969), February 2, 1994, for a 

Gould-Shawmut fuse. 

PEG Package 912816 (RIMS T49920131812), January 31, 1992, for a Foxboro 

saturabl e reactor. 

a 

a PEG Package 9300000184 (RIMS T49940302805), March 1, 1994, for a GouldShawmut fuse holder (B1 ock) . 

PEG Package 920013, June 9, 1992, for 125-Volt dc (Vdc) closing spring 

charging motors, part number 0105C9393P002, for GE model AM-7.2-500-6HD 

Magne-Blast circuit breakers used in 6.9-kVY Class IE service. 

a 

QA Level I11 

a PEG Package 9400023282 (RIMS No. T49940402832), Revision 0, April 2, 

1994, for Sorrento Electronics, Incorporated, read-out module, part 

number 03573000-004. 

a PEG Package 9400005161 (RIMS No. T49940409852) , Revision 0, April 8, 

1994, for a GE transmitter, part number 50-555-111BDAA4WBE. 

PEG Package 9400021420 (RIMS No. T49940322854), Revision 0, March 22, 

1994, for an ASCO solenoid valve, part number 8316654E. 

a 

a PEG Package 9400011803, Revision 0, (PO P-90NJC-44792B-001 , Re1 ease No. 

1010689) for a ball valve. 

6 

111 
PEG Package 9400013143 (RIMS No. T49940212841), February 12, 

1994, for valve parts for gate valves. 

PEG Package 9400012190 (RIMS No. T49940201847), February 1, 1994, for a 

stem assembly with disc for a 2-inch angle valve. 

PEG Package 9400014204 (RIMS No. T49940216841), February 16, 

PEG Package 4821302, May 22, 1992, for non-critical Limitorque 

1994, for a machined shaft sleeve key for safety injection pumps. 

parts. 

mechanism that allows manual or electric operation of the actuator. 

The part in question is a metallic clutch ring in the clutch 

The team reviewed the information contained in the PEG packages in order to 

establish that technical and quality requirements had been adequately 

trans1 ated into the appropriate procurements documents, to verify that the 

bases for the safety classification of the structures, systems, and components 

could be traced to the corresponding design bases documents, and to confirm 

that dedication activities conformed to the applicable regulatory criteria and 

guidance 

information of sufficient detail to support the conclusions stated. 

confirmed that the PEG packages complied with TVA procedures and the 

appropriate regulatory requirements except as discussed below. 

The inspection team found that certain packages did not include 

The team 

0 PEG Packaqe 9400018459 - This QA Level I1 package, March 9, 1994 (RIMS 

No. T49940309810), documented TVA's dedication of a standard, straight 

plug she1 1 , three-contact , Amphenol connector (part number MS 3106A-14s7s). 

stored in TIIC AXB4794 which had been identified as a suitable 

rep1 acement for unique identifier (UNID) 1-SM-046-0057B-S. The PEG 

package described the "Host Equipment Description/UNID(S)" as "Various 

Class 1E and Non-1E Instruments and Devices, UNID 1-SM-046-0057B-S 

Typical , 'I and the "Host Equipment Safety function( s) It was documented as 

"Provide Signal .I1 

The "Critical Characteristics For Item Functions" portion of the 

Technical Evaluation Form (TEF) identified the following as critical 

characteristics for this item: part number, as stated; 

configuration, vendor catalog pin location, nominal dimensions are 

acceptable; the presence of the Amphenol logo on a blue insert; and 

insulation resistance, short circuit across pins, open circuit pin 

and case. All critical characteristics were to be verified during 

receipt inspection and the bases for accepting these critical 

characteristics in establishing the ability of the connector (or 

according to the procurement package, the "Host Equipment") to 

"Provide Signal" were identified as "Contract," "Catalog Excerpts," 

and "Engineering Judgement, I' respectively. 

The package documented the evaluation of indeterminate material 

The manufacturer's published catalog referenced in the package 

stated that standard Mil-Spec contacts are silver plated and have 

pre-tinned solder pockets, the 97 prefix indicates an Amphenol 

design, and they may have special purpose contacts, variations from 

7 

112 
Mil-Spec requirements, or a.feature which has been assigned a MilSpec. 

considered by the engineer when he concluded, on page 6 of 22 of the 

PEG package, that part number MS 3106A-14S-7S is the same as 

973106A-14S-7SY except it’s built to a Mil-Spec, and that both are 

acceptable. The basis for this conclusion was not documented in the 

package, nor was there documentation to support the conclusion that 

the connector was a “1 i ke-for-1 i ke” rep1 acement. 

However, these differences in design appear not to have been 

Furthermore, since EPRI NP-5652, acceptance Method 2, was not part 

of the dedication, TVA could not assume that the connector(s) 

conformed to the design and performance and fabrication requirements 

of a given Mil-Spec as the manufacturer’s pub1 ished catalog stated. 

The inspection team concluded that the procurement package included 

no information on (1) the frequency and amplitude of the signal 

expected to be carried through the connector, (2) how the value of 

the impedance measured across the contacts or pins correlated to the 

transmission ability of the connector, (3) the function performed by 

the connector within the system (later identified as the EG-R 

actuator in the Woodward governor control circuit of the turbine 

driven auxiliary feedwater pump), (4) the design bases documents or 

sources that were reviewed or consulted to ensure that the design 

integrity of the host system was not degraded below an acceptable 

level and, (5) the basis for the engineer’s determination that the 

part number differences were acceptable. 

During the inspection, the inspection team asked TVA to clarify the 

basis for the engineer’s conclusions that a resistance test yielding 

a short circuit across the connector’s pins would demonstrate the 

abil i ty of the connector to carry the required safety-related 

signal , and that the Amphenol connector being replaced was a “1 i kefor-like“ replacement of the original. TVA agreed to amend the 

package to resolve these issues. 

designated as Inspector Followup Item 94-201-11. 

As a result, this item is 

PEG Packase 9400018340 - This QA Level I package (RIMS No. 

T49940311858), Revision 0, March 11, 1994 (which supersedes RIMS No. 

T49911127964) , documents TVA’s procurement of refurbishment services 

from TVA’s Power Service Shops (PSS), Muscle Shoals facility, for a GE 

250-horsepowerY 460-Vol t-ac (Vac) , 3-PhaseY C1 ass 1E Motor used in the 

main control room chiller package A-A. Page 2 of 4 of the TEF stated 

that the motor shown on the attached form was reworked by PSS on shop 

order 92-7609 and that the PEG originating document for transfer of the 

motor was RIMS No. T49911127964. It was also stated that the motor, 

UNID 0-MTR-31-80/2-AY failed after installation and initial run tests. 

This PEG package returned the subject motor to PSS for disassembly and 

examination to determine the cause of motor failure and to initiate any 

needed repairs. 

Examination of RIMS No. T49911127964 (PEG Package No. 920953, 

November 27, 1991), revealed that the technical and quality requirements 

for this service had been established and documented under RIMS No. 

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113 
B26891031760 (Nuclear Engineering - Contract Engineering Group, EPG 

Package 8903107E, October 31, 1989). Attachment 1 to PEG Package 920953 

states, The motor specified for transfer to Muscle Shoals for rewind 

(see attached 2652) is identical to the motor transferred for rewind by 

PEG Package B26891031760 (which was supplemented by B26901207779). All 

characteristics and specifications contained in 826891031760 and 

826901207779 still apply. Page 12 of 20 of EPG Package 8903107E 

documents the analysis upon which the engineer established how seismic 

qualification of the refurbished motor would be maintained and listed 

motor performance acceptance criteria which, once confirmed by testing 

at Muscle Shoals, would confirm that the rewound motor 'meets NEMA 

Design B performance requirements. The documented results of the tests 

would be evaluated before receipt inspection at WBN. 

The inspection team could not identify from the information provided by 

TVA, that the test and weight data required to be furnished by Muscle 

Shoals in accordance with EPG Package 8903107E had been "evaluated prior 

to receipt inspection" by the responsible organization. Additionally, 

it could not be verified that once the motor had prematurely failed 

after installation and initial run tests that an examination or 

evaluation had been performed by the responsible organization to 

determine the cause of motor failure and to initiate corrective actions 

accordi ngl y . 

During the inspection, Problem Evaluation Report (PER) WBNPER940354, 

Revision 0, July 14, 1994, was initiated by TVA to address these 

deficiencies. TVA's failure to review test results that were part- of 

safety-related activities intended to assure that design integrity of a 

Class 1E motor was maintained; provide documentation to support that 

appropriate corrective actions had been taken once a failure occurred 

shortly after post-installation tests; and to demonstrate suitability of 

appJication, was identified as Deficiency 94-201-D1. 

0 PEG Packaqe 9400002182 - Under Contract No. P94N3F-102173-000, TVA 

procured for WBN from United Controls Incorporated (UCI) several types 

and quantities of molded-case circuit breakers (MCCBs), some for 

nonsafety-related applications, designated QA Level 111, and some for 

safety-re1 ated applications designated QA Level I, which were all 

evaluated under different parts of this PEG package. 

The QA Level TI1 portion of the package covered contract Item 1 (three 

Westinghouse LA3400PR MCCBs, TIIC BTN245B), Item 3 (two Westinghouse 

JA3175W MCCBs, TIIC BVA124F) and Item 4 (one Westinghouse EHB2060 MCCB, 

TIIC BTW598A). In UCI COC for shop order lOQX002645, dated January 19, 

1994, UCI stated that the testing was done in accordance with UCI's 

nuclear qualification test procedure NQTP-1816, which was to have been 

prepared per the TVA PO based on TVA/WBN Maintenance Instruction (MI) 

57.27, Revision 11. Upon review of this portion of the PEG package, 

deficiencies in the apparent test methods and documentation provided 

with a COC from UCI were identified. 

concerned were not installed, nor intended for safety-related service, 

and as verified by warehouse inspection were appropriately tagged, the 

However, because the items 

9 

114 
inspection team reviewed only those items on this PEG package (and 

contract) that were intended and tagged for safety-related applications. 

Item 2.(one ITE/Siemens FXD63B100, TIIC BTK256M), and Items 5 through 8 

(two ITE/Siemens ED63A050, TIIC BRH8283; six Square D FA26020AC, TIIC 

BWF356Q; three ITE/ Siemens ED63A025, TIIC BTK204A; and three 

ITE/Siemens ED63A010 MCCBs, TIIC BVB013A) of TVA Contract P94N3F-102173- 

000, were all designated QA Level I and tested by /IC1 using the same 

procedure as the QA Level I11 MCCBs according to the associated UCI COCs 

(same shop order, but dated December 30, 1993) provided for these 

contract i tems. 

During review of the UCI data sheets in the package (marked NQTP1816.1.2) , the TVA-approved UCI test procedure NQTP-1816.1.1, Revision 

4, dated February 9, 1993, and the TVA/WBN test specification, MI-57.27, 

on which the NQTP was based, deficiencies were noted that indicated that 

MI-57.27 did not adequately translate design basis requirements for the 

intended MCCB applications into design output documents. As a result, a 

proper review for suitability of application would not have been 

performed. 

The following are examples of inadequate translation of design basis 

requirements into design output documents in that the TVA test 

specification (MI-57.27, Rev. 11) on which the UCI test procedure was 

based would not be adequate to verify that the MCCBs would perform their 

safety functions under all design basis conditions: 

UCI's test procedure called for testing shunt trips at 120 Vac, but 

did not address 125 Vdc or degraded voltage conditions (as low as 90 

Vdc at some plants) and equalizer battery charge voltage (as high as 

140 Vdc). TVA stated in response that "no degraded voltage should 

exist when the shunt trip is used." 

inspection, TVA did not provide documented objective evidence to 

demonstrate that no shunt trips in MCCBs tested under the UCI test 

procedure (or any others in the plant tested in a similar manner) 

would ever be required to operate to perform their safety function 

under degraded (or excess) voltage conditions. 

However, during this 

Manufacturers' time-current characteristic curves for the inversetime thermal overload trip response of the breakers under test were 

not identified, nor were curves from TVA/WBN's TI-108 curve book as 

required by MI-57.27. Therefore, the source of the acceptable trip 

times for the thermal overload trip tests listed on the UCI data 

sheets was not documented. This was a concern particularly with 

ITE/Siemens breakers because of past problems with this manufacturer 

as discussed in NRC Information Notice 85-16, "Time/Current Curve 

Di screpancy of ITE/ Si emens-A1 1 i s Molded Case Circuit Breaker I' 

dated February 27, 1985, and in NRC Information Notice 89-21, 

"Changes in Performance Characteristics of Molded Case Circuit 

Breakers," February 27, 1989. TVA is verifying that proper curves 

were used for the WBN design basis and will document the results. 

10 

115 
0 The full-load hold-in test of 30-minute duration was not consistent 

with industry standards (e.g., NEMA AB 4-1991 or UL-489) or practice 

in which MCCBs would normally be loaded 1 or 2 hours, depending on 

rating. TVA stated that the 30-minute test was reasonable based on 

the existing loads relative to the continuous current rating of the 

breaker or the nominal rating of its overload trip unit. However, 

there was no documented analysis to show adequate margin between 

breaker ratings and existing loads, accounting for worst-case 

service conditions, nor was the question of future additional loads 

addressed. 

Step 1.2.6, the instantaneous magnetic trip test instructions, of 

the test procedure stated that test current values are "typically" 

1.5 times the trip setting of an adjustable magnetic-only MCCB or 15 

times rated current for a thermal-magnetic MCCB and "shall be within 

the range stated on the trip curves or 15 times the rated current." 

According to WBN design specification WB-DC-30-28, "LOW and Medium 

Voltage Power Systems" (Revisicjn 9, December 29, 1993), the [safety] 

functions of MCCBs in the Class 1E 480-Vac shutdown boards is to 

provide fault (short circuit) protection for thnir associated branch 

circuits (cables), to provide selective coordirIation of their 

instantaneous trips with the short time trips of their upstream 

feeder breakers (West i nghouse DS type metal cl ad ai r circuit 

breakers), and to allow normal starting of their loads without 

premature tripping. 

However, the procedures and acceptance criteria in the test 

procedure (which reflected those of TVA MI-57.27, Revision 11) did 

not provide for testing the adjustable magnetic trips at both the 

low and high ends (or recording this data) or even at the expected 

setting values based on end use. They did not provide for verifying 

that the MCCBs would not trip at current values lower than the lower 

tolerance limit of the trip curves with which they were selected, 

and did not in all cases ensure that the MCCBs would trip before the 

feeder breaker (i.e. trip in less than 70 milliseconds (ms) at 

current values near the pickup of the feeder short time trip). 

Also, the test current value of 15 times rated current was too high 

to provide meaningful performance data to verify that the MCCBs 

would trip within their design current ranges and times indicated on 

the trip curves, over their adjustable range where applicable (with 

appropriate tolerance for field testing). Therefore it was not 

clear, for example, that. in all cases, insulation thermal limits of 

associated protected cabling would not be exceeded if MCCB ratings 

and settings for cable protection were selected on the basis of the 

trip curves. 

The procedural deficiencies cited above were attributed to the failure 

to adequately translate design basis requirements into design output 

documents (i .e., instructions, procedures and drawings). As stated 

previously, TVA-provided specification, MI-57.27, Revision 11, formed 

the basis for UCI test procedure NQTP-1816. TVA's failure to adequately 

translate design basis requirements (e.g., system design voltage ranges 

11 

116 
or variations) into design output documents (i .e., instructions, 

procedures and drawings), resulted in an inadequate review for 

suitablilty for application. 

Deficiency 94-201-D2. 

As a result, this item is identified as 

In addition to the deficiencies in that test procedure described above, 

which were reflective of provisions of the MI, the MI itself contained 

the following discrepancies: 

The note below Paragraph 5.7 stated that inrush current withstand 

capability (e.g., during motor starts) is verified by so-called 

circuit function test (presumably a post installation test). This 

statement appeared to be a justification for riot performing the 

magnetic trip test in a manner that would verify that the breaker 

will not trip below the lower tolerance limit of its magnetic trip 

band or setting during an inrush transient. The post-installation 

test is normally performed to provide a final check of MCCB 

reliability and for making fine adjustments to the magnetic trip 

setting for adjustable MCCBs as required (providing a sufficient 

number of starts are performed to eliminate the random asymmetrical 

current (DC offset) factor). However, an MCCE procurement 

acceptance test is supposed to verify the breaker to be performing 

within its design parameters before releasing it to the plant for 

testing and cycling plant equipment unnecessarily or requiring 

excessive adjustment. 

Caution below Step 6.6.2 [18] to stop test if breaker doesn't trip 

in ten seconds is too long to wait at that high current level. 

The 10-percent increase adjustment in Step 6.2.2 E211 was expressed 

as a "factor of 0.1" instead of 1.1. 

Section 7.0, Step [3] at the end of the test procedure calls for 

verifying that all test equipment is within its calibration period. 

Checking this after the test instead of before may result in having 

to pzrform the test again. 

The "trip free" function of the breakers was not checked. 

The procedural deficiencies cited above involved wording that would not 

as written, and with the presumed verbatim compliance, accomplish their 

intended purposes or were deviations from accepted industry practice 

without apparent reason. The instructions used for this procurement 

were not appropriate to the circumstances, lacking appropriate 

qualitative and quantitative acceptance criteria to ensure that the 

breaker would perform its safety-related function and not fail in a 

manner adverse to safety under all design basis conditions. 

Accordingly, these procedural deficiencies are identified as part of 

Deficiency 94-201-D9. 

12 

117 
In response to the above concerns, TVA provided information that 

satisfactorily addressed the issue of premature tripping for future 

testing with its revised procedure. TVA also provided analysis to show 

that selective coordination should be maintained with the particular 

design and conditions at WBN. It was still not clear, however, that TVA 

had demonstrated that all protected circuits would in fact be protected 

under all design basis conditions. 

Although many of the problems found with Revision 11 of MI-57.27 were 

corrected in Revision 12, the test procedure used to test these breakers 

(and presumably, others in previous procurements) was based on Revision 

11, and therefore, UCI did not adequately test the breakers described in 

PEG Package 940002182 for either specific or general purpose Class 1E 

service at WBN. At the time of this inspection, Revision 12 of MI-57.27 

was already in use. 

of this inspection the question of the adequacy of testing of other 

breakers tested under Revision 11. 

However, the team did not address within the scope 

In addition to the apparent deficiencies identified in PEG Package 

9400002182 described above, the following are instar.,>s in which 

objective evidence of quality furnished by the suppiier was not 

adequately evaluated. TVA did not in all instances ensure that 

purchased material, equipment and services met the original procurement 

(design) requirements. 

No documentation in the package that breaker nameplate data is fully 

recorded by vendor as called for in the test specification. 

The information referred to as the Underwriter's Laboratories (UL) 

"rating" is not a rating per se and referring to it as such could be 

misleading. The information that is checked as one indicator of 

authentic components is the UL "listing" number and "issue" number, 

normally printed on the MCCB label decal. 

traceable to the general design and type of the MCCB. The issue 

number should correspond to a period of time following original 

certification of the design, during which periodic followup testing 

is done, and covers date codes as should be identified in the 

manufacturer's records. 

The issue number is 

Paragraph 1.2.3.2 of test procedure NQTP-1816 required that the 

tester record the applicable data for insulation resistance (IR) 

tests, yet no data was recorded; only subjective "Yl' or "N" check 

blocks are provided on the data sheet. 

It was also noted that the UCI data sheet gave the units of maximum 

acceptable and actual trip time in terms of Hertz, the unit of 

frequency or cycles per second; whereas the appropriate units in 

which instantaneous trip times are expressed are a number of cycles 

or a time interval in milliseconds (ms). 

13 
0 

Review of the data sheets for this job indicated that the minimum 

current for the magnetic trip test was to be set at 15 times rated 

current (not within the trip curves) for thermal-magnetic MCCBs. 

Although this current level is itself too high for a valid test of 

circuit fault protective function, in several cases the actual test 

current applied to thermal-magnetic MCCBs was more than 1.5 times 

the 15-times rated current value. For example, for serial number 

2645-10-6, a 20-amp thermal-magnetic MCCB, minimum amps were 1 isted 

as 300, with 476 amps applied. For some adjustable magnetic-only 

MCCBs, test currents of 15 times the setting, far above the 

tolerance band, were applied. For example, for serial number 2645- 

12-1, a 10-amp, adjustable magnetic-only MCCB, minimum amps were 

specified at 150 with 151 applied. 

of MCCBs, were not only much too high for mean.ingful results, but 

also indicated a lack of adherence based on an apparently inadequate 

understanding of the UCI test procedure or the TVA MI on which it 

was based. 

These values, for either types 

The above instances in which objective evidence of quality furnished by 

the supplier was not adequately evaluated, and in which TVA did not in 

all instances ensure that purchased material , equipment and services met 

the original procurement (design) requirements, are identified as 

Def i ci ency 94-201-D3. 

PEG Packaqe *EPEG 920013 - (*EPEG was an interim designation). This QA 

Level I1 package dedicated 125-Vdc closing spring charging motors, GE 

part number 0105C9393P002 (or 2B also acceptable per PEG Package 927254) 

for GE model AM-7.2-500-6HD Magne-Blast circuit breakers used in 6.9-K~~ 

Class 1E service at WBN. Five of these motors were procured at the old 

QA Level I1 (CGIs, by part number only) under TVA contract 85PN8-357877- 

02, dated January 17, 1985, a copy of which was in the package. The PEG 

package identified one additional previous rep1 acement parts contract, 

79P19-270009, but did not state the quantity of motors received or their 

disposition, nor was a copy of the contract included. 

According to Form 575N issue document number A118749 in the file, one of 

the five motors from contract 85PN8-357877-02 was installed under MI 

57.1 (no work order number given) in breaker serial number 256A4605-015 

designated UNID "2-BKR-211-Comp 8." As part of the RIP CAP plan review 

discussed later in this report, Previous Procurement Substantiation 

Process (PPSP) 27724 downgraded this motor from QA Level I1 to QA Level 

I11 rather than dedicating it. The original QA Level I1 TIIC, AFT403B, 

was changed to a QA Level I11 TIIC, BPL195W, per PEG package 90-2728 for 

motor brushes. However, as part of the dedication process, EPEG called 

for changing the QA Level I11 TIIC for the motors back to QA Level I1 

TIIC, AFT403f3, and stated that future procurements would be QA Level I. 

EPEG 920013 included a copy of work order (WO) 92-05903-00, October 15, 

1992, which documented the dedication testing of six motors; five from 

nuclear stores and one instal led in breaker 0-BKR-569-4603017-SY 

designated with a UNID of l-BKR-211-A/5-A. Only five motors had been 

procured on contract number 85PN8-357877-02 and one of those had been 

14 

119 
installed in 2-BKR-211-Comp 8 which for an unknown reason had not been 

included in the testing. The inspection team reviewed WO 91-00149-00 

which documented the installation of TIIC, BPL195W, charging motor 

issued on Form 575N BO5606 in breaker l-BKR-211-A/5-AY in compartment 5 

of 6900V shutdown board 1A-A, GE breaker serial number 256A4603-017. 

The issue record for this WO, 575N No. B05606, August 27, 1991, 

identified the same breaker location and showed procurement document 

number 357877. 

stores and one in l-BKR-211-A/5-A) had been procured under 85PN8- 

This indicated that four of the motors tested (three in 

357877-02. 

Therefore, presumably the other two of the motors in stores were from 

contract 79P19-270009; although, EPEG 920013 did not specifically 

identify their origin. 

inventory adjustment form, document number 62-94-16(9) (1 ast digit 

appears to be 9), January 29, 1994, that indicated that two AFT4036 

motors were to be surplused to close inspection report WBN-SWEC-R93.2, 

but this disposition was not explained or mentioned elsewhere in the 

file. 

The EPEG 920013 folder also contained an on-hand 

Although the dedication of the motors under EPEG 920013 appeared to be 

satisfactory, the motor installed in 2-BKR-211-Comp 8, erroneously 

downgraded to QA Level 111 in PPSP 27724, had at the time of the 

inspection not yet been dedicated (Refer to report section 5.3, 

Deficiency 94-201-D6). Furthermore, the question of the origin of two 

of the motors tested and the status of the two surplused remain 

unconfirmed and the existence of one or more additional procurements or 

additional unaccounted for motors procured under Contract 79P19-270009, 

which the inspectors did not review while on site, has not been ruled 

out. 

One additional discrepancy identified was that the breaker with UNID 

l-BKR-211-A/5-A, in which one of the tested motors had been installed, 

was described in the file as the supply breaker to 480V shutdown 

transformer 1A-A (in compartment 5 of 6900V shutdown board 1A-A); 

whereas, configuration control drawing 1-45W724-1 , Revision 12, dated 

September 21, 1990, reportedly the effective, as-built, revision, showed 

the UNID of the breaker in compartment 5 and with this same load, but 

with a system designator of 212 (480V Class 1E distribution) instead of 

system 211 (6900V Class 1E distribution system) used in the PEG package. 

Although the 1990 drawing designation is consistent with TVA’s practice 

of redesignating breakers with the system designator of their loads 

rather than the system designator of the distribution system of which 

they are a part, the file did not agree with what was supposed to be the 

as-built drawing. 

The concern therefore remains unresolved regarding the disposition of 

any other charging motors that may have been procured in a similar 

manner, installed in safety-related breakers that must operate more than 

once to perform their safety functions, but not dedicated due to 

misidentification and/or misclassification. As a result of this 

ambiguity and the potential for more unaccounted for motors, this item 

15 

120 
is designated as Unresolved Item 94-201-U1 pending further review by the 

1 i censee. 

0 

0 

PEG Packaqe 4821-3-02 - This QA Level 111 package, "Limitorque Parts/Non 

Critical , I' May 22, 1992, was supplemented by PEG Package 9400024180 for 

Limitorque parts. The part in question was a clutch ring which 

functions as part of the clutch mechanism that allows manual or electric 

operation of the actuator. When the manual operation lever is depressed 

and latched, the actuator is set up for manual handwheel operation. 

Return to electric motor operation is effected automatically by the 

clutch mechanism upon the next motor operation, which at the same time 

disengages the handwheel. This and other 1 imitorque spare parts were 

purchased on TVA Indefinite Quantity and Time (IQT) contract 86XNQ838100. The item in question was classified nonsafety-related by TVA 

based on its "non-critical" designation on Limitorque's "Critical 

Component Document," EC-0001, Revision 2, October 9, 1990. 

According to a telephone conversation with Limitorque, TVA and members 

of the inspection team during the inspection, the clutch ring is 

considered to have an active safety function in that it helps to effect 

the automatic return to motor operation after manual operation. 

However, Limitorque explained that the clutch ring is a sturdy piece of 

well lubricated steel that is under very low stress and therefore is 

considered by Limitorque to have no credible failure modes that would 

impair its safety function, nor is its application very sensitive to 

dimensional tolerances or material considerations. On this basis, 

Limitorque determined it to be non-critical. 

This rationale is not inconsistent with the guidance in Appendix B of 

EPRI NP-5652 on part classification. 

endorsed by the NRC in GL 89-02 without taking exception to this 

reasoning. 

NRC staff position on this subject to accept the Limitorque rationale in 

this particular case only. 

that they have the ultimate responsibility for proper safety 

classification; and that while it is not unreasonable to start with a 

manufacturer's functional part classification basis as a basic 

guideline, the specific plant application requirements that the 

component manufacturer may not consider must always be taken into 

account. 

EPRI NP-5652 was conditionally 

Therefore it would not be inconsistent with the promulgated 

The NRC inspection teain pointed out to TVA 

PEG Packaqe 9300006087 - This package, RIMS No. 49930120800, TVA PO P93PJX-36732H-000, April 22, 1993, was for various shapes and sizes of 

structural steel products purchased from Mid-South Nuclear, Incorporated 

(MSN). The same TVA PO was also reviewed by the NRC during its 

inspection at MSN in April 1994 (NRC Inspection Report 99901270/94-01, 

dated March 8, 1994) and concluded that MSN did not adequately dedicate 

structural steel items supplied to TVA.. TVA actioris to address the 

NRC's findings at MSN are as follows: 

16 

121 
Following its review of the NRC inspection report, TVA performed an 

audit at MSN in March 1994, and issued findings against MSN’s CGI 

dedication activities. Several of these findings identified 

discrepancies in MSN’s selection of critical characteristics and the 

lack of adequate documented technical justification to reasonably 

assure that the materi a1 furnished met the designated materi a1 

specification requirements. 

dedication activities by MSN and requested additional testing and 

information from MSN for the material supplied. 

TVA also performed an evaluation of the 

TVA accepted MSN’s response and corrective action to its findings 

and informed the NRC that MSN has improved and upgraded its CGI 

dedication activities to a level acceptable to TVA. 

TVA’s audit checklist for Appendix B audits contains specific 

requirements to review CGI dedication activities if performed by the 

supplier. 

qualified MSN to supply ASME Code and non-code safety-related 

material. MSN’s QA program, including its controls for CGI 

dedication, were reviewed by TVA during this audit. 

TVA performed its initial audit at MSN in July 1992 and 

TVA’s implementation audit of MSN’s QA program in July 1993 included 

a review of its CGI dedication activities. However at the time, MSN 

was only processing and dedicating sheet metal. 

TVA reviewed several sheet metal dedication activities and MSN’s 

ASME Section I11 Code material upgrade activities, but did not 

review any other CGI dedications. 

During the audit 

TVA informed the NRC inspection team that it has developed a new 

standard procurement clause applicable for steel products such as 

those suppl ied by MSN. 

supplier assure that the material meets the requirements of the 

appl i cab1 e materi a1 speci f i cation. 

The new clause would require that the 

Several items supplied by MSN on TVA PO P-93JX-36732H-00 were reviewed 

during the inspection including verification of heat numbers, material 

traceability, and observation of hardness testing of MSN products. No 

deficiencies were identified. 

Within the procurement package review part of the inspection, the inspection 

team identified four deficiencies, one unresolved item and one inspector 

foll owup i tem. 

5 PLANT INSTALLED ITEMS VIA PREVIOUS MAINTENANCE’ ACTIVITIES 

REPLACEMENT ITEMS PROGRAM (RIP) (TI 2512/27, 35065, 35746, 35747) 

In 1989 WBN implemented a Quality Release Program that required all material 

released to the plant to have a PEG evaluation package. Prior to the 

implementation of this program, the current inventory of material was 

evaluated for acceptability. 

procurement (and dedication where applicable) of the subject items per Site 

Procedure SSP-10.C and documented each evaluation in a file called a Previous 

The TVA RIP CAP contractor evaluated the 

17 

122 
Procurement Substantiation Process Package (PPSP). However, due to the large 

backlog of evaluations, it was determined that material could be released 

before it had been evaluated and that this released material would be tracked 

using a Release Tracking Log (RTL) and evaluated at a later date. The RTL was 

in effect from about June 1989 through May 1990. In order to determine the 

adequacy of selected PPSPs, particularly those in the RTL, the inspection team 

reviewed TVA POs , vendor documentation , receiving inspection reports , audit 

and survei 11 ance reports , materi a1 issue records, engineering eval uations , and 

quality records and also discussed the contents of these packages with appropriate TVA personnel in each discipline. 

5.1 QA Level I PPSP Review 

The following PPSPs were reviewed and found acceptable for evaluating the 

item’s specific use at WBN: 

PPSP 28824, 40854 and 40693 - Metallic bar 

PPSP 27152 - Limitorque parts (clutch, sliding sleeve and shaft) 

PPSP 28194 - Butterfly valve 

PPSP 29210 - Valve stem with disc assembly 

PPSP 27725 - Speed changer motor for a Woodward governor 

PPSP 28613 - Voltage regulator 

PPSP 28819 - GE, 6.9-Kv Magne-Blast circuit breakers 

PPSP 29380 - Limitorque limit switch qear assembly 

PPSP 28453 - Solid state relay stored in TIIC ARV824E 

5.2 QA Level I1 PPSP Review 

The following QA Level I1 PPSPs were reviewed and found acceptable for 

evaluating the item’s specific use at Watts Bar: 

PPSP 27747 - Pipe clamp 

PPSP 29107 - Carbon steel plate 

PPSP 28630 - 7/8-inch diameter continuous threaded rod 

PPSP 29277 - Water pump assembly 

PPSP 20376 - Limitorque operator parts (gear and shaft) 

PPSP 29227 - W-2 type cell switch 

PPSP 40853 - Cartridge fuse stored in various TIICs 

PPSP 29102 - Westinghouse Type W-2 rotary switches 

PPSP 28079 - Limitorque limit switch gear assembly 

PPSP 28921 - Temperature transmitter 

The following QA Level I1 PPSPs were reviewed and found to be deficient: 

PPSP 27975 - November 6, 1992, for a printed circuit board for a GEMAC 

power supply, TIIC AJW144B. The critical characteristics for acceptance 

related to normal operability for this item were reasmable and appeared 

to be satisfactorily verified by the post-installation test in 

accordance with the technical manuals. However, the seismic similarity 

analysis for this CGI was weak and illustrates a weakness in the RIP 

equivalency determination methodology (and imp1 ementati on) as prescribed 

by SSP-1O.C. 

18 

123 
The seismic argument was based on two factors: (1) the assertion that 

the circuit board "does not provide seismic quality to the host," and 

(2) an equivalency (like-for-like) determination. The first point does 

not conclusively substantiate seismic adequacy (apparently using an 

approach similar to that of the Seismic Qualification Utility Group's 

Generic Implementation Procedure, GIP-E) because it neglects the 

potential for the item's introducing seismic fragility or sensitivity 

into the host device if, for example (assuming for this discussion 

equivalency to the original, presumably seismically qualified part), the 

replacement part has any loose parts or weak connections or solder 

joints, or the circuit board plug-in socket is not tight enough. 

Second, the inspection team questioned the adequacy of the equivalency 

determination itself. The equivalency determination form, page 3 of 5 

of Appendix B of SSP-1O.C (Revision l), lists several factors to be 

considered in the determination, but does not provide guidance on how to 

apply them, e.g, what to verify if the part number had not been the same 

(first factor), or what is required or can be used if the item was not 

procured from the original manufacturer or supplier (second factor, 

Ilthough not checked in this PDSP). 

to a "national standard" (fourth factor, also not checked) would not 

necessarily ensure consistency of product attributes without some 

knowledge and verification of the manufacturer's controls on batch or 

lot homogeneity and traceability. 

Also, procuring a "commodity item" 

Finally, the third listed factor (also not checked in this PPSP) was the 

statement, "TVA had no reason to believe there were changes in the 

design, materials, or manufacturing processes of the item since the 

original item was procured." It was not clear, first, how a RIP 

evaluator could reasonably make such a determination. 

does not provide any conclusive, objective evidence that there were no 

such changes. NRC GL 91-05 promulgated the staff position that even for 

an item of the same part number (which may or may not indicate changes) 

that was not procured from the same source at the same time, a review of 

design, material and process changes would need to be made. That a 

purchasEr merely "had no reason to believe" that there were changes 

clearly would not serve for such a review to support a like-for-like or 

equivalency determination. 

This assertion 

The equivalency determination form then listed, appropriately, vendor 

contact and document review (and "other") as methods of determination, 

but again, no further guidance was provided. Hence, in this case, the 

reviewer simply 1 isted a contract number as the sole reference document, 

then marked the rest of the form, which called for an equivalency 

determination to be made, and the Appendix completed, I'NA." It was not 

clear how review of one contract, unless the item was received under an 

RD from and original IQT contract, could support an equivalency 

determination. The deficiencies with the Attachments to Procedure 

SSP-1O.C discussed above, that it does not logically accomplish its 

stated purpose and is inconsistent with the guidance in GL 91-05, render 

that procedure inappropriate to the circumstances. 

identified as part of Deficiency 94-201-D9. 

Therefore they are 

In addition, as the PPSP 

19 

124 
did not contain explanations of the identified technical concerns, this 

issue is designated as Inspector Followup Item 94-201-12. 

AVA550G. 

documentation in the file whether the prescribed "stroke test" per the 

cited maintenance request (MR) ensured that the motor met design 

requirements (e.g., full- flow cutoff differential pressure capability). 

It was also not evident how the production dedicated motors were 

demonstrated to be similar to test specimens. Therefore, this item is 

considered Unresolved Item 94-201-U2. 

PPSP 20406 - Dated October 12, 1993, for Limitorque motors, TIIC 

At the time of the inspection, it was not clear from 

PPSP 27732 - Dated November 22, 1992, TIIC ALQ229B, for 125-Vdc trip 

coils, GE part number 6174582601 (also used as closing spring release 

coi 1 according to GE Instruction Book GEI-88763F) , for GE AM-7.2-500-6HD 

Magne-Blast circuit breakers in WBN Class 1E 6900V shutdown boards. 

Five of the coils were procured QA Level I1 under 1VA Contract 78P13- 

234973, received on GE Switchgear Products Department (Chattanooga) 

Packing Slip 19141, and receipt inspected on February 17, 1978 (Receipt 

lnspecticn Report (RIR) 152014). The PPSP associate2 the contract 

number 'wi th what it call s a procurement reference i tem, N3-219-1 , but 

the relationship of these two numbers was not evident. According to 

Form 575N 628809626, dated September 13, 1988, in the PPSP, one of the 

TIIC (ALQ229B) coi 1 s from contract N3-219-1 was instal 1 ed under 

MR A605597 in breaker Serial No. 256A4603-013, UNID 2-BKR-211-2B/lOB. 

The other issue document in the PPSP, Form 575N A118747, December 1, 

1989, indicated that one of these coils, also from contract N3-219-1, 

was installed as a trip coil in breaker with serial number 256A4605-015, 

UNID 2-BKR-211/ERCW-DA (Note that this is the same breaker (also called 

2-BKR-211/Comp 8) in which an undedicated charging motor was installed 

and downgraded under PPSP 27724). 

The PPSP technical evaluation form (Appendix B of the package) listed 

voltage range (design criterion 70 to 140 Vdc) as a critical 

characteristic for design, citing TVA Specification 1765. The list of 

critical characteristics for acceptance (Appendix C) appropriately 

included voltage range, also specifying 70 to 140 Vdc for the acceptance 

criterion. The prescribed acceptance method was post-installation test 

and the PPSP took credit for MI 57.99.4 performed under MR A-605597 for 

verification that this critical characteristic was met. The copy of MR 

A605597 in the PPSP references Form 575N 628809626 and indicates that 

the coil was used to replace the breaker's closing coil in accordance 

with MI 57.99.4, Revision 11. 

The data sheets from this MI are included in the MR package, but it was 

only stated that the breaker operated satisfactorily and the control 

voltage used was not mentioned. Also, there was no documentation in the 

file to dedicate the trip coil installed in 2-BKR-ERCWID-A. 

which documented the issue of the trip coil for this breaker listed 

MI-57.1. TVA was unable to produce a copy of any revision of MI 57.99.4 

to determine what control voltage, if any, had been specified by the 

procedure. The current versi on of MI-57.1 , whi ch superseded 57.99.4 , 

The 575N 

20 
was reviewed in accordance with which the other coil had been installed. 

Revision 21 of MI 57.001, "6900 Volt Breaker Inspection," May 13, 1994, 

did not specify any particular control voltage to use when testing the 

trip coils or closing spring release coils of the breakers. 

neither the list of critical characteristics for design nor for 

acceptance included coil insulation or winding resistance. 

Finally, 

Failure to perform an adequate review of the coils for suitability of 

application and the failure to verify that the item meets the 

procurement requirements expressed in the PPSP is designated as 

Deficiency 94-201-D4. 

(1 acking appropriate quantitative acceptance criteria) rendered it 

inappropriate to the circumstances. Accordingly, it is identified as 

part of Deficiency 94-201-D9. 

The deficiency in MI-57.1 discussed above 

PPSP 27732 also contained two more examples of the use of UNIDs that 

were inconsistent with the UNIDs on the drawings said by TVA to reflect 

as-built conditions. The breaker in which the closing coil was 

installed was identified in the PPSP as it was on the issue and work 

documents as 2-BKR-211-2B/lOB; whereas, Configurat' -.I Control Drawing 

1-45W724-4, "Wiring Diagrams, 6900V Shutdown Board 2B-B, Single Line," 

Revision 12, September 22, 1990, this breaker (supply to the motor of 

auxiliary feedwater pump 2B-B) as UNID 2-BKR-3-128B. The other issue 

document in the PPSP, but not mentioned in the PPSP procedure forms, 

identified the breaker in which the trip coil was installed as 2-BKR211-ERCW/D-A (essenti a1 raw cool ing water). 

2-BKR-211/Comp 8 in PPSP 27724), supply to the motor for ERCW pump D-A), 

on Configuration Control Drawing 1-45W724-3, "Wiring Diagrams, 6900V 

Shutdown Board 2A-A, Single Line," Revision 12, dated 9-22-90, was shown 

as 0-BKR-67-40-A. 

This breaker (a1 so call ed 

NEMA Size 1, full-vol tage, non-reversing, 3-pole, open configuration, 

motor starters (each comprised of a contactor, an overload relay, and an 

auxiliary switch) with 480-Vac coils. Four of the starters (TIIC 

AJP795T) had been purchased QA Level I1 on TVA Contract 87NLC381927, 

dated December 29, 1986, with two issued on Form 575Ns. The PPSP 

indicated that both of these had subsequently been replaced, leaving 

two, the disposition of which was not apparent in the PPSP. The PPSP 

indicated that eight AJP795Ts were later dedicated, qualified and 

supplied by Farwell & Hendricks (F&H), a third-party dedicator, under 

TVA Contract 91NNA42765CY dated February 26, 1991, Change 1, March 25, 

1991 (item 3, QA Level I). According to F&H COC No. 80056.1, F&H 

Technical Procedure 3-001, Revision 0, was used and the results were 

recorded in Data Package 80056.1, said to be on file at F&H. 

PPSP 27739 - Dated October 21, 1992, was for Telemechanique A203C48, 

However, the PPSP did not show how F&H dedicated these CGIs. Although 

F&H had been audited by a Nuclear Procurement Issues Council (NUPIC) 

joint utility team, the PPSP did not indicate that WBN engineering had 

reviewed this specific data package. In particular, system interface 

requirements were not specified. For example, the control voltage used 

was not shown therefore it could not be determined whether the units 

21 

126 
were tested with worst-case degraded/elevated voltage, highest ambient 

temperature, and other WBN-specific service conditions. Also, it was 

not clear how the production units were verified to be similar, 

particularly in terms of seismic performance/sensitivity to the 

prototype test specimens. The units’ Seismically Sensitive Electrically 

Active (SSEA) status was not given. 

The inspectors’ concerns about the details of the F&H dedication arise 

from experience in recent years that indicates an increasing number of 

instances in which the dedications performed by third-party dedicators 

(who supply the material under Appendix B QA programs and accept 10 CFR 

Part 21 reporting responsibility) are not adequate to assure that the 

items are suitable for safety-related service under all of the 

customers’ plant-specific design-basis conditions. Purchasers often 

have not completely or accurately communicated all the specific 

application requirements to the dedicator, typically ordering, as in 

this case by part number only. 

taken these plant-specific factors into account in identifying critical 

characteristics, verification methods, and acceptance criteria that are 

relevant to the particular plant application for which the items are 

being dedicated. 

Nor have dedicators always adequately 

In several responses to inquiries on this matter, TVA did not provide 

information to address the inspector’s specific questions (as documented 

in licensee transcriptions of the inspector’s verbal questions) such as 

(1) How did F&H dedicate the starters (i.e., critical characteristics, 

verification methods, acceptance criteria, and results of tests and 

inspections) (2) How did F&H demonstrate similarity of the production 

units to the seismically tested prototype, (3) what was the extent of 

TVA’s review of Data Package 80056.1, (4) what control voltage was used 

by F&H for testing the starters (pull-in and drop-out tests); and was 

this voltage consistent with or lower than the worst-case degraded 

voltage expected at WBN? 

that F&H was an audited and approved Appendix B/Part 21 supplier and 

therefore, subject to receipt inspection, the Telemechanique starters, 

were acceptabl e. 

Instead, TVA continued to restate its position 

Review of the data sheets eventually sent to the inspector from F&H Data 

Package 800056.1 revealed that the pull-in and drop-out voltages were 

measured. Although the voltages appeared to the inspector to be 

adequate for most credible plant conditions, TVA would be expected to 

review them in light of actual WBN service conditions to confirm 

adequacy. A comparison of WBN service conditions to the manufacturer’s 

specifications may already been done when the original starters were 

specified , but the actual tested starter performance apparently was 

never reviewed heretofore by TVA. 

revealed that contrary to manufacturer’s and general industry guidance, 

the critical characteristics of full-load hold-in capability and starter 

insulation resistance (of contactors, coil, and auxiliary devices) were 

not verified by F&H. 

Review of the data sheets also 

22 
In addition, although, F&H may have verified the dimensions and 

resi stance of separately dedicated overl oad heaters , measurement of 

these parameters for the heater elements installed in the overload 

relays in the starters in question was not documented in the data sheets 

provided for review. Furthermore, although the data sheets did indicate 

that the overload trip function was checked by passing three-times rated 

current through the heater and verifying a trip, the required time delay 

for the class of overload relaylstarter was not checked. 

review, the inspector concluded that in, this case, F&H had not 

adequately dedicated the starters for general purpose Class 1E service 

in a nuclear power plant. 

Finally, the inspector reviewed the copy of the audit checklist used in 

the NUPIC audit of the vendor. 

supplement the reports of the audit and the supplemental surveillance 

(that only covered a Tel emechanique overl oad heater) provided 

previously. The starters were being ordered, in addition to quality 

assurance (QA) and 10 CFR Part 21 (rart 21) requirements, by part number 

only. Therefore, the inspector evaluated the audit reports to determine 

vhether they -veri f i ed that the vendor s program to dedicate the 

commercial grade starters (or other CGIs) included verification of all 

the design specifications associated with that model or part number upon 

which the licensee was relying to ensure suitability for safety service 

at WBN. 

verification was either programmatically required or routinely 

performed, or included in this case. Therefore, on this basis, the 

inspector determined that the audit was not adequate to support 

procurements in which no additional technical requirements other than 

part or model number and minimal description were imposed in procurement 

documents. 

Based on this 

The checklist was requested to 

The inspector could not conclude from this review that such 

TVA may have reviewed the technical data on the model starters in 

question for suitability of application, and finding this acceptable, 

ordered the starters purely on the basis of part number, with Appendix B 

QA and Part 21 specified. However, it was apparent that TVA did not 

verify, either during the audit or surveillance of the vendor, nor by 

examination of products upon delivery, that all critical characteristics 

pertinent to WBN that were being relied upon on the basis of the design 

characteristics associated with the part number were being verified. 

This failure to verify that safety-related equipment meets the 

procurement requirements is designated as Deficiency 94-201-05. 

5.3 QA Level I11 PPSP Packaqe Review 

The following QA Level I11 PPSP packages were reviewed and found to be 

properly classified as nonsafety-related items. 

classifications were found acceptable only for their specific end use and if 

used in similar applications in other components and systems, such 

cl assi f i cati on may change. 

It is noted that these 

23 

128 
PPSPs 27992 and 27905 - Stainless steel wire used to secure equipment 

PPSP 29149 - Pump wear ring 

PPSP 28019 - Limitorque operator handwheel clutch pinion 

PPSP 27067 - Rotary control switch stored in TIIC ACN683G. 

PPSP 27977 - Positioner/controller module stored in TIIC ARV417V. 

The following QA Level I11 PPSP packages were reviewed and found to be 

inadequately dedicated: 

PPSP 28039 - Lockwire used to secure the set screw for motor pinions in 

certain Limi torque operators. 

to document the inadequate technical evaluation of this PPSP. 

following discussion, the inspection team expressed a concern that failure of 

the lockwire could potentially affect the host component from performing its 

safety-related function and that the PPSP package should have been more 

perceptive with justification for not classifying the lockwire as QA Level I1 

and dedicating it as a basic component. The PPSP did not specifically address 

the end use or function of the lockwire as discussed in Limitorque Maintenance 

Update 89-1 (TVA Document No. WBN-VTD-L200-0280, Revision 0, October 3, 1991). 

Part lB, "Item Data," of Appendix By Technical Evaluation, to PPSP 28039, 

states in part, that the purpose of the lockwire is to prevent nuts from 

becoming undone due to vibration; that the lockwire does not perform a safetyrelated function; and that the failure of the lockwire does not influence the 

ability of the host equipment (the Limitorque actuator) to perform its safetyrelated function. 

During the inspection, TVA generated PER 940337 

Based on the 

The PPSP stated that the end use of the lockwire was to prevent a set screw, 

not a nut, from backing out. The set screw is used to fix the motor pinion 

once it is properly orientated on the actuator motor shaft. The inspection 

team also reviewed the Limitorque Maintenance Update and participated in a 

conference call with TVA and Limitorque to discuss the safety functions of the 

lockwire and the set screw. Limitorque confirmed that the motor pinion gear is 

attached to the motor shaft by means of a key (to transmit the rotary force) 

and a set screw (to prevent axial movement of the gear on the shaft) and 

stated that the set screws are normally set by staking, which for the size - 000 actuators, is sufficient to retain the set screw. 

However for all other size actuators, the set screws cannot be staked due to 

the hardness of the steel motor pinion gears. 

from backing out, Limitorque recommends the use of thread locking compounds or 

1 ockwire inserted through grooves. Limi torque a1 so stated that it intended to 

review its critical parts list and evaluate the need for classifying the 

lockwire and set screw as critical. 

To prevent these set screws 

Page 17 of TVA Document WBN-VTD-L200-0280, lists the lockwire as 0.047-inch 

diameter wire. PPSP 28039, identifies the lockwire as 0.06-inch diameter 

(TIIC No. AXR 68510, however the correct TIIC for the 0.047-inch diameter 

stainless steel lockwire, according to TVA, should have been AXR 684M. TVA 

stated that it would document the use of the incorrect lockwire on the PER and 

appropriately disposition the PER once the Limitorque evaluation of the proper 

classification of the set screw and lockwire is completed. As a result of the 

misclassification of the lockwire, the inspection team selected two additional 

24 

129 
QA Level I11 PPSP packages for wire and found that they were properly 

classified as QA Level I11 (nonsafety-re1 ated) and performed no safety 

function for their identified end use. Satisfactory resolution of the 

lockwire classification concerns is identified as Inspector Followup Item 

94-201-13. 

PPSP 27724 - Dated April 6, 1992, for a closing spring charging motor, GE 

part numbers 0105C9393P002 and 2B (original QA Level I1 TIIC AFT403B changed 

to QA Level I11 TIIC BPL195W per PEG package 90-2728 for motor brushes), for 

AM-7.2-500-6HD GE Magne-B1 ast circuit breaker , instal 1 ed in breaker seri a1 

number 256A4605-015, designated UNID "2-BKR-211-Comp 8," per Form 575N A118749 

and MI 57.1. This PPSP, on the same basis as the PEG package for the brushes, 

downgraded the motor from QA Level I1 (the old QA Level I1 at which it and 

four others were originally purchased under TVA contract number 85PN8-357877- 

02, dated January 17, 1985) to QA Level 111. 

The rationale documented was that (1) the motor was not supposed to impact 

automatic function of the safety-related (Class 1E) breaker, (2) the motor is 

isolated from Class 1E power supply by fuse, and (3) closing springs may be 

charged manually. 

motors were not needed for the breaker to perform its safety function for 

several reasons. First, although the breaker will be capable of closing one 

time automatically once the closing spring has been charged by the motor, it 

may be required under certain design basis event (DBE) scenarios within the 

licensing basis of WBN to close more than once automatically. A normally open 

bus feeder breaker with a charged closing spring (exhibiting a spring charged 

white light on the switchboard), could be reasonably expected to close the 

breaker once upon demand such as an engineered safety feature (ESF) actuation 

signal, and although it may trip upon a subsequent loss of offsite power, it 

should not normally be required to reclose automatically during the DBE 

because the 6.9-Kv shutdown board should be re-energized under these 

circumstances by the emergency diesel generator (EDG). Therefore in such an 

instance, the closing spring charging motor would not be required to operate 

during the DBE and so would have no safety-related function. 

This rationale did not support the assertion that the 

However, a normally-open breaker supplying a Class 1E load from the 6.9-Kv bus 

would need to shut automatically upon an ESF actuation, but would be required 

to trip upon a subsequent loss of offsite power and then reclose automatically 

with a signal from the EDG load sequencer after the EDG had re-energized the 

bus. In this case the load breaker's motor would be required to operate 

during the DBE and perform a safety function and would be itself classified 1E 

accord i ngl y . 

Although the PPSP did not establish that the former scenario was the case for 

the installed motor, TVA's response to the inspection team's inquiry as to the 

purpose/function of breaker 2-BKR-211-Comp-8 was that it was the alternate 

power supply to 6.9-Kv shutdown board 2A-A from station supply transformer 

CSST D. This would justify downgrading the installed motor to QA Level I11 as 

was done by this PPSP. 

distribution diagram, "Wiring Diagrams, 6900V Shutdown Board 2A-A, Single 

Line," drawing number 1-45W724-3, Revision 12, dated September 22, 1990, the 

inspector found that the breaker in compartment 8 (indicated by the "comp 8" 

However, upon review of the associated 6.9-Kv 

25 

130 
in the UNID) was in fact breaker 0-BKR-67-40-AY supplying motor 0-MTR-67-40-A 

for essential raw cooling water pump D-A. 

Class 1E load breaker and therefore its charging motor did have a safety 

funct i on. 

This breaker was a normally-open 

Second, the fact that closing springs may be charged manually is irrelevant 

because if the springs should be required to be recharged in order for the 

breaker to reclose during a DBE to perform a safety function, the motor would 

be required to operate to meet system safety function requirements and credit 

for manual charging could not be taken. Also, even for a motor that is 

determined to be non Class lE, the PPSP did not establish that the fuses were 

proper Class 1E isolation devices. This charging motor was procured as a CGI, 

but never dedicated for the safety-related application in which it was being 

used. 

compounded the error. The misclassification resulted in failure to review the 

item for suitability of application as well as failure to verify that the item 

meets the procurement requirements. 

Deficiency 94-20146. 

The improper quality level (safety class) downgrading by PPSP 27724 

Accordingly, this issue is identified as 

PPSP 28973 - Dated November 14, 1992, for a WestinghoL : 592C114G81 direct 

trip actuator (DTA) for tripping Westinghouse Type DS-20b 480-Vac metalcl ad 

air circuit breakers in response to signals from the breaker's Amptector 

overload trip unit. Two of these DTAs (TIIC AMNO16P) were purchased QA Level 

I1 under contract 86PJH-357168 from Mills & Lupton Supply Company (M&L). The 

Westinghouse packing 1 ist, shipment no. 23771791, (referencing Westinghouse 

General Order C072644, the M&L order and the TVA contract) indicated the DTAs 

were shipped direct from the Westinghouse warehouse in Charlotte, North 

Carolina, to WBN, yet the PPSP stated the items were received from the 

distributor, not Westinghouse. 

According to issue document Form 575N A132952, dated April 5, 1990, one of the 

DTAs from contract 357168 was issued for installation under MI 57.2 (no WO 

number given) in a breaker designated UNID 1-BKR-212-Al/81\-Ay seri a1 number 

010, located in compartment 8A of Class 1E 480V shutdown board 1A1-A. 

Although configuration control (as-buil t) drawing no. 1-45W749-1 (Revision 20, 

dated 9-22-90) , "Wiring Di agrams 480V Shutdown BD 1A1-A, Si ngl e Li [le, It shows 

the breaker in compartment 8A as UNID 0-BKR-239-1-AY the 1992 PPSP still 

referred to it by its previous UNID using the system 212 designator instead of 

the load system 239 designator as apparently is the current practice. 

239 is WBN's 250-Vdc battery power supply system and this breaker is the 

alternate feeder to one of the 250-Vdc battery chargers. 

System 

However upon further review of the package, the inspectors identified that the 

justification for downgrading the DTA to QA Level I11 was in error. The 

rationale as documented in the PPSP was that although system 212 which the 

breaker was in is a safety-related system, its load in system 239 (250-Vdc 

battery charger) was a nonsafety-related load. Therefore, the PPSP preparer 

reasoned incorrectly that on this basis the breaker, and hence the DTA, could 

be considered nonsafety-related. This was incorrect because it did not take 

into account the fact that this circuit breaker must act as a Class 1E 

i sol ati on device to cl ear faul ts on the nonsafety-re1 ated 1 oad and associ ated 

buswork and cabling protected by the breaker in order to protect the system 

26 

131 
212 Class 1E 480-Vac bus. 

means of the DTA; therefore the DTA must be considered to have a safety 

function. 

I11 without dedicating it means that it compounded the error of installing it 

without proper dedication initially, leaving an undedicated commercial grade 

i tem DTA of indeterminate qual i ty in a safety-re1 ated appl i cati on. Fai 1 ure to 

review the item for suitability of application and failure to verify that the 

item meets the procurement requirements is identified as Deficiency 94-20147. 

Fault tripping in this breaker is accomplished by 

The fact that this PPSP downgraded the installed DTA to QA Level 

Subsequent to the inspection, TVA submitted information to the NRC that 

described its dedication plan for the DTA in question. Proposed Revision 1 of 

PPSP 28973, dated August 4, 1994 (through the independent reviewer, but not at 

that time approved), was included in the information submitted and was 

reviewed with some discrepancies noted: 

a Attached to the revised PPSP was Work Request (WR) C152372, dated August 

4, 1994. 

question was installed as the component to be worked on, called out 

several tests. First, it specified testing in accordance with the 

vendor manual, Section 11-7, wbich would exercise the DTA. Then, it 

also called out testing insulation resistance (IR) and so-called "coil 

voltage." However, while an IR test of the DTA appeared to be 

appropriate, it was not clear from the WR exactly which component's IR 

was to be tested. 

used to test the breaker, ostensibly after installation of the 

rep1 acement DTA, but the MI only covered insulation resistance 

measurement of the breaker's main poles. 

referenced in two places, but there was no Section 6.10 in the data 

package which only went up to Section 6.3.3. 

similar number was Section 6.1.10, but it had nothing to do with the DTA 

which is operated in the overload trip tests of Section 6.13. 

addition, the "coil voltage" test specified could not be found as such 

in either the referenced vendor manual, WBN-VTD-W120-3023, or in 

MI-57.2. It was not clear what coil voltage was to be tested, since the 

DTA is only energized by the Amptector trip unit in the breaker under 

sufficiat overload or fault conditions. 

This WR, which only identified the breaker in which the DTA in 

The revised PPSP referenced MI-57.2 (Revision 15), 

Section "6.10" of MI-57.2 was 

The only section with a 

In 

a Page 4 of the revised PPSP, the like-for-like determination form, 

indicated (block checked) that "TVA had no reason to believe there were 

changes in the design, materials, or manufacturing process of the item 

since the original item was procured, yet Section 3-1.6 of the 

referenced vendor manual described several changes in the DTAs supplied, 

and it was not clear that all changes involved a change in part number. 

Also, the various possible DTAs are of different colors, slightly 

different terminal or pigtail configurations (one is not seismically 

qualified), yet the PPSP did not mention checking these attributes to 

ensure a DTA fully compatible with the Amptector trip unit in the 

breaker, particularly if the Amptector had been replaced. 

a Page 5 of the revised PPSP, the technical evaluation form, listed part 

number, manufacturer, dimensions/configuration, coil resistance or 

continuity, and coil voltage as critical characteristics for design. 

27 

132 
The list of critical characteristics for acceptance listed part number, 

manufacturer , dimensions/conf iguration, and conti nul ty/coi 1 voltage. 

First, for coil resistance, having been identified as a critical 

characteristic ("for design") there was no justification for not using 

it for acceptance, which is not consistent with GL 91-05. Also, it was 

not clear how "continuity/coil voltage" together as a critical 

characteristic for acceptance describes a meaningful test. Again, coil 

voltage was not explained. And finally, the insulation resistance test 

specified in the WR supposed to be used to complete the dedication, was 

not listed at all. 

It is standard practice to measure IR among phases and to ground from each 

phase either with the breaker closed, or if open, on the load side and the 

line side, followed by a test of each pole from line to load (open). However, 

according to the completed MI in the PPSP, this test was performed with the 

breaker tripped (Step 6.1.13RR), yet only one set of readings for each pole to 

ground and to each other was recorded, and there were no IR readings recorded 

from line to load. This procedure did not appear to adequately check IR of 

the breaker and it would not ensure checking of DTA insulation resistance at 

a1 1 These def i ci enci es rendered the MI i nappropri ate to '.he circumstances , 1acl;ing appropriate qualitative and quantitative acceptance criteria. 

Accordingly they are identified as part of Deficiency 94-201-Dg. 

5.4 Review of Generic Dedication Packaqes 

The inspection team reviewed the list of bulk commodities listed in TVA 

Procedure SSP 10.C and based on discussions with RIP and PEG personnel, determined that TVA presently performs no generic dedications and that in its 

evaluation of plant installed items, only one generic technical evaluation was 

used for accepting installed fasteners. 

both commercial and safety-related fasteners used at NBN were examined and 

tested to assure compliance to specification requirements. 

TVA stated that since the late 1970s 

TVA a1 so performed overchecks and screened 1 ots/heats of nonconforming 

fasteners as documented in nonconformance reports, Institute of Nuclear Power 

Operations (INPO) assessments, NRC inspection reports, TVA's response to NRC 

GL 87-02, fastener test results from TVA's test laboratories, and internal TVA 

memorandums for the period 1979 through 1989. 

The inspectors also reviewed TVA Construction Specification 6-53, "ASME 

Section I11 and Non-ASME Section I11 (including AISC, ANSI/ASME B31.1 and ANSI 

831.5) Bolting Material .'I Selective revisions were reviewed and it was determined that the specification required overchecks of all fasteners since its 

issuance in 1978. 

justification of the WBN fastener program, it was concluded that TVA had an 

adequate basis for not testing fasteners under the RIP CAP plan. 

Based on this review and discussions of the engineering 

5.5 Control of Product Contaminants 

TVA-WBN through its Process Specification 4.M.l.l and PF material purchase 

specifications identifies technical requirements for chemical limitations for 

elements such as halogens, sulfur and water leachable halides, for products 

28 

133 
used in austenitic stainless steel components and systems. TVA has 

established a pol icy that based on satisfactory historical performance of 

suppliers of these type of items, no audits, surveys, surveillance of supplier 

or overchecks of the items received is necessary to ensure that these items 

meet PF chemistry limitation requirements. 

The inspectors di scussed with TVA the rzqui r-ements of American National 

Standards Institute (ANSI) N45.2.13-1976, "Quality Assurance Requirement for 

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

endorsed by NRC Regulatory Guide 1.123-1977. 

"Certificate of Conformance," was a permitted method for ensuring that an item 

meets specification. It was pointed out to TVA that Subparagraph F of Section 

10.2 provides additional requirements and states, in part, that means shall be 

provided to verify the validity of supplier certificates and the effectiveness 

of the certification system. The inspectors concluded that even though the 

past performance of these suppliers may indicate satisfactory implementation 

of their QA program, Appendix B to 10 CFR Part 50 requires assessment of the 

effectiveness of the control of quality of these suppliers at some interval 

based on the importance, complexity, and quantity of the product received. 

TVA stated that Section 10.2, 

TVA General Specification 6-29, Volume IVY Process Specification 4.M.1.1, 

"Material Fabrication and Hand1 ing Requirements for Austenitic Stainless 

Steel," Revision 17, dated March 25, 1994, and several lower tier process 

specifications such as PF-1060, "Purchase Specification for Gaskets," 

Revision 4, dated November 23, 1993, and PF-1061, "Purchase Specification for 

Lubricants Unrestricted Use on Stainless Steels Primary and Secondary 

Syst'ems," Revision 3, dated July 31, 1992, permit the use of a COC from a 

supplier, not on TVA's ASL, certifying that actual chemistry test results meet 

the required product limitations required by the applicable TVA specification 

for certain products used in safety-related austenitic stainless steel systems 

and components. 

for consistency with Section 10.2 of ANSI N45.2.13-1976, and will determine 

the need to revise its policy based on the results of that review. 

result, this item is identified as Inspector Followup Item 94-201-14. 

TVA representatives stated that they will review the policy 

As a 

5.6 Replacement Items Installed by Previous Maintenance Activities 

5.6.1 Maintenance History Data Base 

To address replacement items installed by previous maintenance activities, WBN 

developed the Maintenance History Data Base (MHDB). This data base was 

devel oped by revi ewi ng some 91,000 work control documents, survei 11 ance 

instructions, and operating instructions. The work control documents were 

generated for (1) corrective maintenance, (2) preventative maintenance, (3) 

maintenance performed during surveil1 ance testing, and (4) maintenance 

performed during plant modifications. However, 

it was useful to the 10 CFR 50.49 review task because of the extensive amount 

of data it contained which related to MRs and work requests (WRs), as well as 

the cross references between documents such as contract numbers and material 

issues tickets. 

Systems Group. 

The MHDB is no longer in use. 

The MHDB is now held in inactive status by the Information 

29 

134 
These work control documents were issued in the time period between original 

system(s) turnover and January, 1989, the date when the Quality Release Program was implemented. The review of the work control documents was performed 

to determine (1) did the maintenance activity install parts? (2) Was the host 

equipment safety-related? and (3) Was the host equipment subject to 10 CFR 

50.49 requirements. A l'yesl' answer to questions number 1 and 2 meant the work 

document would be added to the data base along with other information such as 

(1) MR Number, (2) Contract Number, (3) TIIC number, (4) 575N Number, and (5) 

Quality Classification. This data base made it possible for use in several 

cross reference checks. A 'lyesll answer to question number 3 allowed the 

grouping of the 10 CFR 50.49 equipment subset for later use. 

The advantage of the data base information being available up front to the 

installed items evaluations is that the process started with a specific "end 

use," the unique identification number for the host equipment, and works 

backward via the work document and the issue document (Form 575N) to the 

procurement documents. Work control or 575Ns referencing unclassified 

equipment, or equipment designated by a temporary identification number, would 

precipitate a 'Pro-TIIC designation in the MHDB. 

Items on the MHDB were reviewed, consolidated, evaluated ana dispositioned. 

Old items identified as installed in host equipment subject to the 

requirements of 10 CFR 50.49 were evaluated for acceptability. The MHDB also 

identified a total of 139 consolidated safety-related potentially problem 

items that were not subject to the requirements of 10 CFR 50.49. 

5.6.2 Problem TIICs 

The population of maintenance installed items reviewed by TVA consisted of 

several incomplete TIIC items listed in the MHDB called Problem or Pro-TIICs. 

These items identify inconsistencies with the correlation of document 

identifiers such as the TIIC number on the MR or the Forrn 575N number. These 

items resulted from in-process reviews of data by the clerks reviewing the 

documents for inclusion in the data base. There were inconsistencies that 

required a detailed evaluation by an engineer knowledgeable in the process to 

resolve exactly what the TIIC number was, or what the material really is. 

Many involved the simple transposition of the alpha-numeric elements of the 

TIIC number so that what shows up on the MR, Form 575N, or contract is an 

invalid TIIC number, or a TIIC number associated with material obviously 

incompatible with the description and other data on the f4R. 

difficulties were that the TIIC number was not legible, not complete, or 

missing. 

A total of 139 Pro-TIICs were identified in the MHDB as being safety-related, 

but not installed in host components subject to the environmental 

qualification requirements of 10 CFR 50.49. The RIP group evaluated these 

items and determined that 116 items were duplicate items or ricnsafety-related. 

The remaining 23 items were included and evaluated by RIP. All MHDB Pro-TIICs 

installed in host components subject to 10 CFR 50.49 were captured and added 

to RIP, and subsequently evaluated as part of the "50.49 EQ Task." 

Other examples of 

30 
5.7 RIP Hold Items 

A large number of items had previously been rejected by the Material 

Improvement Project (MIP) (See Section 7.1) because they were unable to 

qualify existing material in inventory. 

affected material had been already been installed in the plant further 

complicating the issue. The inspection team selected a sample of 15 of these 

items from the RIP On Hold database. 

TIIC numbers that had been placed on hold pending further action necessary to 

resolve the issue. 

adequacy of applicant corrective actions for each of the selected items. 

packages reviewed included the following: 

However in many cases some of the 

This database contained 2246 separate 

PPSP packages were selected for review to determine the 

PPSP 

PPSP 40858 - Identified electrical connectors procured as QA Level I11 

items during July 1990. The items were included under TIIC BBF564V. TVA 

determined that the items had been correctly procured as QA Level I11 and had 

not been used in any safety related application in the plant. 

TVA P:ntract 92KNA-43589CY dated December 10, 1991. The item had been 

procured as QA Level I1 without proper dedication. The item was included 

under TIIC BQE436R and had been installed in the plant as control air pressure 

regulator, 0-PREG-065-0028B. TVA stated that corrective actions associated 

with this item have not yet been completed and the PPSP is considered open. 

It was noted that WR C152395 had been written to remove/replace the regulator 

with a properly procured item. 

had been issued to address reportability and other related issues associated 

with this problem. 

PPSP 40831 - Identified an ITT Conoflow pressure regulator procured under 

Additionally it was noted that PER CHPER930033 

0 

control circuitry for the component cooling water booster pump 1A-A under MR 

631548 on July 12, 1990. 

did not have adequate documentation. 

AHP725N. TVA stated that corrective actions have not yet been completed for 

this item and the PPSP is considered open. It was noted that WR C152300 had 

been written )Lo replace the relay with a properly procured item. 

PPSP 41016 - Identified 10K resistors that had been installed in 125-Vdc 

vital battery power system panels under MR A479681 on April 26, 1985. The 

resistors should have been procured as QA Level I or Level I1 but did not have 

adequate documentation. The resistors were included under TIIC AVC377E and an 

incorrectly referenced contract number may have been the reason for lack of 

traceability of the material. TVA stated that corrective actions have yet not 

been completed for this item and the PPSP is considered open. It was noted 

that WR C152390 had been written to replace the resistors with a properly 

procured items. 

PPSP 41033 - Identified a time delay relay which had been installed in the 

The item should have been procured as QA Level I but 

The item was included under TIIC 

documentation or known receiving date. The bearings were included under TIIC 

APB197E and had been placed in RIP hold status due to lack of traceability of 

the material. 

present in stock and no records could be found which would show that any of 

PPSP 40782 - Identified various bearings in stock without adequate 

The applicant had determined that a total of three items were 

31 

136 
the items had been used in the plant. It was verified through the Material 

Management System (MAMS) program that the three bearings had been surplused 

and the TIIC had been deleted. 

procured under TVA Contract, 256489, dated April , 1979. The items were 

included under TIIC ART495C and had been procured as QA Level 111. 

two of the items had been issued for use and installed in plant equipment 

under TVA Form 575-628609643 during July 1986. 

instal 1 ed on radiation monitor, 1-RM-090-0010. 

not a basic component and failure of the item would not affect the ability of 

equipment to perform its intended function. 

surpl used. 

PPSP 40944 - Identified the possible incorrect use of insulator washers 

However 

Both washers had been 

TVA eval uated the washer as 

The remainder of the items were 

PPSP 40018 - Identified the possible lack of adequate documentation for a 

Crosby, three-inch, relief valve bellows assembly used in relief valve, 1-RFV74-505 for the residual heat removal system. The bellows had been procured as 

QA Level I. However the item was placed on RIP hold status to resolve a 

concern about lack of proper documentation. Crosby had been an approved 

sur,-lier and a COC had been furnished along with the item. However, TVA's 

original contract had specified the bellows as one of the various replacement 

parts that was required to meet specifications for pressure retaining 

components. This would have required a CMTR for material used to manufacture 

the bellows. 

evaluated the item as not requiring a CMTR. 

been in error as the bellows was not actually a pressure retaining component. 

The bellows was acceptable as QA Level I. 

Crosby had not furnished a CMTR. Subsequently the applicant 

The original contract wording had 

PPSP 40957 - Identified the possible improper usage of thermocouple wire in 

the plant. The thermocouple wire was included under TIIC AXX798H and had been 

procured under TVA Contracts 322205, 362363, and 371807 from 1981 through 1985 

as QA Level I1 without proper dedication. The wire was placed on RIP Hold 

Status pending resolution of the issue. No stock remained in stores. TVA 

evaluated the wire as requiring downgrade to non-quality related and 

determined that all wire under this TIIC had been used in nonsafety-related 

applications (heat tracing), which was not required to function during an 

accident. 

and that the TIIC had been deleted. 

(BQR841P) had been issued for future material which was classified as QA 

Level 0. 

It was verified through the MAMS program that no stock was on hand 

Additionally it was noted that a new TIIC 

without proper dedication. 

Contract RD-619204 as QA Level I1 but no supporting documentation was required 

by the contract. Five of the items had ,remained in stores but the sixth item 

could not be accounted for. The items were included in TIIC ABW859C and were 

placed on RIP hold status pending resolution of the issue. TVA evaluated the 

issue as acceptable. The five items known to be in stores were surplused and 

with no record of issue or other transaction history for the item existed it 

was assumed that the sixth item was lost. 

program that no stock was on hand and that the TIIC had been deleted. 

PPSP 40293 - Identified possible usage of bourdon tubes in the plant 

Six bourdon tubes had been procured under TVA 

It was verified through the MAMS 

32 

137 
that had an improper TIIC designation. 

originally included under TIIC BBH991X and had been procured under various TVA 

Contracts during 1988 and 1989 as QA Level I material. 

had been incorrectly placed under a TIIC number which reserved for QA Level I1 

material. 

material. 

remained in stores. 

resolution of the issue. 

Level I. 

for TIIC BBH991X and that the TIIC had been deleted and that all remaining 

pipe hangers in stores were transferred to a new TIIC Number BHQ340A which 

properly designated the material as QA Level I. 

PPSP 40967 - Identified the usage of piping hanger material in the plant 

One hundred pipe hangers were 

However the material 

There had not been a problem with the actual qualification of the 

Seventy of the pipe hangers had been installed in the plant and 30 

The pipe hangers were placed on RIP hold status pending. 

TVA evaluated the pipe hangers as acceptable for QA 

It was verified through the MAMS program that no stock was on hand 

capacitors in the plant. 

separate TIICs had been procured under various TVA Contracts from Foxboro 

during 1989 and 1990. 

but had originally been thought not to meet COC requirements and were placed 

in RIP Hclcl Status pending resolution of the issue. A n..,,,ber of the 

capacitors had been installed in various nuclear steam system supply 

instrumentation located in the plant. 

reached end of shelf life and had been surplused. TVA evaluated the 

capacitors as like-for-like replacements purchased from the original supplier 

of host equipment and therefore acceptable for QA Level I1 without further 

dedication. This determination was based on the existence of an original COC 

provided by Foxboro who, at the time, was on TVA’s ASL. 

PPSP 40974 - Identified the possible usage of improperly procured 

Various electrolytic capacitors included under 18 

The capacitors were procured as QA Level I1 material 

The remainder of the material had 

TVA had conducted an audit at the suppliers facility prior to that time. This 

Audit, 89\1-97, eval uated the suppl i ers program for control of qual i ty requi red 

by TVA contracts for providing commercial grade modules and spare parts. The 

inspection team reviewed the audit report for this applicant vendor audit and 

noted that review of capacitors as spare parts had been included as part of 

the audit. 

PPSP 41039 - Identified the possible usage of improperly procured conduit 

bushings in the plant. Three hundred, 1.5-inch galvanized electrical bushings 

under TIIC AWD041V had been procured under TVA Contract 43501B during 1989 as 

QA Level I11 material. 

surplused and the remainder installed in the plant. 

under this contract remained in stores. 

used in safety related applications the issue was placed on RIP Hold Status 

pending resolution of the issue. 

QA Level 111. 

component. 

conduit does not serve a safety related function. 

Two hundred fifty one of the bushings had been 

No bushings procured 

Because some of the bushings were 

TVA evaluated the bushings as acceptable for 

This determination was based on the bushings not being a basic 

Bushings were installed in various conduit applications and 

PPSP 41040 - Identified the possible usage of improperly procured cable in 

a safety-related application in the plant. 

had been procured as QA Level I1 without proper dedication and installed in 

Module, 1-LPT-24-69, under MR A619637 on February 21, 1990. The item was 

originally included under TIIC BHX678Y. However the modified module (TIIC 

A special purpose cable assembly 

33 

138 
ARD287N) containing the special cable assembly was subsequently determined to 

be non qualified and returned to stores under a TVA Form 575-A031233 for 

credit. It was verified from the MAMS program and by a review of PEG package 

01493 that the module that contained the cable assembly had been downgraded to 

QA Level 0. 

cable. Following return of cable (Form 575N credit) from cable reel WB14735 a 

discrepancy of 352 feet of cable existed. 

been previously received satisfactorily and documentation for only 1730 feet 

which had been returned for credit. 

issuance existed. The cable had 

been procured under TVA Contract 290617-2 as QA Level I and the qualification 

of the cable was never in question only the usage. 

included under TIIC AXK180W and the issue placed in RIP hold status pending 

resolution of the issue. TVA evaluated the issue as acceptable based on a 

review of the Computerized Cable Routing System (CCRS). This review revealed 

that this type cable was not listed or referenced and therefore not installed 

in any safety-related applications. 

PPSP 41009 - Identified a possible discrepancy in usage of electrical 

Two thousand eighty two feet had 

No other documentation for any other 

The returned 1730 feet was later surplused. 

The 4AWG cable was 

PPSP 40853 - Identified various cartridge type fuses for which procurement 

documentation could not be located. These fuses were included under a series 

of 54 separate TIICs and had been procured as QA Level 11. 

documentation was not available to support their use in safety-related 

applications. 

applications the issue was placed on RIP hold status pending resolution of the 

issue. TVA had resolved this issue based on corrective actions which had been 

included under WBFIR910192112 and as part of TVA's Master Fuse List (MFL) 

Special Program. TVA had verified that all fuses in Class 1E circuits 

identified on the MFL have a solid path of traceability. 

installed and non-installed fuses placed in plant storage areas. 

However 

Because some of the fuses had been used in safety related 

This included both 

Non-traceable fuses were discarded or returned to stores to be surplused. 

Inplant storage facilities for fuses were established and must meet the 

requirements of SSP-10.04, "Material Issue, Control , and Return." 

Additionally, TVA's MFL Special Program was reviewed by -the NRC as documented 

in NRC Inspection Reports 50-390 , 391/92-27 and 93-31. 

the staff determined that fuses listed on the MFL were observed to have the 

correct identification, location, manufacturer, type, and size of fuse. 

iluri ng those reviews 

For each of the 12 selected RIP on hold items that were closed, it was 

determined that corrective actions had been adequate and that any related 

installed items were properly procured. For the three remaining RIP on hold 

items which were still considered as open, the inspection team verified that 

documentation existed to ensure that removal of the related incorrectly 

procured items would occur. 

5.8 IMPELL Review of Seismicallv Sensitive-Electricallv Active (SSEA) Items. 

The inspection team reviewed the status of TVA's corrective actions associated 

with Significant Corrective Action Report WBP8900634SCAY which was initiated 

as the result of a previous material evaluation performed for TVA by IMPELL 

34 

139 
Corporation, between August and October 1989. The SCAR remains open pending 

completion of corrective actions. During the evaluation, SSEA devices in 

stock and installed in safety-related applications associated with 416 TIICs 

were evaluated. As the result of that review, installed SSEA devices 

associated with 56 TIICs were determined to be inadequate. These electrical 

devices had been released from stores inventory for safety related 

applications and were not procured to requirements of the Institute of 

Electrical and Electronics Engineers (IEEE) Standard IEEE-344 1975, "Seismic 

Qualification Requirements for Nuclear Power Generating Stations," nor were 

they subsequently determined to satisfy those requirements. 

also reported to the NRC under Construction Deficiency Report (CDR) 390/91-31. 

Because there can be multiple installations associated with the same TIIC 

number, the 56 TIIC items resulted in a total of 66 separate installations and 

classified under 56 Case Numbers (separate case number for each affected 

TIIC). The affected devices included relays, breakers, motor starters and 

magnetic contacts. 

installations was to be replaced with a properly qualified component. 

This failure was 

When corrective actions are complete each of the affected 

The inspection team was provided a copy of a matrix used by TVA personnel to 

identi'; a specific work document such as a design change notice (DCN), WO, WR 

or other document which was intended to ensure the replacement of each of the 

66 devices. The team reviewed the matrix and determined that TVA's corrective 

actions were complete for all but three of the 66 devices which had originally 

been determined to be inadequate. In order to assess the adequacy of TVA's 

corrective actions in this area, TVA was requested to provide documentation to 

support corrective actions for the following five devices: 

Case 14 - 1 and 322172-1 which were not procured with any unique nuclear 

requirements for use as basic components. 

motor starter had been requisitioned under TVA Form 575-628602989 and 

installed under MR 498392. That starter had been installed in 1MCC-215- 

1B2/2B-B for the EDG Room 1B-B Exhaust Fan, 1-MTR-30-453B. 

verified that the affected motor starter had been replaced with a 

qualified device under WO 90-16222-63. 

Identified motor starters procured under TVA Contracts N3-865- 

TVA had determined that one 

It was 

Case 16 - for installation under MR 412375. 

Contract N3-632-1 and was never certified as a basic component for use 

in a Class 1E circuit. However the MR which was worked during 1984 to 

troubleshoot an oi 1 pressure fluctuation on Chi 11 er 0-CHR-031-0049/2-BY 

does not indicate that the relay was actually used in the system as a 

spare part. 

was canceled when TVA determined that the chiller control panel did not 

contain this type relay. Additionally TVA personnel inspected the 

associated chiller control panel and did not identify any evidence that 

re1 ay rep1 acement had occurred. 

Identified a relay requisitioned under TVA Form 575-628501768 

The relay had been procured under TVA 

MR A622272 had been written by TVA to replace the relay but 

Case 23 - 

2578-1 which was not procured as a qualified Class 1E device. 

determined that the circuit breaker had been requisitioned under TVA 

Form 575-628706537 and installed under MR 511626. That circuit breaker 

Identified a circuit breaker procured under TVA Contract N3- 

TVA had 

35 

140 
had been instal 1 ed in 1MCC-214-A1-A Compartment 2D. Thi s appl i cat i on 

requires the use of a Class 1E circuit breaker so that the nonsafetyrelated primary makeup water pump 1A circuity may be isolated from the 

Class 1E power source during an accident. It was determined that 

replacement of the affected circuit breaker had been included under DCN 

M-19745-A as implemented by WP D-19745-12. Additionally the inspection 

team verified that WP 0-19745-12 was field complete on October 25, 1993. 

for installation under MR 408313. The relay had been procured under TVA 

Contract 303615 and was never certified as a basic component for use in 

a class 1E circuit. However the referenced MR is not available on 

microfilm and could not be traced. Contract 30361.5 also stated that the 

subject relay had been procured for use in the Incore Instrument System. 

TVA reviewed the Nuclear Instrumentation System (PIIS), Incore 

Thermocouple System (ITS) and Incore Flux Mapping System (IFMS) to 

determine if this type relay was used in any safety-related 

applications. 

electronic processing portion of the ITS and the source and intermediate 

range portions of the NIS have been replaced since TVA Form 575N was 

issued. 

portions of the NIS. 

any record or work document which indicated that the relay was actually 

used in the system as a spare part, that this issue had been adequately 

di sposi ti oned. 

0 Case 43 - Identified a rel'ay requisitioned under TVA Form 575-628504307 

The IFMS is not considered safety-related and the 

Additionally no relay of this type exists in the power range 

TVA determined that since they did not identify 

0 Case 54 - Identified a circuit breaker procured under TVA Contract 

468257 which was not procured as a qualified Class 1E device. TVA had 

determined that the circuit breaker had been requisitioned under TVA 

Form 575-628705034 and instal 1 ed under MR 413626. That circuit breaker 

had been installed as the power supply for the 120 Vac vital invertor 2- 

IV which supplies power to the vital instrument board. It was verified 

that the affected circuit breaker had been replaced with a qualified 

device under WO 90-16570-29. 

As a result of this review, the inspection team determined that corrective 

actions associated with the above selected cases were adequate. 

will receive further review during the followup review associated with CDR 

This issue 

390/91-31. 

5.9 RIP QA Level I11 Item Review 

The inspection team reviewed "Final Report, Rep1 acement Items Program Task 4A, 

Revision 1, issued June 30, 1994," which was prepared by TVA. The scope of 

this task was to investigate the QA Level I11 population in TVA's computerized 

Material Management System (MAMS) to determine if items had been installed 

into safety-related applications. As part of this effort, TVA performed a 

sample review of QA Level I11 items to determine if these items had been 

inadvertently installed into safety-related applications and thus negated 

environmental or seismic qualifications or degraded the capability of the host 

equipment to perform its intended safety function. 

36 
‘I 

TVA used MIL-STD-105E as a guide for selection of the sample lot and a 

predetermined rejection rate of no more than five. 

percent level of confidence that at least 95 percent of the installations were 

not adversely affected by the replacement parts installed as QA Level 111. 

This review was performed on a sample lot of 315 items out of a total of 5215 

QA Level I11 items that were installed in the plant. Because there can be 

multiple installations associated with the same TIIC number, the 315 TIIC 

items selected as the sample population resulted in 1417 installations. 

The inspection team reviewed the listing of 315 sample QA Level I11 TIIC items 

that had been the subject of this review. The sample list included items such 

as light bulbs, conduit, identification markers, etc., that would not have 

been expected to serve any safety-related application along with items that 

would have required detailed reviews to ensure that the items were not used in 

safety-related applications. The team selected several items from TVA’s 

sample listing and requested that they provide documentation to establish the 

extent of their review effort. 

items selected, TVA had ensured that the related items had not been installed 

in a safety-related application. 

This would provide a 95 

The team determined that, for each of the 

As the result of TVA’s review, three items were found to have been installed 

into safety-related applications but were declared to be of indeterminate 

quality because the documentation was not sufficient to demonstrate 

traceability. As a result, the items were replaced with acceptable piece 

parts and the team agreed that this rejection rate met the acceptance 

criteria. 

had been adequate to ensure that the 95 percent confidence level had been 

achieved. 

Based on the above review, the team determined that TVA’s review 

6 REPLACEMENT ITEMS INSTALLED BY PREVIOUS CONSTRUCTION ACTIVITIES 

REPLACEMENT ITEMS PROGRAM (RIP) (TI 2512/27 , 35065, 35746, 35747) 

The RIP CAP plan states that its scope includes the review and evaluation of 

spare or rep1 acement parts from the construction warehouse inventory installed 

in WBN by the construction organization. The purpose of the NRC’s review of 

this area during this inspection was to provide an independent determination 

of the scope and depth of the TVA review completed thus far in the area of 

construction installed materials at WBN. It was not intended to evaluate the 

disposition of TVA’s findings but rather to summarize TVA’s review and provide 

NRC management with information necessary to evaluate the sui tabi 1 i ty of 

materials used in the construction of WBN. 

In its initial response to concerns raised regarding the adequacy of this 

portion of the RIP CAP plan, TVA contracted Digital Engineering, Incorporated 

(DE), to develop the methodology and conduct a review of this material. 

results of this review were published in DE’S preliminary report, “WBN 

Construction RIP, Phase 1 Summary Report,” dated October 3, 1989. On 

August 17, 1991, TVA Nuclear.Quality Assurance (NQA) issued the report of its 

audit of RIP, QA Audit Report No. WBA92211. The report contained three 

recommendations (numbers 2, 3, and 4) regarding construction installed 

materials. 

had only addressed the ledger card recorded items (about 55,000) and not the 

The 

Recommendation 2 was that because the review completed thus far 

37 

142 
so-called direct charge items (about 19,000), RIP should review the direct 

charge items. 

inventory adjustments made without issues on Form 575Ns. 

that associated 575Ns be sampled for verification of the validity of the 

information thereon. 

Recommendation 3 was that RIP should disposition ledger card 

Recommendation 4 was 

In a subsequent audit report , NQA Assessment Report NQA-IJB-92-028, dated 

September 23, 1992, NQA accepted the RIP proposal, in response to 

Recommendation 2 of WBA92211, to sample a homogeneous group of direct charge 

items listed in their "System 3" computer inventory system, stating that the 

completion assurance group will follow up on the acceptability of the sample 

and its results. On the basis of procedural controls in--place requiring the 

use of Form 575N for all material issues, NQA accepted the RIP explanation 

that the no-issue inventory adjustments were most likely the result of 

breakage or theft. 

the information contained on the forms because the response addressed only 

maintenance installed items and not construction installed items. 

Finally, NQA rejected RIPS taking credit for confidence in 

The revised response to Recommendation 4 agreed to be provided by RIP as a 

result of this disapproval , was submitted in a memorandw from the TVA site 

vice president, dated October 19, 1992, which transmitted to R.F. Driscoll , the WBN responses to the recommendations in NQA-WB-92-028. In the revised 

response (Attachment F to the memo), several additional factors in support of 

substantiating the validity of Form 575N information were presented. Finally, 

on January 25, 1993, RIP issued "Final Report, Replacement Items Program 

Task 3, Construction Installed Replacement Items, Revision 1." Most notably 

this report provided the results of the review of direct charge materials as 

we1 1 as additional support i ng documentat i on on 1 edger card materi a1 s . 

On the basis of this review, the inspection team did not find any obvious 

significant deficiencies in the logic, methodology, or conclusions of the 

final report. However to date, the formal review by TVA of this report has 

not been documented. Since TVA needs to make a formal, final submittal of its 

review and proposed final disposition of this issue, this item is identified 

as Inspector Followup Item 94-201-15. 

7 CURRENT WAREHOUSE INVENTORY-MATERIAL IMPROVEMENT I PROJECT (MIP) 

(TI 2512/27, 35065, 35746, 35747) 

7.1 Implementation Review of the MIP 

As a part of the RIP CAP plan, TVA committed to the sanitization of the 

current inventory of materials purchased prior to June 5, 1991, before these 

materials were released for installation in the plant. Sanitization was the 

term used to define a process of review for technical adequacy and re-receipt 

inspection of these materials. This effort was referred to in the RIP CAP as 

the Material Improvement Project or "MIP." 

conducted to review the implementation of the MIP and was conducted primarily 

by an evaluation of the adequacy of the sanitization packages which had been 

created during the process. 

This part of the inspection was 

38 

143 
The sanitization packages included the following information: a wal kdown sheet 

which identified the on hand stock; a PEG package which provided the 

engineering evaluation of the technical and QA requirements for the material; 

the original contract/purchase order/request for del ivery; the old receipt 

inspection records, if avail able; COCs/CMTRs and other vendor documentation; 

the sanitization receipt inspection; a plant issue history from the MAMS 

computer data base for the material; and verification that the vendor was on 

the ASL, as applicable. The inspection team conducted an independent review 

of all of these documents to verify technical adequacy and inclusion of 

appropriate regulatory requirements. 

7.1.1 Acceptable Sanitization Packages 

Many of the items reviewed by the team were surplused during the sanitization 

process. For each of ‘these items the inspection team verified that there was 

an entry on the computer data base, which indicated that the material had in 

fact been surplused. The team also verified that there was no issue history 

for the material, and where possible reviewed inventory sheets, which verified 

that the material had been moved to TVA’s surplus yard. 

packyes were determined to be adequate: 

The following 

SANM 14347 SANM 13295 SANM 13234 SANS SP0070 

SANM 13199 SANM 12585 SANM 11030 SANE 2949 

SANM 13289 SANM 13968 SANM 15093 SANE 3092 

SANM 14061 SANM 11545 SANM 14345 SANE 3648 

SAN SP0023 SANE 4623 SANE 2130 SANE 4480 

SANE 3368 SANE 3973 SANE 3934 SANE 3755 

SANE 4123 (partial surplus) 

The following items reviewed by the inspection team successfully completed the 

sanitization process and were maintained in inventory by TVA: 

SANM 13126 

SANM 15684 

SANM 10337 

SANM 10901 

SANE 4468 

SANE 3718* 

SANE 3736 

SANM 12193 

QA level I 

QA level I 

QA level I 

QA level I 

QA level I 

QA Level I 

QA Level I1 

QA level I1 

Pump collar 

Ball bearings 

Valve pin 

Metal Bar 

Penetration repair kits 

ASCO sol enoi d Val ves 

Gould fuse 

Structural angle 

* Determined acceptable after TVA corrective action reported subsequent to 

inspection 

The following items had been rejected during the sanitization process. Review 

of these packages by the inspection team determined that the materials had an 

issue history, and therefore, required further engineering evaluation. 

Further review of additional documentation determined that these materials had 

been referred to engineering for an evaluation of the installed items: 

39 

144 
SANE 4601 QA level I1 Digital microcircuits 

SANE 3407 QA level I1 Rotary control switches 

SANE 4013 QA Level I1 Resi stors (TI IC AQD7SOB) 

7.1.2 Unacceptable Sanitization Packages 

The inspection team's review of sanitization packages identified the following 

def i ci enci es : 

e SANE 2953, SANMs 13888. 14894 and 14175 - These sanitization packages 

rejected stock material and, due to a material issue history, required 

an engineering evaluation of the specific installation application in 

the plant (i.e., the items had been referred to RIP for evaluation). 

The team requested copies of these engineering evaluations (PPSP 

packages) for review and was advised that no packages had been developed 

for these items. Additionally, the team's review of the RIP computer 

data base determined that the items were not carried on this data base 

as outstanding work items for the RIP project. Note: TIIC numbers in 

these packages requiring engineering eval uati on were as fol 1 ows : SANE 

2953 (TIIC Numbers AKN-613T, AKN-712QY AMN-341QY AMN-414LY AKN-912G); 

SANM 13888 (TIIC Numbers AVX-142JY BAY-349Q); SANM 14894 (TIIC Number 

AVF-311X); and SANM 14175 (TIIC Number BRK-U44A, Heat # M7780). 

This issue was identified to TVA as a problem area requiring an extent 

of condition evaluation and resolution. As a result of this finding, 

TVA initiated WBPER 940351. As a part of the corrective action to the 

PER, TVA reviewed 885 sanitization packages in an attempt to identify 

the extent of the problem. 

identified that were to have been referred to the RIP project for 

further evaluation. Of the 158 referred packages, 84 were not found on 

the RIP computer data base and no additional engineering reviews had 

been performed. TVA' s extent of condition, eval uati on and corrective 

actions in this area were ongoing at the completion of this inspection. 

Of the 885 packages, 158 packages were 

TVA Procedure SSP-lO.B, "Materials Improvement Project," Revision 5, 

requires that: "Material evaluated as unacceptable for use at WBN will 

be investigated to determine if any of the material has been issued from 

inventory. If none has been issued, then future issues will be denied. 

Material issued, but not installed, will be recalled to the warehouse. 

In both cases, the material will be surplused, dedicated, or discarded. 

If an item has been installed, all pertinent information will be logged 

on Appendix D and forwarded to the appropriate Engineering Group for 

evaluation and resolution." Since TVA personnel did not follow this 

procedure, previously installed material was not evaluated for 

suitability. Failure to follow procedures is identified as Deficiency 

94-201-D8. 

e SANE 3718 - This package was for solenoid operated valves (SOVs) 

manufactured by the Automatic Switch Company (ASCO) , model NP8320A182E 

(TIIC BEE211G). 

NP-1 valves, environmentally qualified, yet the procurement master data 

sheet (PMDS) was checked ''no'' under 10 CFR 50.49 appl icabil ity (a 

These SOVs are referred to in ASCO's product catalog as 

40 
typical use is MSIV pilot valves). 

not checked and the shelf life was marked "NA." It was not clear where 

one SOV from the each of the two contracts covered was used. The stock 

verification sheet has shop order marked "NA," yet there are shop order 

numbers associated with the SOVs. 

Also, the 10 CFR Part 21 block was 

The printout from MAMS of the associated V300 item history screen 

indicated five SOVs which were surplused or possibly scrapped due to 

damage. 

Test Report AQR67368, Revision 1, possibly by Wyle Laboratories, had no 

associated environmental qualification (EQ) binder number reference, 

e.g., "WBN-SOV-XXX, If and pages 2 and 4 of the ASCO COC were missing. The requested temperature profile attached to the COC was the wrong 

type. It was an EQ accident profile (temperature in containment versus 

time); whereas, the profile intended was the chart of temperatures at 

different locations on the SOV when it is continuously energized, which 

is needed to perform qualified life calculations. 

energies of the limiting components and their bases were not evident. 

The COC for Contract 85PN7-360966, Shop Order 65076N (quantity 4), 

supplied through M&L did not reference TVA's PO nr-ber, yet this COC was 

accepted. The other contract in the package, SOVs supplied through 

another ASCO distributor, Power & Control, Incorporated, did have the 

TVA RD number on the COC. 

Attachment A to release RD 111893 of IQT Contract 89NNR-34440BY 

The activation 

However, in another SAN package, TVA rejected a COC for this deficiency. 

In response to these concerns, and subsequent to the inspection, TVA 

submitted copies of the missing COC pages and first page with the proper 

contract reference it had obtained from ASCO. The correct temperature 

profile diagram was on the missing page 4. Also submitted was a new PEG 

package (9400042184) with which TVA supplemented SANE 3718 with the 

required information. TVA also reported that the deficient PMDS is 

under review to incorporate the proper requirements. 

actions were considered acceptable. 

These corrective 

0 SANE 3845 - This package was for QA Level I, Dunham-Bush motor starters, 

TIIC AQQ696J. The package indicated that the material was retained, yet 

inventory adjustment sheet number 62-92-357 indicated the materi a1 

"failed SAN" and ''code 95 TIIC AQQ696J.I' It was also not clear from the 

package how the items were dedicated by the supplier, Southern Testing 

Services (STS), and how production units sent to WBN were verified to be 

similar to the STS seismic test specimen. 

In response to these concerns, TVA reported subsequent to the inspection 

that three starters were procured under contract 90LK-85122B from STS. 

One was used for the electrical board room chiller (Form 575N 254779) 

and when found not operating properly, it was replaced by another one 

from stock (Form 575N 245741). PPSP 35545 identified the rejected 

starter as being put on RIP hold and then surplused. TVA reported that 

the remaining unit was still being held in "indeterminate material" and 

would require a PEG package to put it in stock. 

inspection team's questions on dedication and seismic qualification, TVA 

provided a copy of STS Report S254-RP-02, dated October 5, 1990. This 

In response to the 

41 

146 
report consisted of an analysis for additional starters in question for 

seismic qualification similarity to starters seismically tested by STS 

for WBN as reported in STS Report S254-RP-01, dated September 21, 1990, 

under the same TVA contract. 

visual examination of the starters and the fact that they were of the 

same part number as the original units supplied as well as the seismic 

test specimen and had been determined to have been all manufactured 

within a 3-week period in the same factory. 

The similarity analysis was based on 

Concluding that the subsequently supplied starters were similar, or even 

“like-for-like” to those qualified on the basis described above is 

consistent with the staff position promulgated in NRC GL 91-05. 

However, the similarity analysis was intended only for seismic 

qualification. 

process, and is only one critical characteristic in a dedication 

process, it cannot substitute for a complete dedication. Although STS 

energized a sample starter with nominal 240 Vac during some of the 

seismic test runs, verification of other critical characteristics 

considered standard by the industry, such as insulation resistance, 

individual pole resistance, full load hold-in capabilitv, overload trio 

performance, and pul l-in and drop-out voltages (degraded voltage 

conditions) was not documented in the packages reviewed by the team. 

Since seismic qualification is a design verification 

In addition, the original seismic report contained a notice of anomaly 

describing 10- to 12-millisecond (ms) chatter in the normally closed 

auxiliary switch of the seismic test specimen starter during the first 

run in which it was deenergized and in one of four horizontal 

orientations. 

that there had been no other runs with this or another starter in the 

same orientation (apparently the unit’s most seismically sensitive axis) 

in which the NC auxiliary switch did not chatter in order to rule out 

common mode failure. 

criterion had been established for this performance parameter, yet STS 

determined, apparently unilaterally, that this condition would not 

adversely affect the starters’ performance. 

The description of the test and the procedure indicated 

It was not clear that any acceptance or rejection 

As far as could be determined from the documents reviewed, this 

determination was apparently acceptable to TVA because the WBN Civi’l 

Engineering group had accepted the report for seismic qualification, yet 

it was not clear how or whether the 10- to 12-111s contact bounce of the 

NC auxiliary switch had been evaluated for its effect on the safety 

function(s) of the starter, or its parent equipment or system as 

installed at WBN. Accordingly, this issue is identified as Unresolved 

Item 94-201-U3. 

8 EXIT MEETING 

On July 20, 1994, the inspection team conducted an exit meeting which 

summarized the inspection scope and the findings. 

described the areas inspected and discussed in detail the inspection results. 

The following people were in attendance. 

The inspection team 

42 
- TVA 

D. Nunn, Vice President 

J. Scalice, Vice President 

R. Johnson, Acting Engineering Manager 

W. Clothier, Jr., Manager, Nuclear Stores 

R. Brown, Licensing 

T. Whittemore, Materials Program Administrator 

W. Rogers, PEG Manager 

J. Lewis, Material Supervisor 

V. Patuzzi, Quality Assurance Specialist 

J. Seeley, RIP Project Manager 

P. Pace, Compliance Licensing Manager 

D. Malone, Manager 

B. Schofield, Licensing Manager 

L. Spiers, Site Quality Assurance Manager 

G. Pannell , Manager Speci a1 Programs 

W. Elliott, Engineering and Modifications Manager 

D. Koehl, Technical Support Manager 

D. MooA:~, Plant Manager 

P. Phillips, Senior Technical Lead Auditor 

M. Fecht, Manager Quality Programs 

J. Overturf, RIP Project Manager 

United Enerqv Services Corporation 

F. Yurich, Consultant to TVA 

W. Craig, Consultant to TVA 

Nucl ear Requl atorv Commi ssi on 

R. Pettis, Jr., P.E., Team Leader, NRR 

U. Potapovs, Section Chief, NRR 

P. Frederickson, Construction Branch Chief, RII 

L. Campbell, Reactor Engineer, NRR 

S. Alexander, EQ and Test Engineer, NRR (via telephone) 

J. Peralta, Operations Engineer, NRR 

R. Gibbs, Reactor Engineer, RII 

W. Bearden, Project Engineer, RII 

P. Van Doorn, Senior Resident Inspector-Operations, WBN 

G. Walton, Senior Resident Inspector-Construction, WBN 

43 

148 
Appendix A 

Li st of Deficiencies 

Number 

94-201-Dl 

94-201-DZ 

94-20 1 -D3 

94-201 -D4 

94-201-D5 

94-20 1-D6 

94-201-D7 

94-201-08 

94-20 1-D9 

Title 

Weaknesses in Correct i ve Action Program 

Implementation - PEG Package 9400018340 

Inadequate Translation of Design Basis 

Requirements into Design Output Documents 

Which Resulted in an Inadequate Review for 

Sui tabil i ty of Appl ication for QA Level I 

Molded Case Circuit Breakers - PEG Package 

9400002182 

Failure to Ensure that QA Level I Molded 

Case Circuit Breakers Meet Procurement 

Requirements - PEG Package 9400002182 

Inadequate Review for Suitability of 

Appl ication and Failure to Verify 

Compl i ance to the Procurement Requirements 

for GE 125-Vdc Trip Coils for Magne-B1 ast 

Circuit Breakers used in Class 1E 6900V 

Shutdown Boards - PPSP 27732 

Failure to Verify the Procurement 

Requirements of Telemechanique Starters 

Eva1 uated Under PPSP 27739 

Inadequate Commercial Grade Dedication 

of a GE Magne-Blast Circuit Breaker 

Spring Charging Motor - PPSP 27724 

Inadequate Commercial Grade Dedication 

of a Westi nghouse Direct Trip Actuator 

for a 480-Vac Air Circuit Breaker 

Evaluated Under PPSP 28973 

Failure to Follow TVA Procedures for 

Previously Installed Items Not Captured 

in the RIP Data Base - SANE 2953 

Inadequate TVA Procedures Associated With 

the Evaluation of PEG Package 9400002182, 

PPSP 27975, PPSP 27732 and PPSP 28973 

A- 1 

149 

Section/Paqe 

4.1/9 

4.1/12 

4.1/14 

5-2/21 

5.2/23 

5.3/26 

5-3/27 

7.1.2/40 

4.1/12, 5.2/19, 

5.2/21, 5.3/28 
Number 

94-201-U1 

94-201-U2 

94-201-U3 

Appendix B 

List of Unresolved Items 

Ti tl e 

QA Level I1 PEG EPEG 920013 for GE 

125-Vdc Closing Spring Charging Motors 

That Must Operate More Than Once to 

Perform Their Safety Function 

Questions Relating to Dedication 

Testing of Limi torque Motors Eva1 uated 

Under PPSP 20406 

Questions Concerning QA Level I 

Dunham-Bush Motor Starters Evaluated 

Under SANE 3845 

Section/Paqe 

4.1/16 

5.2120 

7.1.2/42 

B- 1 

150 
Appendix C 

List of Inspector Fol lowup Items 

Number Title Section/Paqe 

94-201-11 Several Issues Requiring Amendment to PEG 4.1/8 

Package 9400018459 for Amphenol Connectors 

94-201 - I2 Documentation Contained in PPSP 27975 for 5.2/20 

a GEMAC Power Supply Did Not Contain 

Expl anati ons of the Technical Concerns 

Identified by TVA 

94-201-13 Failure to Perform an Adequate Technical 5.3/25 

Evaluation for Lockwire Used in Certain 

Limitorque Operators - PPSP 28039 

94-201- I4 Questions Concerning TVA's Process S. 5/29 

Specification 4.M.l.l and PF Material 

Purchase Specifications Regarding 

Control of Contaminants 

94-201- I5 Questions Concerning the Evaluation of 6/38 

Rep1 acement Items Instal 1 ed by Previous 

Construction Activities (RIP) 

c- 1 

151 
Appendix D 

Glossary of Specific Terms 

BKR 

BFN 

CAP 

CCRS 

COMP 

CSST 

DCN 

DE 

EPEG 

EPG 

1 QT 

ITS 

I FMS 

MI 

MR 

MAMS 

MTR 

MIP 

MFL 

MHDB 

NIS 

PER 

PEG 

PPSP 

PSS 

PMDS 

QA 1 

RT L 

RIR 

RIP 

RIMS 

SAN 

SANE 

SANM 

SSEA 

SQN 

SSP 

SA I 

TIIC 

TEF 

UNID 

V300 

WBN 

wo 

WR 

575N 

Breaker (UNID Abbrevi ation) 

Browns Ferry Nuclear (P1 ant) 

Corrective Action Program 

Computerized Cable Routing System 

Switchboard Compartment (UNID Abbreviation) 

Combined Station Supply Transformer 

Design Change Notice 

Digital Engineering (RIP Contractor) 

Equipment Procurement Engineering Group 

Engineering Procurement Group (Contract Procurement Group) 

Indefinite Quantity and Time (Contract) 

Incore Thermocouple System 

Incore Flux Monitoring System 

Maintenance Instruction 

Material Requisition 

Material Management System 

Motor (UNID Abbreviation) 

Materi a1 Improvement Project 

Master Fuse List 

Material History Data Base 

Nuclear Instrumentation System 

Probl em Eva1 uat i on Report 

Procurement Engineering Group 

Previous Procurement Substantiation Process (RIP) 

Power Service Shops (TVA Repair Facility, Mussel Shoals, LA) 

Procurement Master Data Sheet 

Quality Assurance Instruction 

Release Tracking Log (PPSPs) 

Receipt Inspection Report 

Rep1 acement Items Program 

Reccrd Information Mana ement System 

Sanitization (MIP) 

El ectri cal Sanitization 

Mechanical Sanitization 

Seismically Sensitive-E 

Sequoyah Nuclear 

Site Standard Practice 

Package 

Package 

ectri ca ly Active 

Site Administrative Instruction 

TVA Item Identification Code 

Technical Evaluation Form (SSP-10.05 and -lO.C) 

Unique Identifier (Equipment Tag Number) 

MAMS Materi a1 Hi story Screen 

Watts Bar Nuclear 

Work Order 

Work Request 

Materi a1 Requi si tion/Issue Form 

D- 1 

152 
Selected Bulletins, Generic Letters, and Information Notices 

Concerning Adequacy of Vendor Audit:; and 

Qual i ty of Vendor Products 

DOCUMENT TYPE AND NUMBER 

Information Notice 94-71 

Information Notice 94-76 

Information Notice 94-85 

Information Notice 94-86 

TITLE 

Degradation of Scram Solenoid Pi1 ot Valve 

Pressure and Exhaust Diaphragms 

Recent Fai 1 ures of Chargi ng/Safety Injection 

Pump Shafts 

Problems With the Latching Mechanisms in Potter 

and Brumf i ei Ci RlO-E3286-2 Re1 ays 

Legal Actions Against Thermal Science, Inc., 

Manufacturer of Thermo-Lag 
CORRESPONDENCE RELATED TO VENDOR ISSUES 

154 
UNITED STATES 

NUCLEAR REGULATORY COMMISSIOlN 

WASHINGTON, D.C. 20555-0001 

November 4, 1994 

Mr. Jim Fitzwilliam 

Qual i ty Assurance Manager 

Nova Machine Products Corporation 

P.O. Box 30287 

Middleburg Heights, Ohio 44130 

SUBJECT: REQUEST FOR INTERPRETATION: NRC INFORMATION NOTICE NO. 86-21 

Dear Mr. Fi tzwi 11 i am: 

By letter dated August 19, 1994, you requested the U.S. Nuclear Regulatory 

Commission (NRC) to provide an interpretation of Information Notice (IN) 

No. 86-71. "Recognition of American Society of Mechanical Engineers 

Accred'tation Program for N Stamp Holders," dated March 3:1, 1986, including 

its Supplements 1 and 2, issued on December 4, 1986, and April 16, 1991, 

respectively. IN No. 86-21 and its supplements provided no new requirements, 

but provided guidance and clarification on existing NRC positions that may be 

used by purchasers of certain American Society of Mechanical Engineers Boiler 

and Pressure Vessel Code (ASME Code) , Section 111, "Rules for Construction of 

Nuclear Power P1 ant Components, 'I i tems to verify that the ASME-accredited 

suppliers of the items are effectively implementing their quality assurance 

(QA) program. 

Your August 19, 1994, letter to the NRC identified a method, other than those 

discussed in IN No. 86-21, Supplement 2, for verifying that an ASME-accredited 

supplier's QA program is being effectively implemented antl you requested an 

interpretation by the NRC for its acceptability. 

understanding of your interpretation request. 

The following is our 

I. 

11. 

BACKGROUND 

Purchase orders (POs) received by Nova from utilities (NRC licensees) 

and support groups such as Architect Engineers, Nuclear Steam System 

Suppliers, Original Equipment Manufacturers, and pump and valve 

manufacturers, invoke requirements for Nova to veriffy that 

ASME-accredited suppliers are effectively implementing their QA program. 

METHOD PROPOSED BY NOVA FOR VERIFYING THAT ASME-ACCREDITED SUPPLIERS ARE 

EFFECTIVELY IMPLEMENTING THEIR OA PROGRAMS 

1. Through a review of the ASME publication, "Companies Holding 

Nuclear Certificates of Authorization," which 1 ists suppliers 

accredited by ASME and their scope and expiration date, Nova will 

establish that ASME has performed an audit antl accepted a 

supplier's QA program and issued the supplier a certificate. 

155 
J. Fitzwilliam -2- 

2. The supplier is then placed on Nova's approved supplier's 

list (ASL) for the scope identified in the ASME publication with a 

restriction that prior to procurement from the supplier an 

implementation audit will be performed. 

Nova is informed that Nuclear Procurement Issues Committee (NUPIC) 

performed an audit at the supplier and issued the supplier a NUPIC 

audit report \,erifying that the supplier is effectively 

implementing its the ASME-accredited scope activities. 

Nova requests and the supplier forwards a copy of the NUPIC audit 

report. The audit report identifies the audit participants, the 

ASME certificate number(s), QA program revision, and details the 

audit results. 

3. 

4. 

5. Nova evaluates the NUPIC audit report and concliides that the NUPIC 

audit verified the supplier is adequately implmenting its QA 

program for the ASME-accredited scope. Nova maintains its 

evaluation of the NUPIC audit report and the audit report. 

6. Based on its evaluation, Nova lises the NUPIC audit as the means to 

verify that the supplier is effectively implementing its 

ASME-accredited scope QA program and removes the restriction from 

its ASL to perform a QA program implementation audit of the 

supplier. 

of the supplier is that a NUPIC member, an NRC licensee, has 

performed the .verification audit and, therefore, the intent of the 

requirements discussed in IN 86-21 have been met. 

Nova's justification for not performing its own audit 

111. INTERPRETATION REQUEST 

Nova's specific question presented to the NRC in its August 19, 1994, 

letter is as follows: 

"Under the above noted condi ti ons , can Nova procure ma teri a1 s cert i f i ed 

under an ASME Certificate holder's program and, meet the intent of 

IN 86-21 without a) performing an audit of the ASME Certificate holder's 

program, or b) performing any type of program verification?" 

IV. NRC EVALUATION OF NOVA'S QUESTION 

BACKGROUND 

Criterion VI I , "Control of Purchased Materi a1 , Equipment , and Services , of Appendix B of Part 50 to Title 10 of the Code of Federal Requlations 

(10 CFR) requires, in part, that measures shall be established to assure 

that purchased material conform to procurement documents and 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 or service. 

156 
J. Fitzwilliam -3- 

Additionally, in NRC Regulatory Guide (RG) 1.144, "Auditing of Quality 

Assurance Programs for Nuclear Power Pl ants, I' Revision 1 , dated 

September 1980, the NRC has provided a position oil the use of joint 

audits. Regulatory Position C.5 of RG 1.144 provides additional 

guidance on performing audits and supplements the provisions of 

Section 1.3 of ANSI/ASME N45.2.12-1977, "Requirements for Auditing of 

Qual i ty Assurance Programs for Nuclear Power P1 ants. 'I 

Section 1.3 of ANSI/ASME N45.2.12-1977 states, in part, that the work of 

establishing practices and procedures and providing the resources in 

terms of personnel, equipment and services necessary to meet the 

requirements of this standard may be delegated to other organizations 

and such delegation shall also be documented. 

of RG 1.144 states, "Where more than one purchaser buys from a single 

supplier, a purchaser may perform an audit of the supplier on behalf of 

more than one purchaser in order to reduce the number of external audits 

of the supplier. The results of this audit should Se distributed to all 

purchasers for whom the audit was conducted." 

Regulatory Position C.5 

Also, RG 1.144 acknowledges that differences may exist between the 

requirements of ANSI/ASME N45.2.12 and Section I11 of the ASME Code. 

Regulatory Position C.2 of RG 1.144 addresses these differences and 

states, in part, that subject to the exceptions of this regulatory 

position, ANSI/ASME N45.2.12 should be used in conjunction with Section 

111 of the ASME Code for auditing QA programs where the ASME Code does 

not address activities covered by ANSI/ASME N45.2.12. 

NRC RESPONSE 

Nova's proposed method for verifying that ASME-accredited suppliers are 

effectively implementing their QA program does not appear to meet the 

requirements of Appendix B to 10 CFR Part 50, the provisions of 

ANSI/ASME N45.2.12-1977, or the guidance provided in Regulatory 

Position C.5 of RG 1.144. 

the supplier would have to perform the following. 

In order for a supplier to use a NUPIC audit 

e 

e 

The supplier would have to delegate to NUPIC the responsibility to 

perform audits on its behalf and this delegation would have to be 

documented as provided for in Section 1.3 of 

ANSI/ASME N45.2.12-1977. 

The results of the NUPIC audit should be distributed to the 

supplier using the audit by NUPIC, not the company audited by 

NUPIC, as provided for in Regulatory Position C.5 of RG 1.144. 

157 
J. Fitzwilliam -4- 

Should you have any further questions, please contact Mr. Larry L. Campbell of 

this office at (301) 504-2976. 

1 

Si ncerel y , 

Robert M. Gallo, Chief 

Special Inspection Branch 

Division of Technical Support 

Office of Nucl ear Reactor Regul ati on 

158 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 2055MW)(ll 

November 28, 1994 

Mr. Roger F. Reedy, President 

Reedy Associates, Inc. 

15951 Los Gatos Blvd., Suite 1 

Los Gatos , Cal i forni a 95032 

SUBJECT: REQUEST FOR INTERPRETATION: 10 CFR PART 50, APPENDIX B, INTERNAL 

AUDITS PERFORMED BY VENDORS 

Dear Mr. Reedy: 

We have reviewed your November 2, 1993, letter to the U.S. Nuclear Regulatory 

Commission (NRC) that provided additional information with respect to your 

previous inquiry to the NRC, dated July 27, 1993. 

opinions in your November 2, 1993, letter concerning the degree of self audits 

(internal audits) necessary to satisfy the requirements of Appendix B to 

Part 50 of Title 10 of the Code of Federal Requlations (10 CFR). 

Additionally, in this letter you stated that Appendix B does not state that 

Criteria XVIII, "Audits," was mandatory for consulting firms such as yours. 

Also, in your November 2, 1993, letter you proposed an alternative in-lieu of 

internal audits performed by your firm, yi~. , "1 icensees should be able to 

accept an audit of a vendor performed by another licensee, provided the audit 

results are reviewed and accepted by the first licensee prior to procuring 

products or services from this vendor. (Please note that these audit results 

are also reviewed and evaluated by us as equivalent to internal audits)." Our response to your inquiry is as follows. 

enclosure to this letter supports our response to your inquiry. 

You expressed specific 

The information provided in the 

Licensees are responsible for determining the quality and technical 

requirements applicable for a contractor's safety-related design service. 

Certain design services may require extensive quality assurance design 

controls through all stages of development and implementation while others may 

require only a limited quality assurance involvement in selected phases of 

development and implementation. 

controls to be used in safety-related design services which it contracts for 

in its procurement documents for such services. 

The 1 icensee specifies the quality assurance 

It is the NRC staff's expectation that (a) where the licensee has determined, 

pursuant to Criteria I, 111, and IV of Appendix By that the activities 

performed by its contractor must be subjected to an internal audit by the 

contractor in accordance with Criterion XVIII, and (b) the licensee has 

imposed such compliance on its contractor through the licensee's procurement 

documents, then the contractor is responsible for ensuring the required 

internal audits are performed. Contractors performing safety-related design 

services that have accepted procurement documents that invoke tke requirements 

of Appendix B to 10 CFR Part 50, or ANSI N45.2, "Quality Assurance Program 

Requi remnts for N~cl ear Power P1 ants, I' or NQA-1 , "Qual i ty Assurance 

Requirements for Nuclear Power Plants," or other similar criteria that require 

audits, should perform internal audits of their safety-related design 

activities. These internal audits would serve to verify that the contractor's 
R. Reedy -2- 

quality assurance program, including design control, for the contractor's work 

scope was being effectively implemented. 

There are two alternatives to a contractor actually performing the internal 

audits. First, the contractor could advise ljcensees that the contractor 

declines to perform the auditing described in Appendix By and that the 

licensee must have total responsibility for compliance with these provisions. 

Alternatively, the contractor can authorize and delegate, in writing (as 

provided for in Criterion I and IV), the responsibility for the performance of 

the internal audits to another organization. However, because the performance 

of internal audits is a safety-related service, the contractor would have to 

assume the responsibility for meeting the requirements of Criterion VII, 

"Control of Purchased Equipment and Services," of Appendix B by ensuring that 

the organization (licensee or contractor) performing internal audits on behalf 

of the contractor was qualified to do so. Additionally, applicable 

requirements from procurement documents issued to the contractor by licensees 

would have to be passed on to the organization performing the internal audits. 

Also, the contractor would have to clearly establish, in writing, that the 

internal audit was performed on behalf of Reedy and the organization 

performing this quality service would have to assume certain responsibilities 

for performing a safety related function. Finally, even though the contractor 

has delegated the performance of the audit to a third party, the contractor is 

nonetheless ultimately resDonsi ble for the adequacy of the audits. 

If, for whatever reason, your firm does not wish to have the responsibility for complying with the auditing provisions of Appendix By you must: 

explicitly advise any and all licensees whom you are providing or intend to 

provide safety-related design services that your firm will not comply with the 

auditing requirements of Criterion XVIII, such that the licensee will have 

total responsibility for meeting the pertinent audit provisions of Appendix B 

and must be responsible for any programmatic and regulatory burden thereof, 

and (b) assure that the licensee's procurement document retaining your firm's 

services does not contain requirements for compliance with the auditing 

provisions of Appendix B. 

(a) 

Should you have any further questions regarding this matter, please feel free 

to contact Mr. Larry L. Campbell at (301) 504-2976 or Mr. Robert Gramm at 

(301) 504-1010. 

Si ncerel y , 

Suzanneh!. B1 ack, Chief 

Quality Assurance and Maintenance Branch 

Division of Technical Support 

Office of Nuclear Reactor Regulation 

Enclosure: As Stdted 

160 
INQUIRY DISCUSSION 

BACKGROUND 

Appendix B to 10 CFR Part 50 identifies the quality assurance (QA) criteria 

for nuclear power plants. The pertinent requirements of this appendix apply 

to all activities affecting the safety related functions of structures, 

systems, and components that prevent or mitigate the consequences of 

postulated accidents that could cause undue risk to the health and safety of 

the pub1 ic: these activities include designing, purchasing, fabricating, 

handling, shipping, storing, cleaning, erecting, installing, inspecting, 

testing, operating, maintLining, repairing, refueling, and modifying. The NRC 

staff evaluates the adequacy of QA programs used by both holders of 

construction permits and operating 1 icenses with respect to the requirements 

of Appendix B to 10 CFR Part 50. Additionally, when the NRC staff conducts 

inspections at nonlicensed vendors, the bases for the inspection are 10 CFR 

Part 21 and the qual i ty and technical requirements contractually imposed on 

the vendor by its customers. 

DISCUSSION ON TEXT OF THE REEDY ASSOCIATES, INC. NOVEMBER 2, 1993, LETTER 

In your November 2, 1993, letter to the NRC, you stated, "Criterion I points 

out that the permit holder or licensee responsible for delegating to others, 

such as contractors, agents, or consultants, the work of establishing and 

executing the quality assurance program, or anv part thereof, but must retain 

responsibility for quality assurance." We agree that Criterion I of Appendix 

B permits the l3censee to delegate the responsibilities to others, however 

Criterion I also states, in part, that the authorization for delegating the 

performance of these responsibilities (activities) shall be clearly 

established and delineated in writing, including both the performing functions 

of attaining quality objectives and the quality assurance function. Criterion 

I continues by stating "The quality assurance functions are those of (a) 

assuring that an appropriate quality assurance program is established and 

effectively executed.. .'I Therefore, a contractor accepting the responsibility 

of performing safety-related activities is also responsible for assuring that 

its QA program controlling these activities is effectively executed. 

In your letter, you also stated, "Criterion IV requires permit holders and 

1 icensees to have measures that assure that applicable regul atory requirements 

necessary to assure adequate quality are suitably included or referenced in 

the documents for procurement of material, equipment, or services. It also 

requires that , to the extent necessary. Procurement documents reauire 

contractors or subcontractors to provide a quality assurance program 

consistent with the provisions of Appendix By" and that "Criterion VI1 

requires permit holders and licensees to assess the effectiveness of the 

qual i ty by contractors and subcontractors at intervals consi stent with the 

imDortan.ce, complexi tv. and qu_a_n_titv of the Dj-oduct or services." Licensees 

determine the importance, complexity, and quantity of services performed by 

1 Enclosure 1 
consultants such as Reedy Associates, Inc. (Reedy), and based on their 

evaluation of the consultant's capabilities and its QA program and 

implementation, invoke the applicable quality and technical requirements in 

their procurement documents. If a licensee invokes the requirements of 

Appendix B in its procurement document, consultants such as Reedy would be 

contractually obligated to comply with all applicable requirements including 

the performance of internal audits required by Criterion XVIII, "Audits." 

A1 though contractors who perform safety-re1 ated design activities are not 

explicitly mandated by Criterion XVIII of Appendix B to perform internal 

audits, as previously discussed, Criterion I, "Organization,"and Criterion IV, 

"Procurement Document Control , 'I 

and impose applicable requirements to assure the adequate quality of the 

delivered item or service. 

by which the licensee specifies the necessary aspects of Appendix B which a 

contractor such as Reedy must comply. 

It is the NRC staff's expectation that (a)-where the licensee has determined, 

pursuant to Criteria I, 111, and IV of Appendix By that the activities 

performed by its contractor must be subjected to an internal audit by the 

contractor in accordance with Criterion XVIII, and (b) the licensee has 

imposed such compliance on its contractor by the licensee's procurement 

documents, then the contractor is responsible for ensuring the required 

internal audits are performed. 

requires that procurement documents specify 

Therefore, procurement documents are the vehicle 

APPLICABLE REGULATORY GUIDES AND ANSI/ASME STANDARDS 

The NRC, through its regulatory guides (RG) has endorsed several ANSI/ASME 

standards, as acceptable methods for meeting the requirements of Appendix B. 

In some instances, the NRC has taken exception to portions of the ANSI/ASME 

standards or provided additional positions not addressed in the standards. 

Most licensees have committed to these ANSI/ASME standards, as endorsed by the 

NRC RGs, as part of their licensing commitments (some licensees may have 

proposed alternatives to parts of these RGs and standards, but in general, 

they form the bases for licensee's QA program commitments). 

ANSI N45.2-1971, "Qual i ty Assurance Program Requirements for Nuclear Power 

Plants," as endorsed by RG 1.28-1972, "Quality Assurance Requirements (Design 

and Construction),n and ANSI N45.2-1977 as endorsed by RG 1.28, Revision 2, 

dated February 1979, provide a general set of requirements for meeting 

Appendix B and refer to the ANSI N45.2 series standards to provide details for 

implementing selective portions of Appendix B. In August 1985, the NRC issued 

Revision 3 to RG 1.28 endorsing ANSI/ASME NQA-1-1983, "Quality Assurance 

Requirements for Nuclear Power Plants," and the ANSI/ASME NQA-1-la-1983 

Addenda. 

N45.2 series standards. 

several of documents that address your inquiry are provided below. 

NQA-1 incorporates the requirements of ANSI N45.2 and selected ANSI 

Although not a complete listing, selective text from 

2 Enclosure I 

162 
Section 1.2, "Scope," of ANSI N45.2-1971 states, in part, that this 

standard applies to the plant owner as well as other organizations 

participating in activities affecting quality. 

similar text (e. g. NQA-1, states in part, that the organization upon 

which this standard, or portions thereof, is invoked shall be 

responsible for complying with the specified requirements). Both 

ANSI N45.2 and NQA-1 impose similar requirements for internal audits. 

Section 11, "Audits," of ANSI N45.2.11-1974, "Quality Assurance 

Requirements for the Design of Nuclear Power Plants," as endorsed by 

RG 1.64, "Quality Assurance Requirements for the Design of Nuclear Power 

Plants," clearly describes provisions for internal audits of 

organizations performing design activities. 

Section 1.2, "Applicability," of ANSI N45.2.12-1977, "Requirements for 

Auditing Qual i ty Assurance Programs for Nuclear Power P1 ants, I' as 

endorsed by RG 1.14f-1980 states, in part, that the requirements of this 

standard apply to both internal and external audits performed by or for 

the plant owner, contractors, or other organizations participating in 

safety-related activities. Section 1.3, "Responsibility," continues by 

addressing the interface aspect where the licensee and a contractor both 

have the responsibility to perform audits and states "Where multiple 

organizational arrangements (for auditing) exist, the interface 

responsibilities for each organization shall be clearly defined and 

documented. In no way shall the performance of audits by an 

organization diminish the responsibility of the audited organization or 

contractor of his designated portion of the quality assurance program or 

the quality of his product or services." 

Other standards have 

Section 1.3, "Responsibility," of ANSI N45.2.12-1977 states, in part, 

"The work of establishing practices and procedures and providing the 

resources in terms of personnel, equipment, and services necessary to 

meet the requirements of this standard may be delegated to other 

organizations and such delegation shall also be documented." 

NRC Regulatory Position C.5 of RG 1.144, "Auditing of Quality Assurance 

Programs for Nuclear Power PI ants, "Revision 1 , dated September 1980, 

states: "Where more than one purchaser buys from a single supplier, a 

purchaser may perform an audit of the supplier on behalf of more than 

one purchaser in order to reduce the number of external audits of the 

supplier. The results of this audit should be distributed to all 

purchasers for whom the audit was conducted." 

Although this position addresses external audits, its provision that the 

audit be performed on behalf of a purchaser would also be applicable for 

internal audits (the audit would have to be performed on behalf of the 

contractor). 

Section 1.2, "Responsibility," of ANSI N45.2.13-1976, "Quality Assurance 

Requirements for the Control of Procurement of Items and Services for 

3 Enclosure 1 
Nuclear Power Plants," as endorsed by RG 1.123-1977, states, in part, 

that the purchaser shall assure that the quality assurance requirements 

incorporated in its procurement documents satisfy the requirements of 

ANSI N45.2 and applicable supplementary standards as applicable to the 

items or services procured. 

an example of a supplier performing design reviews and the degree of 

quality assurance that is expected for compliance with ANSI N45.2. 

example indicates that it is appropriate to expect the suppl ier to 

perform audits. Also, Section 7.1, "General," of Section 7, 

"Verifications to be Performed by the Purchaser," of ANSI N45.2.13, 

requires, in part, that the purchaser's verification activities are not 

intended to relieve the supplier of its responsibilities for 

verification of quality requirements. 

The appendix to ANSI N45.2.13-1976 provides 

The 

The NRC, through its endorsement of standards such as those referenced in this 

letter, has established a position that there is a need for both licensees and 

contractors to perform internal audits of their activities. However, both the 

licensee and its contractors, as provided for in Criterion I of Appendix 6, 

may delegate the performance of internal audits to others. 

Contractors performing safety-re1 ated design services that have accepted 

procurement documents that invoke the requirements of Appendix B to 10 CFR 

Part 50, or ANSI N45.2, or NQA-1, or other similar criteria that require 

audits, should perform internal audits of their safety-related design 

activities. 

quality assurance program, including design control, was being effectively 

implemented for the contractor's work scope. 

These internal audits would serve to verify that the contractor's 

GENERAL COMMENTS 

A licensee's procurement documents for safety-related services normally impose 

the requirements of 10 CFR Part 21 and Appendix B to 10 CFR Part 50 on the 

contractor performing these services and require that these requirements be 

passed onto subcontractors performing safety-re1 ated services for the 

contractor. 

It is not uncommon for contractors, such as Reedy, that employ a small number 

of personnel to qualify and contract an independent organization to perform 

its internal audits. 

4 

164 

Encl OSL 
UNITED STH~ES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 2055m1 

November 15, 1994 

EA 94-049 

Mr. Steve M. Quist, President 

Rosemount Nuclear Instruments , Incorporated 

8200 Market Boulevard 

Chanhassen, Minnesota 55317 

Dear Mr. Quist: 

SUBJECT: NOTICE OF VIOLATION (NRC INSPECTION REPORT 99900271/93-01 & OFFICE 

OF INVESTIGATIONS CASE NO. 4-90-009) 

This 1 etter concerns the Rosemount Nuclear Instruments, Incorporated 

(Rosemount) presentation given to NRC staff at a June 23, 1994, enforcement 

conference and the NRC staff conclusions based on NRC review of and 

deliberation on the circumstances in this matter. 

consideration all of the information obtained and developed by NRC staff from 

1988 to the present. 

In 1993, NRC staff conducted an inspection to follow up on an NRC concern 

regarding an oil-loss problem identified in Rosemount Model 1150-series 

pressure transmitters that could have resulted in undetectable degraded 

operation of Rosemount nuclear safety-re1 ated pressure transmitters. 

that inspection and on the findings of a related Office of Investigation (01) 

investigation, certain of your activities were found to be in violation of 

10 CFR Part 21 (Part 21) requirements, as specified in the enclosed Notice of 

Violation (Notice). The violation is of concern because Rosemount did not 

fulfill its Part 21 responsibilities to its customers between 1984 and 1988, 

when Rosemount became aware of degraded operation of its 1150-series 

transmitters in nuclear safety-related applications due to oil-loss problems in the transmitter’s sensor cell. Although the initiating events and 

Rosemount’s initial Part 21 violation occurred a considerable time ago, they 

represent a significant weakness in the implementation of Rosemount’s 10 CFR 

Part 21 and 10 CFR Part 50, Appendix B programs. 

The NRC staff determined that in the mid-1980s, Rosemount provided incomplete 

and inaccurate information to NRC licensed facilities regarding the scope of 

the Rosemount transmitter oil-loss problem and that some Rosemount staff were 

not candid about providing comprehensive information regarding the oil-loss 

matter. Several licensees subsequently used this information, in part, for 

their evaluation of the deviation pursuant to 10 CFR Part 21. 

and testimony that were reviewed by NRC staff indicated that Rosemount limited 

the oil-loss information it disclosed and the information disclosed was 

typically site specific. Further, Rosemount did not describe the sensor cell 

manufacturing and test weaknesses which could have a1 1 owed transmitters that 

exhibited slow leaks to be shipped to NRC licensees. These weaknesses in the 

manufacturing and test processes preceding the 1 ate 1980s were potentially 

generic, could have caused degraded transmittzi- operation in any of 

Our deliberations took into 

Based on 

The documents 
. .. .. 

Mr. Steve M. Quist 2 

Rosemount's 1150-series transmitters, and should have been sent to NRC 

1 icensees for evaluating their specific transmitter safety-related 

applications pursuant to 10 CFR Part 21. 

Based on 01's investigation and the NRC staff's inspection, the NRC has 

concluded that Rosemount acted in careless disregard of 10 CFR Part 21 

requirements and its own procedures by failing to adequately evaluate or to 

inform its customers of the potential for degraded transmitter operation as a 

result of sensor cell oil-loss. We have determined that Rosemount's actions 

in this period between 1984 and 1988 - the mult.iple opportunities to have 

recognized the generic implications of numerous 1150 series transmitter 

problems, the repeated failure to recognize these problems by experienced 

Rosemount personnel, and the reluctance of Rosemount personnel to allow candid 

communications in order to adequately inform customers of the deviations - reflect careless disregard for the requirements of 10 CFR Part 21. Therefore, 

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

Enforcement Actions" (Enforcement Policy), 10 CFR Part 2, Apendix C, the 

failure to perform an adequate Part 21 review and to inform your customers of 

the potential for degraded operation of Rosemount's 1150-series transmitters 

in safety-related applications is being classified as a Severity Level I1 

violation. Your past performance in this matter was unacceptable. A civil 

penalty has not been proposed inasmuch as we were not able to identify any 

Rosemount Director or Responsible Officer subject to 10 CFR Part 21 who 

knowingly and consciously failed to provide the notice required by 10 CFR 

21.21. In addition, the general standard for imposition of a civil penalty 

against a licensee under Section 234 of the Atomic Energy Act of 1954, as 

amended, which is less specific and stringent than the "knowingly" and 

"consciously" standard for imposition of a civil penalty against a vendor 

pursuant to 10 CFR Part 21, does not apply in these circumstances. 

The failure of the Rosemount staff to candidly inform NRC licensees of the 

potential oil-loss failure mode in the four years that Rosemount was aware of 

the problem constitutes a very significant regulatory concern and indicates 

significant weaknesses in your implementation and compliance with 10 CFR 

Part 21. The NRC staff does acknowledge that after the Part 21 

responsibilities were discussed with Rosemount during the 1993 inspection 

effort, Rosemount undertook corrective action to preclude recurrence. For 

example, Rosemount has revised, and provided NRC staff with, its 10 CFR 

Part 21 procedure DP-N-1626, "Implementation of 10 CFR Part 21 for Deviations 

and Failures to Comply in Nuclear Products." The NRC staff also notes that 

the current Rosemount staff interactions with NRC seem more open and candid 

then in the past. 

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

specified in the enclosed Notice when preparing your response. 

response, you should document the specific actions taken and any additional 

actions you plan to take in order to prevent recurrence. 

response to this Notice, including your proposed corrective actions and the 

results of future inspections, the NRC will determine whether further NRC 

In your 

After reviewing your 

166 
Mr. Steve M. Quist 3 

enforcement action i s necessary to ensure compl i ance w'i th NRC r e g u l a t o r y 

requirements. The NRC staff will review the adequacy of Rosemount's 

corrective actions during a future inspection. 

Further, as a result of the inspection and investigation activities conducted 

by the NRC, the staff concluded that Rosemount provided inaccurate and 

incomplete information to the NRC during an April 13, 1989, public meeting 

regarding the failure experience of the Rosemount Model 1152 transmitter. 

However, the NRC staff found that Rosemount did not deliberately supply the 

NRC staff with inaccurate and incomplete information at this meeting. The NRC 

has not proposed any enforcement action for the .inaccurate and incorrect 

statements made to the NRC and its licensees because the applicable regulation 

(10 CFR 50.9) applies to applicants and licensees and not an entity like 

Rosemount under the circumstances of this case. However, the staff has 

substantial concerns about this matter and the staff must emphasize that 

Rosemount's actions resulting in the submittal of inaccurate or incomplete 

information to the NRC and licensees are wholly unacceptable. The NRC expects 

all licensee and vendor communications to be complete and accurate, and to 

properly reflect situations that could have implications to public health and 

safety, especially where an evaluation pursuant to 10 CFR Part 21 is involved. 

Therefore, in addition to your response to the Notice, ,you are requested to 

provide, under oath or affirmation, a written description of those actions 

that Rosemount has taken or intends to take to ensure that information 

provided to the NRC or its licensees is complete, candid, and accurate in all 

material respects. The response to this request should be provided within 

30 days of the date of this letter. 

Additionally, it is noted that the NRC staff received a Rosemount letter dated 

September 28, 1994. In this unsolicited letter, Rosemount stated that it 

agreed with the NRC "views" expressed at the June 23, 1994 enforcement 

conference on the importance of 10 CFR Part 21 and was "pleased with the 

conclusion.. . that no deliberate violation occurred." Flosemount also stated, 

however, that it cannot concur in the view that Rosemount acted in careless 

disregard by failing to adequately identify and report potential defects in 

its Model 1153 pressure transmitters prior to December 1988. The NRC staff 

noted that Rosemount attached a 40 page enclosure to its letter that takes 

exception to a number of statements and conclusions delineated in NRC 

Inspection Report 99900271/93-01 which provided some of the bases for the 

enforcement action that is set forth in the attached Notice of Violation. The 

NRC staff has reviewed the information in the Rosemount letter and has 

determined that the letter does not provide new information or arguments that 

would cause any change in the staff's position or enforcement action. 

167 
Mr. Steve M. Quist 4 

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. 

Si ncerel y , 

Docket No. 99900271 

Enclosure: 

Notice of Violation 

cc: See next page 

168 
Mr. Steve M. Quist 5 

cc: Mr. Kenneth E. Ewald, Business Unit Manager 

Rosemount Nuclear Instruments , Incorporated 

12001 Technology Drive 

Eden Prairie, Minnesota 55344 

Mr. Paul Blanch 

135 Hyde Road 

West Hartford, Connecticut 06117 

Mr. Ernest Hadley 

414 Main Street 

Post Office Box 3121 

Wareham, Massachusetts 02571 

169 
NOTICE OF VIOLATION 

Rosemount Nuclear Instruments, 

Incorporated 

Chanhassen, Minnesota 55317 

Docket No. 99900271 

EA 94-049 

During an NRC inspection conducted between February 1 and 4, and March 8 and 

12, 1993, and an NRC investigation that was conducted between February 1990 

and November 1993, the staff identified a violation of NRC requirements. In 

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

Enforcement Actions," 10 CFR Part 2, Appendix C (1993), the violation is 

1 i sted bel ow: 

A. Section 21.21, "Notification of failure to comply or existence of a 

defect," of the version of 10 CFR Part 21 in effect from January 1978 

until July 1991, stated, in part: 

Each individual, corporation, partnership, or other entity 

subject to the regulations in this part shall adopt 

appropriate procedures to (1) provide for (i) evaluating 

deviations or (ii) informing the licensee or purchaser of 

the deviation in order that the licensee or purchaser may 

cause the deviation to be evaluated unless the deviation 

has been corrected. 

Rosemount Nuclear Instruments, Inc. is subject to 10 CFR Part 21. 

Contrary to the above, the Measurement Division of Rosemount, 

Incorporated (Rosemount) , Quality Implementation Procedure (QIP) 126(N), 

"Potential Defect or Deviation in Products for Nuclear Application," 

issued March 18, 1981, was not adequately established and implemented to 

ensure that deviations, as defined in 10 CFR 21.3(e), about which 

Rosemount did not have adequate knowledge of each of the applications at 

NRC licensee facilities, were either evaluated or transmitted to the 

appl i cab1 e customer or 1 i censee for eval uati on. Speci f i cal ly, between 

1984 and December 1988, Rosemount failed to properly inform licensees of 

a potential for a sensor cell oil-loss problem that could occur in its 

nuclear-safety-re1 ated 1150-series pressure transmi tters which could 

have caused safety limits to be exceeded or-caused substantial safety 

hazards to exist at licensee facilities. 

This is a Severity Level I1 violation (Supplement VII). 

170 
Notice of Vi ol at i on 2 

Pursuant to the provisions of 10 CFR 2.201, Rosemount Nuclear Instruments, 

Incorporated is hereby required to submit a written statement or explanation 

to the U.S. Nuclear Regulatory Commission, ATTN: 

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

Branch, Division of Reactor Inspection and Licensee Performance, Office of 

Nuclear Reactor Regulaticrl, within 30 days of the date of the letter 

transmitting this Notice of Violation (Notice). 

marked as a "Reply to a Notice of Violation" and should include: 

reason for the violation, or, if contested, the basis for disputing the 

violation, (2) the corrective steps that have been taken and the results 

achieved, (3) the corrective steps that will be taken to avoid further 

violations, and (4) the date when full compliance will be achieved. Where 

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

time. Under the authority of Section 182 of the Atomic Energy Act of 1954, as 

amended, 42 U.S.C. 2232, this response shall be submitted under oath or 

aff i rmat i on, 

Document Control Desk, 

This reply should be clearly 

(1) the 

Dated at Rockville, Maryland 

this /56 day of November 1994. 
NRC FORM 335 

- 

12491 

NRCM 1102. 

US. NUCLEAR REGULATORY COMMISSION 

3201,3202 BIBLIOGRAPHIC DATA SHEET 

2. TITLE AND SUBTITLE 

(See instructions on the reverse) - 

Licensee Contractor and Vendor Inspection 

Status Report 

Quarterly Report 

October - December 1994 

5. AUTHOR(S) 

- 

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC. provide Division. OfficeorReghn, U.S. Nuclear Rrguhtory 

- 

Commission, andmai1ingaddres;if contractor,pmvide 

name and mailing addreal 

Division of Technical Support 

Office of Nuclear Reactor Regulation 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC. ~~e'Ssmeasabove'~ilcontractor.provideNRCDivision, Officeor Region, U.S NudearRwulatory Commirrion, 

and mailhg addresd 

Same as above 

10. SUPPLEMENTARY NOTES 

11. ABSTRACT 12W words or /err1 

This periodical covers the results of inspections performed by the NRC's Special 

Inspection Branch, Vendor Inspection Section, that have been distributed to the 

inspected organizations during the period from October through December 1994. 

12. KEY WORDS/DESCR:PTORS (List wodsorphrares that willassist mearchers in 1ocatingthereporc.l 13. AVAILABILITY STATEMENT 

Unl imi ted 

Vendor Inspection 14. SECURITY CLASSIFICATION 

(This Page) 

Unclassified 

Uncl ass i f i ed 

(This Report) 

15. NUMBER OF PAGES 

16. PRICE 

NRC FORM 335 (2891 

1. REPORT NUMBER 

(Anlgned bv NRC Add Vol., Supp., Rev., Md Addendum Numben, If MY.) 

NUREG-0040 

Vol. 18, No. 4 

3. DATE REPORT PUBLISHED 

I YEAR MONTH 

February ' 1995 

I. FIN OR GRANT NUMBER 

5. TYPE OF REPORT 

Quarterly 

October - December 1994 

1. PERIOD COVERED llncluriveDares1