uarterly Report 

ctober - December 1995 

S. Nuclear Regulatory Commission 

NUREG-0040 

Vol. 19, No. 4 
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dOCUm€?Ilt 
NUREG-004c 

Vol. 19, No. 4 

Licensee Contractor 

and Vendor Inspection 

Status Report 

Quarterly Report 

October - December 1995 

Manuscript Completed: February 1996 

Date Published: February 1996 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

ABSTRACT 

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

Special Inspection Branch, Vendor Inspection Section, that have been 

distributed to the inspected organizations during the period from October 1995 

through December 1995. 

CONTENTS 

- PAGE 

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

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

Inspection Reports ....................................................... I 

Crane Val ves Nuclear Operations 

Joliet, IL 

(99901285/95-01) ............... 2 

GE Nuclear Energy 

King of Prussia, Pennsylvania 

Paci f ic Scientific 

Yorba Linda, California 

(99901001/95-01) ............... 9 

(99900255/95-01) ............... 31 

Westi nghouse Electric Corporati on (99900404/95-02) ............... 42 

Pittsburgh, Pennsylvania 

Select Generic Correspondence on the Adequacy of Vendor .................. Audits and the Quality of Vendor Products 

54 

Correspondence Re1 ated to Vendor Issues .................................. 55 

V 

INTRODUCTION 

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

licensing and inspection program is that licensees are responsible for the 

proper construction and safe and efficient operation of their nuclear power 

plants. 

for the inspection of commercial nuclear facilities to provide for multiple 

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

participates in a quality verification process in compliance with requirements 

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

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

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

their contractors, while the major inspection effort is performed by the 

industry within the framework of quality verification programs. 

The Federal government and nuclear industry have established a system 

The licensee is responsible for developing and maintaining a detailed quality 

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

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

is responsible for ensuring that suppl iers, contractors and vendors a1 so have 

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

codes, and standards. 

The Vendqr 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 qual ification tests, and holders of NRC 

construction permits and operating 1 icenses in vendor-re1 ated areas. 

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

probl ems are determined and appropri ate corrective actions are devel oped. The 

inspections also review vendors to verify conformance with applicable NRC and 

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

coordination between 1 i censees and vendors. 

These 

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

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

and review of equipment problems found during operation and their corrective 

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

the inspected organization is required to take appropriate corrective action 

and to institute preventive measures to preclude recurrence. 

implications are found, NRC ensures that affected licensees are informed 

through vendor reporting or by NRC generic correspondence such as i nformat i on 

notices and bulletins. 

When generic 
This quarterly report contains copies of all vendor inspection reports issued 

during the calendar quarter for which it is published. 

report lists the nuclear facilities inspected. This information will also 

alert affected regional offices to any significant problem areas that may 

require special attention. Appendices list selected bulletins, generic 

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

correspondence i nvol vi ng vendor i ssues . 

Each vendor inspection 
INSPECTION REPORTS 

1 
November 6, 1995 

Mr . Henry G. McCul1 ough , Manager , Qual i ty 

Crane Valves Nucl ear Operations 

104 North Chicago Street 

Joliet, IL 60431 

SUBJECT : NRC INSPECTION NO. 99901285/95-01 

Dear Mr . McCul1 ough : 

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

(NRC) inspection of Crane Valves Nucl ear Operations (CVNO) , Jol i et and 

Romeoville, Illinois, conducted September 25-28, 1995. The NRC inspection 

team, led by David H. Brewer and other inspectors named in the report, 

conducted a performance-based evaluation of CVNO management , staff , and 

quality programs and the implementation of those programs related to the 

fabrication of valves and valve components. 

provide a basis for confidence that CVNO products supplied to the U.S. nuclear 

industry in fuel assemblies would perform their safety function. 

The inspection was conducted to 

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

representative records, (b) held discussions, and (c) made various 

observations. On the basis of this inspection, the NRC team determined that 

the implementation of the CVNO quality assurance program, documented in the 

CVNO Nuclear Assurance Manual, Edition 2, Revision 4, March 5, 1994, either 

met or exceeded the requirements of Appendix B to Part 50 of Title 10 of the 

Code of Federal Reaulations (Appendix B to 10 CFR Part 50). The enclosed 

inspection report contains a detailed discussion of the areas examined and the 

NRC team observations. 

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

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

Should you have any questions concerning this report, we will be pleased to 

discuss them with you. Thank you for your cooperation during this process. 

Sincerely , 

Original signed by Don Norkin for: 

Robert M. Gallo, Chief 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Docket No.: 99901285 

Enclosure: Report No. 99901285/95-01 
REPORT NO.: 

ORGANIZATION: 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACT I V I TY : 

INSPECTION DATES: 

LEAD INSPECTOR: 

OTHER INSPECTORS: 

REVIEWED BY : 

APPROVED BY: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF TECHNICAL SUPPORT 

99901285/95-01 

Crane Valves Nuclear Operations (CVNO) 

104 North Chicago Street 

Joliet, Illinois 60431 

Henry G. McCullough 

Manager, Engineering 

Crane Valves Nuclear Operations manufactures, repairs, 

modifies or replaces ASME Section 111, Division 1, 

Class 1, 2 and 3 (includes non-code nuclear safetyre1 ated, government products and nonsafety-re1 ated 

manufacture) valves, valve parts and appurtenances. 

September 25 through 28, 1995 

Vendor Inspection Section (VIS) 

Special Inspection Branch (PSIB) 

3 

Richard P. McIntyre, VIS/PSIB 

a1 ina, Chief, VIS/PSIB 

{hkiDate 

Robert M. Gallo, Chief, PSIB Date 

Enclosure 
1 SCOPE AND SUMMARY OF INSPECTION FINDINGS 

During this inspection, the NRC inspection team (team) evaluated CVNO 

management, staff, and quality programs and the implementation of those 

programs related to the manufacture, repair, modification and replacement of 

nuclear safety-re1 ated valves and valve parts. The i nspect i on was conducted 

to provide a basis for confidence that these items and services supplied to 

the U.S. nuclear industry would perform their safety function. 

basis consisted of the following: 

The inspection 

CVNO Nuclear Quality Assurance Manual (QAM), Edition 2, Revision 4, 

March 5, 1994 

Appendix B, "Qual i ty Assurance Criteria for Nuclear Power P1 ants and 

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

Federal Requlations (10 CFR Part 50) 

Part 21, "Reporting of Defects and Noncompl iance, 'I of 10 CFR 

1.1 Yiolations 

No violations were identified during this inspection. 

1.2 Nonconformances 

No nonconformances were 

2 STATUS OF PREVIOUS 

This was the first NRC 

3 INSPECTION FINDINGS 

identified during this inspection. 

NSPECTION FINDING 

nspect i on of CVNO. 

AND OTHER COMMENTS 

3.1 Backwound 

Crane Company began manufacturing valves in 1855. 

their first nuclear valves and supplied them to the U.S. Navy. 

valves supplied to a commercial nuclear power plant went to Duquesne Power and 

Light, Shipppingport, PA, in 1956. 

In 1952 Crane manufactured 

The first 

In 1959 Crane made the first of a series of acquisitions by obtaining the 

Chapman Valve line (tilting disc check valve). 

a leading stainless steel valve manufacturer, and nuclear valve operations 

became known as Crane-Aloyco, a division of the Crane Company. Also in 1985 

Crane obtained the 1 icense for original equipment manufacturer production of 

Walworth nuclear spare parts and valves and acquired the Mark Controls service 

centers (Romeoville, Houston, Gonzales, San Leandro, Signal Hill and 

Woodbury). 

Flowseal and Center1 ine products to its manufacturing and supply capability. 

In 1985 Crane acquired Aloyco, 

In 1994 Crane acquired Mark Controls adding Pacific Valves, 

-2- 
In 1988 the Chapman facility, Indian Orchard, MA, was closed and 

safety-related products were transferred to Crane's Romeoville facility. 

1992 the Flowmatics control valve line was transferred from the Chempump 

division to Crane's Nuclear Operations facility in Romeoville. 

In 

In 1971 Crane became the first manufacturer in the United States authorized to 

use the Nuclear "N" symbol of the American Society of Mechanical Engineers, 

for both valves and weld fittings. 

Crane-Aloyco became Crane Valves Nuclear Cperations in 1994. 

3.2 Entrance and Exit Meetinas 

During the entrance meeting in Joliet, IL, on September 25, 1995, the team met 

with members of CVNO management and staff, discussed the scope of the 

inspection, reviewed mutual responsibilities for handling proprietary 

information and established contact persons for the team within the management 

and staff of the CVNO organization. 

During the inspection, the NRC team conducted a performance-based tnspection 

of CVNO organizations through technically directed observations and 

evaluations of processes, activities, and documentation. The NRC team 

(a) examined technical documentation, procedures, and representative records, 

(b) conducted interviews, (c) held discussions, and (d) made various 

observations. 

Inspection participants and contacts are listed in Section 4. 

During the exit meeting on September 28, 1995, the team summarized the 

inspection results with CVNO management and staff. 

3.3 Oualitv As surance Manual Control 

The team reviewed QAM, Section I I-A, "Quality Assurance Program," Paragraph 

4.0, "Manual Control ," to determine the appropriate QA program requirements. 

The team reviewed implementation of the QAM revision process for the changes 

from Revision 2 to Revision 3. Revision 3 was made to incorporate changes 

requested by an ASME Survey conducted August 2-4, 1993. 

review, numerous suggestions and comments were made which were incorporated 

into Revision 3 of the QAM. The team verified that QAM Revision 3, August 4, 

1993, had been reviewed and approved by the QA manager and by the ASME 

Authorized Nuclear Inspector (ANI). 

and 4 had been distributed to all listed holders of a controlled copy of the 

QM. 

During the ASME 

The team also determined that Revisions 3 

3.4 Indoctrination and Training 

The team reviewed QAM, Section II-A, "Quality Assurance Program," Paragraph 

6.0, "Indoctrination and Training," and Procedure Number 02-101, 

"Indoctrination and Training," Revision 4, May 18, 1990, to determine the 

appropriate QA program requirements. 

and three design support engineers for a review of their training records. 

The team chose three quality assurance 

-3- 
For each position, one individual was relatively new to the position and ope 

was the senior engineer in the group. 

indoctrination and training records that each of the six engineers had 

attended indoctrination, formal training classes or had reviewed and signed 

for all the training required for their positions as listed on their personal 

Indoctrination and Training Matrix and as described and required in Procedure 

Nt'imber 02- 101. 

The team verified through review of 

When reviewing and verifying the training records of the engineers for 

Engineering Bulletin, EB-013, "Preparation, Review and Approval of Design 

Calculations," Revision 0, November 24, 1994, the team questioned the 

circumstances under which an Engineering Bulletin was issued. 

stated that Engineering Bulletins were typically issued to document and 

discuss technical issues that the engineers use in performing their jobs. 

Procedure Number 03-104 stated that an "Engineering Bulletin is a document 

issued by Crane-Aloyco's Engineering Department for the purpose of informing 

operating personnel and/or customers of the design, material, methods or 

practices uti1 ized by Crane-Aloyco." 

The team noted that EB-013 contained required processing information for the 

preparation, review and approval of calculations which are integral to, or 

which support engineering, design and analysis. Based on the definition of an 

Engineering Bulletin contained in Procedure Number 03-104, "Preparation and 

Issue of Engineering Bulletins," Revision 1, May 1, 1992, the team suggested 

that EB-013 would be more appropriately classified as a Procedure because it 

implemented the QAM. CVNO personnel agreed to make this change. 

3.5 gesian Control 

CVNO personnel 

CVNO manufactured gate valves, globe valves, check valves and butterfly valves 

and supplied replacement parts. They repaired, but did not manufacture, power 

operated re1 ief Val ves and safety re1 ief valves. 

The team reviewed QAM, Section 111, "Design Control ,'I Paragraph 4.0, "Design 

Basis and Design Inputs," which stated that, "The Owner's [customer's] 

Certified Design Specification and the requirements of the [ASME] Code shall 

be the basis for CVNO's design review for complete valve assemblies. 

valves 4 in. nominal size and less, CVNO does not provide its own Design 

Specification as allowed by the [ASME] Code, but uses the Owner's Design 

Specification as basis for construction." 

The team reviewed Procedure Number 03-103, "Control of Design Interface," 

Revision 2, July 14, 1992. This document provided design guidance for 

internal design interfaces as we1 1 as custcmer interfaces. 

For 

The team reviewed Engineering Bulletin EB-004, "Valve Parts Classification," 

Revision 0, April 10, 1989, which contained CVNO baseline design for all 

valves and valve parts by specifying the classification of each component. 

EB-004 stated that CVNO uses ASME Boiler and Pressure Vessel Code Case N-62-4 

as the primary basis for classifying components of "Code Nuclear" and "NonCode Nuclear Safety-Re1 ated valves. I' EB-004 pl aced a1 1 valve components into 

-4- 
one of three classifications; "N", Nuclear, ASME Code Section 111, Class 1, 2 

& 3; "Q", Non-code, Nuclear Safety Related; and 'IC", Non-code, Non-SafetyRe1 ated Nuclear Power P1 ant Use. 

The QA manager stated that when a purchase order was received from a customer, 

CVNO personnel reviewed the classification of the components with the customer 

to ensure that the customer was fully aware of the classification. 

customer required one or more components to be upgraded from "C" to I'Q", the 

purchase order was revised to reflect the change. If special testing (e.g., 

seismic), was required, it was stated in the purchase order and the test 

report was included in the documentation furnished with the valve or 

components on delivery. Certified material test reports (CMTRs) provided with 

the product identified components individually and stated their classification 

individually. If a component was "N" or "Q", the CMTR included the heat 

number or lot number to provide traceability. 

number. 

If the 

'IC" items had no heat or lot 

The team reviewed four purchase orders from nuclear utilities to determine 

CVNO adherence to EB-004 classification and determined that in each case there 

was compliance with the requirements of tC-004. 

The team reviewed Engineering Bulletin No. EB-005, "Stem Material 

Standardization," Revision 0, June 29, 1989, which recmmended the use of ASTM 

A-564, Type 630, Condition H1150 (17-4PH), in the place of 410 and 304 

stainless steels for stem and hinge pin applications. EB-005 also instituted 

17-4PH, H1150, as the CVNO standard for such applications. The team 

considered the recommendation of 17-4PH usage relative to NRC Information 

Notice 92-60, "Valve Stem Failures Caused by Embrittlement," August 20, 1992. 

The Information Notice alerted addressees to the fact that 17-4PH, H900 

through H1150, could become excessively embri ttled on exposure to temperatures 

of 600°F and higher for a period of several thousand hours or longer as the 

result of secondary aging. 

valves mounted on the pressurizer of a pressurized water reactor are the 

valves subjected to this type of service. The team determined that CVNO did 

not produce this type of valve and although they provided replacement parts 

for such valves, the replacement parts were supplied to the design 

specificatiors of the owner of the valves as they were less than 4 inches in 

nominal diameter. 

materials was consistent with internal procedures and regulatory guidance. 

Power operated relief valves and safety relief 

The team determined that CVNO application of 17-4PH 

The team determined that CVNO had established measures for the selection and 

review for suitability of application of materials, parts and processes 

essential to the safety-related functions of the components they provide to 

the nuclear power industry. The team also determined that CVNO had 

established measures that identified and controlled the various design 

interfaces. 

-5- 
3.6 JO CFR Part 21 Prowam 

The team observed the posting of required documents in prominent places in the 

CVNO Joliet, IL, office and in the Romeoville, IL, facilities at 12 East 

Devonwood and 720 North Parkwood. 

The team reviewed Procedure Number 15-100, "Reporting of Defects and Nonconformances (IOCFREI)," Revision 2, October 29, 1993, and determined that it 

complied with the requirements of 10 CFR Part 21. 

Although Procedure Number 15-100 was first issued January 10, 1986, no issues 

were entered into the Part 21 log until January 1994. The team made the 

observation that it was unusual that no items were entered into the log during 

the eight year period between January 1986 and January 1994. CVNO personnel 

made the point that during that period they were almost exclusively involved 

with producing replacement parts for existing va'lves on a small lot basis and 

thereby not likely to encounter conditions that would require evaluation or 

reporting. The team told CVNO personnel that it was as important to document 

issues that were raised, evaluated and not reported to the NRC as it was to 

document those issues that were reported to the NRC. 

4 PERSONS CONTACTED 

The NRC staff participating in the inspection; CVNO personnel contacted during 

the inspection; and the personnel attending the entrance and exit meetings are 

listed below. A bullet (0) indicates that person attended the entrance 

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

Crane Valves Nuclear Operations: 

0 t McCullough, H.G. Manager, Engineering 0 t Bisesto, F.J. Director, Customer Service 0 t Carl son, J . Director, Operations e t Landhol t, W .W. President, Valve Group 

U.S. Nuclear Regulatory Cmission: 

0 t Brewer, D.H. Metal 1 urgi cal Engineer, 

0 t McIntyre, R.P. Senior Reactor Engineer, 

VIS/PSIB/DISP/NRR 

VIS/PSIB/DISP/NRR 

t Cwal ina, G.C. Chief, VIS/PSIB/DISP/NRR 

-6- 

8 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 205Sj-OWl 

December 4, 1995 

Mr. Joe R. Harrell , Manager 

GE Nuclear Energy 

Nuclear Field Services/Power Del ivery Services 

640 Freedom Business Center 

King of Prussia, PA 19406 

SUBJECT: NRC INSPECTION REPORT NO. 99901001/95-01 

Dear Mr. Harrell: 

This letter addresses the inspection of your organizat ,on's activities at the 

Philadelphia Service Shop of the GE Apparatus Service Division and at the 

Specialty Breaker Plant in Philadelphia, conducted September 11-14, 1995, by 

Mr. Stephen Alexander and IYr. Ronald Frahm, Jr., of this office, and the 

discussion of their findings with the members of your staff identified in the 

enclosed report at the conclusion of the inspection and in subsequent 

telephone conversations. The inspection was conducted to assess corrective 

actions implemented by your organization in response to quality assurance 

' audit by customers and GE Nuclear Energy, San Jose, problems with circuit 

breaker maintenance, and problems with dedication of commercial grade items 

for use in safety-related applications. The inspectors also review d GE 

the Code 

inspection 

thin these 

Nuclear Energy procedures adopted pursuant to Part 21 of Title 10 o 

of Federal Reaulations (10 CFR Part 21). Areas examined during the 

and our findings are discussed in detail in the enclosed report. W 

areas, the inspection consisted of an examination of procedures and 

representative records, interviews with personnel, and observations 

inspectors. 

by the 

During this inspection, we determined that the implementation of your quality 

assurance (QA) program did not meet certain NRC requirements, specifically, 

Criterion XIII, "Handling, Storage, and Shipping," of Appendix B to 10 CFR 

Part 50. We observed unsafe handling practices in moving Magne-Blast circuit 

breakers in the service shop using overhead travelling hoists and found that 

adequate procedures to control this activity had not been established. 

specific findings and references to the pertinent requirements are identified 

in the enclosures to this letter. You are requested to provide us within 30 

days from the date of this letter a written statement in accordance with the 

instructions specified in the enclosed Notice of Nonconformance. 

The 

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

this letter, its enclosures, and your response will be placed in the NRC 

Public Document Room (PDR). To the extent possible, your response should not 

include any personal privacy, proprietary, or safeguards information so that 

it can be placed in the PDR without redaction. 

necessary to include such information, yo6 should clearly indicate the 

specific information that you desire not to be placed in the PDR, and provide 

the legal basis to support your request for withholding the information from 

the public. 

Yowever, if you find it 

9 
Mr. J. R. Harrell -2- 

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

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

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

The cooperation of your staff and that of the GE Philadelphia Service Shop, 

and GE Specialty Breaker Plant in this matter was greatly appreciated. 

you have any questions about the enclosed report, we would be glad to discuss 

them with you. 

Should 

4 Si ncerel Speci obert a1 M. Inspection Gallo, Chief Branch 

Divisim of Inspection and Support Programs 

Office of Nucl ear Reactor Regul at i on 

Docket No.: 99901001 

Enclosures: 1. Notice of Nonconformance 

2. Inspection Report 99901001/95-01 

cc w/encl : 

Mr. Forrest Hatch, Manager NS&PO Quality, GE NE, San Jose 

Mr. Normand ROUX, Manager, Switchgear Services, GE Philadelphia Service Shop 

Mr. Edward Dugan, Chief Design Engineer, GE Specialty Breaker Plant 

10 
NOTICE OF NONCONFORMANCE 

GE Nucl ear Energy, Power Del i very Services 

King of Prussia, Pennsylvania 

Docket No. 99901001 

Report No. 95-01 

Based on the results of an NRC inspection conducted on September 11-14, 1995, 

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

with NRC requirements. 

A. Criterion XIII, "Handling, Storage, and Shipping," of Appendix B to 10 

CFR Part 50, states, in part: "Measures shall be established to control 

the handl ing, storage, shipping, cleaning and preservation of material 

and equipment in accordance with work and inspection instructions to 

prevent damage or deterioration. 

Contrary to the above, PDS did not establish adequate procedures to 

control handl ing of Magne-Blast circuit breakers being moved about the 

service shop in a manner that would ensure that they would not be 

damaged. As a result, unsafe handling practices were used when moving 

Magne-Blast circuit breakers (without arc chutes) in the service shop 

using overhead travelling hoists. 

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

Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555 

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

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

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

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

include for each nonconformance (1) the reason for the nonconformance, or if 

contested, the basis for disputing the nonconformance, (2) the corrective 

steps that have been taken and the results achieved, (3) the corrective steps 

that will be taken to avoid further noncompliances, and (4) the date when your 

corrective action will be completed. Where good cause is shown, consideration 

will be given to extending the response time. 

Dated at Rockville, Maryland 

this 4th day of December, 1995 

Enclosure 1 
ORGAN I ZAT I ON 

INSPECTED: 

INSPECTION 

CONDUCTED AT: 

REPORT NO.: 

PRINCIPAL 

CONTACT : 

CORRESPONDENCE 

ADDRESS : 

NUCLEAR INDUSTRY 

ACTIVITY: 

INSPECT ION TEAM 

LEADER: 

OTHER INSPECTORS: 

REVIEWED: 

APPROVED: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF INSPECTION AND SUPPORT PROGRAMS 

VENDOR INSPECTION REPORT 

GE Nuclear Energy (GE NE) 

Nuclear Field Services/Power Delivery Services (PDS) 

640 Freedom Business Center 

King of Prussia, PA 19406 

(NRC Docket No. 99901001) 

GE Apparatus Service Division 

Phi 1 adel phi a Service Shop 

1040 East Erie Avenue 

Philadelphia, PA 19124 

(NRC Docket No. 99901147) 

GE Speci a1 ty Breaker P1 ant 

6901 Elmwood Avenue 

Philadelphia, PA 19142 

(NRC Docket No. 99900219) 

99901001/95-01 

George Sanders, Lead Engineer, GE NE PDS 

610-992-6049 

640 Freedom Business Plaza 

King of Prussia, PA 19406 

Medium-vol tage swi tchgear manufacturing (SBP) , equipment maintenance and repair services (ASD) 

for electric power generation and distribution 

equ i pment 

Vendor Inspection Section (VIS) 

Special Inspection Branch (SIB) 

Ronald K. Frahm, Jr. 

Quality Assurance and Maintenance Branch 

-A?&. 

Gregory C. Cwalina, Chief VIS 

SIB 

Q)&-#L%& 

Robert M. Gallo, Chief, SIB 

/$I /9r 

Da e 

G+l45- 

Da e 

Enclosure 2 
1.0 

1.1 

e 

0 

0 

1.2 

1.3 

SUMRY OF INSPECTION FINDINGS 

Inspection Basis: 

The inspection basis comprised the following: 

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

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

Requlations (10 CFR Part 50) 

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

General Electric Nuclear Energy (GENE) Quality Assurance (QA) Program 

Documents and Procedures Appl icable to Power Del ivery Services (PDS) 

Scope of Activities at the GE Apparatus Service Division (ASD) 

Philadelphia Service Shop, and at the GE Specialty Breaker Plant (SBP) 

in Philadelphia. 

Vi 01 ati ons : 

None. 

Nonconformance: 

(99901001/95-01-01) Contrary to the requirements of Criterion XI11 of 10 CFR 

Part 50, Appendix B, GE NE PDS did not establish adequate procedures to 

control hand1 ing of Magne-Blast circuit breakers being moved about the service 

shop in a manner that would ensure that they would not suffer inadvertent 

damage or cause injury to shop personnel. As a result, unsafe handling 

practices were used when moving Magne-B1 ast circuit breakers (without arc 

chutes) in the service shop using overhead travelling hoists. 

3.5 of this report). 

(See Paragraph 

2.0 STATUS OF PREVIOUS INSPECTION FINDINGS 

No previous findings were addressed during this inspection. 

3.0 INSPECTION DETAILS 

3.1 InsDection Objectives: 

The approved inspection plan for this NRC Vendor Inspection Section 

inspection included the following objectives: 

3.1.1 Evaluate PDS QA at ASD with emphasis on corrective action and 

improvements prompted by a GE NE (San Jose) QA audit and customer-identified 

deficiencies such as incorrect trip units in serviced breakers for Boston 

Edison Company's Pilgrim Nuclear Station. 

3.1.2 

problems, failures to latch) identified at Maine Yankee and Millstone. 

Evaluate resolution of Magne-Blast circuit breaker problems (switch 

2 

13 
3.1.3 

any changes about which licensees may need +- he informed. 

3.1.4 

Bar Nuclear Plant and Shoreham Nuclear Plant. 

Review the joint PDS/SBP Magne-Blast design reconciliation to identify 

Review resolution of failures of Magne-Blast trip cranks at the Watts 

3.1.5 

items (CGIs) for use by SBP in building or ASD in servicing safetyre1 ated equipment . 

Evaluate GE NE PDS program at SBP for dedication of commercial grade 

3.2 flame-Blast Circuit Breaker Problems 

The NRC investigation of the circuit breaker problems at Maine Yankee, 

Millstone and Pilgrim discussed below revealed manufacturing deficiencies in 

switches manufactured by GE Electrical Distribution and Control (EDK), 

Plainville, Connecticut, deficiencies in workmanship and QA at ASD, 

Philadelphia, and deficiencies in the dedication process at the GE NE PDS 

commercial grade dedication facility at the GE SBP in Philadelphia. 

deficiencies resulted in audits of ASD by Yankee Nuclear Services Division for 

Maine Yankee and audit of ASD and the PDS dedication facility at SBP by GE NE, 

San Jose. In addition, the NRC visited Maine Yankee, ED&C, and ASD and GE NE 

to obtain further information and observe testing and inspection. 

visits resulted in issuing Information Notices 94-54, "Failures of General 

Electric Magne-Blast Circuit Breakers to Latch Closed," (August 1, 1994) and 

95-02, "Problems With General Electric CR2940 Contact Blocks in Medium-Voltage 

Circuit Breakers," (January 17, 1995). During this inspection, the NRC 

inspectors reviewed PDS corrective actions and programmatic improvements made 

as a result of the identified problems and audit findings. 

These 

These 

3.2.1 Inter1 ock Switch Probl ems 

On March 23, 1994, Maine Yankee Atomic Power company, the licensee for Maine 

Yankee, reported that during a pump surveillance test, 4.16-kV Magne-Blast 

breaker (Maine Yankee ID No. 3-17) for the high pressure safety injection 

(HPSI) pump motor (P-14B) failed to close. The affected breakers, GE Type 

AM-4.16-250-9HB, vertical-1 ift, medium-voltage (4.16-kV) , had been overhauled 

on site by a team from GE ASD Philadelphia in the summer of 1993. 

licensee investigated the incident and determined that one of the two circuit 

breaker cubicle interlock limit switches (comprising Interlock Switch Assembly 

52-1s) that had been replaced during the overhaul had not been properly 

manufactured and installed. The affected switch (which enables the breaker 

closing circuit when the vertical-lift breaker is fully elevated in its 

cubicle) was found loose on its mounting and misaligned. 

The 

The two normally-open (shelf state) configuration CR2940U310 contact blocks 

that comprise interlock switch assembly 52/IS are ganged together such that 

they are both held closed by a breaker cubicle interlock mechanism lever when 

the vertical-lift breaker is fully elevated in its cubicle. However, in this 

instance, the hole drilled in the head of one of the two special mounting 

screws in the upper contact block (fastening it to its mounting bracket) had 

no female threads tapped into it. Therefore, inappropriately using the 

existing fasteners, instead of control 1 ing the nonconforming screw and 

3 

14 
replacing it with a new, properly made screw, the new lower contact block 

could was mourlted onto the upper block using only one screw. 

subsequent operations of the breaker, the lower block became loose. 

Consequently, its plunger could not be depressed by the moving contact carrier 

of the upper contact block, thus it remained open, disabling the breaker 

closing circuit. 

During 

Following the HPSI pump breaker’s failure to close on March 23, 1994, the 

licensee discussed the material and workmanship concerns with GE PDS and 

initiated a comprehensive program to reinspect all the safety-related MagneBlast breakers. During subsequent breaker inspections, Maine Yankee and PDS 

discovered additional problems with mechanisms, including non-standard 

interlock paddles (spring steel in stead of rigid), and a damaged set screw in 

one crankshaft. The licensee identified some additional loose mounted power 

switches, interlock switches, and close-latch monitoring switches on several 

breakers. Additionally, other minor deficiencies, such as interference with 

the close-latch monitoring switch return spring and adjustment on the top 

positive interlock switch gaps, were also identified. 

these deficient conditions and the breakers were tested satisfactorily. 

The licensee corrected 

In addition, on September 26, 1993, while the plant was shut down in 

preparation for the loss of power testing, an emergency diesel generator (EDG1A) output circuit breaker (MY ID# 3-12) failed to close. The breaker closing 

spring was found uncharged. 

charging mechanism had failed to recharge its closing spring automatically 

when the breaker was last closed. 

52/IS used to enable the closing spring charging circuit this time (as opposed 

to the breaker remote closing circuit failure discussed above) were stuck 

open. The licensee corrected this problem by replacing the defective 

CR2940U310 contact block. Based on the deficiency identified in this EDG 

breaker, the licensee initiated a work order (WO No. 93-3293) to verify that 

all safety-re1 ated breakers’ charging circuitry and a1 1 breakers positive 

interlock switches were functional. The licensee also verified that the 

breaker functioned adequately overall and the connections in the breakers’ 

components were tight. 

sticking open) was believed to be similar to the failure of a power switch 

discussed below. No additional concerns were identif’zd during this 

verification. 

Troubleshooting revealed that the closing spring 

The contacts of interlock switch assembly 

The failure of the charging motor circuit switch (by 

The 1 icensee determined that during overhaul of the 4.16-kV circuit breakers 

on the site, ASD had installed several defective limit switches. 

to some contact sticking problems, several special fasteners in the contact 

blocks used for the Maine Yankee breaker overhaul were found with incomplete 

threading and/or shallow screwdriver slots. The manufacturer, GE-ED&C, 

determined that the problems occurred when a screw making machine shut itself 

down due to a broken tool. On this occasion, a number of incompletely 

machined screws that passed through during the shutdown sequence were not 

captured, but instead became mixed with that batch of finished fasteners. 

ED&C corrected the problem and confirmed that the use of the defective screws 

was limited to a few CR2940 contact blocks with JlA3XX= (1993) date codes. 

ED&C has reported establishing tighter controls to prevent recurrence, 

inspecting its stock, and weeding out any additional defective fasteners. 

In addition 

4 
3.2.2 Power Switch Fai 1 ures 

On December 6, 1993, while the plant was at fuli power. Emergency feedwater 

pump breaker (GE Serial no. 0224A4126-017, Maine Yankee ID No. 4-34) failed to 

close during a routine surveillance test. The breaker failed to close in 

about one-third of the attempts to close it during shop testing. PDS sent a 

technician to Maine Yankee to assist in the failure analysis. During the 

overhaul of the 4160-Vac Class 1E breakers at Maine Yankee, ASD had replaced 

the over 10-year-old original CR2940 contact blocks used for power switches 

(in addition to the CR2940 interlock switch replacement discussed above). 

Further evaluation on this breaker identified a normally closed set of 

contacts in power switch assembly 52/SM-LS in the closing circuitry would 

stick in the open position. Power switch assembly 52/SM-LS is made up of 

three CR2940 contact blocks (two U310, normally open, and one U301, normally 

c!osed) ganged together. 

by the closing spring charging mechanism. 

has one set of two contacts (numbered 1 through 6). The normally closed 

contacts 5 and 6 of power switch assembly 52/SM-LS are supposed to return to 

their normal closed state as the contact block's plunger is released (when the 

closing spring is fully charged) and enable the breaker closing circuit. 

contact blocks of power switch assembly 52/SM-LS open as soon as closing 

spring charging mechanism leaves the fully charged condition. 

subsequent breaker testing, the 1 icensee identified one other faulty power 

switch on a spare circuit breaker (4-32). 

The three contact blocks are operated simultaneously 

Each of the three contact blocks 

The 

During 

Internal examination of the failed switches revealed that the moving contacts 

(5 and 6) of the normally-closed contact block were sometimes stuck in the 

open position. This appeared to result from mechanical interference that 

prevented reclosure of the contacts by their internal return springs. 

observations raised the question of the reliability of at least this batch of 

new CR2940s (their date codes, e.g. , '1MA316=" indicated 1993 manufacture). 

The NRC pursued this issue with the CR2940 manufacturer, ED&C, the breaker 

manufacturer, SBP, and GE NE/PDS. ED&C eventually reported its determination 

confirming the inspectors' suspicions that the affected contact blocks came 

from a batch or batches that had some excess plastic injection mould flashing 

that was apparently interfering with the movement of the moving contact bar or 

its c:rrier assembly. GE ED&C has reported instituting initial corrective 

actions and preventive measures at its factory in Puerto Rico, but at the time 

of this inspection, had not yet provided definitive information on the scope 

of the problem; although there have been no further occurrences of this 

part i cul ar probl em reported. 

These 

According to the breaker manufacturer, SBP, power switch 52/SM-LS contacts 5-6 

are provided along with another accessory function called [trip] latch 

checking when this feature is supplied as an option. In some applications, 

such as at Palo Verde, contacts 5 and 6 provide a white light indication on 

the switchgear panel that the closing spring is charged and also that control 

power is available (hence the term "white 1 ight switch"). 

and 6 only perform a needed breaker control/interlock function when used in 

conjunction with a so-called "automatic reclosure" feature not used at Maine 

Yankee. 

Yankee Plant Engineering Department, done in consultation with PDS and SBP as 

However, contacts 5 

According to the technical evaluation (455-93) by staff of Maine 

5 

16 
part of a modification screening pursuant to 10 CFR 50.59, contacts 5 and 6 

are not required for a close-permissive function to inhibit closing attempts 

during closing spring charging, because the mechanism's design does not permit 

breaker closure before the closing spring is fully charged. Hence, contacts 5 

and 6 have no required function at Maine Yankee, yet they can disable the 

breakers' electrical closing circuit should they stick open. Therefore, in 

consultation with GE, Maine Yankee determined that under these circumstances, 

it should not impair any required function of the breaker to ''jumper them out" 

(short across or bypass them). According to the licensee's WO 94-42, the 

licensee eventually jumpered out the 52/SM-LS contacts 5 and 6 of 16 safetyre1 ated breakers. 

Other problems reported with the CR2940 contact blocks in Magne-Blast breakers 

have been contacts 3 and 4 of power switch 52/SM-LS in the charging motor 

circuit becoming welded shut when interrupting charging motor current. In 

addition, new contact blocks occasionally have been found inoperative during 

acceptance testing (sometimes as part of commercial grade dedication) , some 

due to misalignment or incorrect installation of internal parts. 

problems with the special screws such as shallow screwdriver slots also have 

been found as discussed above. At the request of GE NE/PDS, SBP is looking 

into using a different switch that is more reliable for these Magne-Blast 

applications, at least for nuclear safety-related service, and is planning to 

issue a Service Advice Letter (SAL) that will address these problems. 

Other 

3.2.3 Breaker Failures to Latch Closed 

During troubleshooting of the interlock switch problems at Maine Yankee, GE 

PDS and the licensee discovered a potential latching problem with the 4.16-kV 

circuit breaker mechanism. 

times during subsequent testing after its power switch failure episode 

discussed above. The affected 250-MVA capacity Magne B1 ast breaker models 

were AM-4.16-250-6, 7, 8 or 9-HB. The suffix "HI' denotes that the breakers 

were equipped with an "ML-13" operating mechanism, and the suffix "B" denotes 

that they were of the so-called "high momentary" design to achieve a 75-kA 

close-latch rating. This capability was added to 250-MVA breakers in 1975 

(350-MVA breakers were already high momentary) to meet current ANSI/IEEE 

Standard C37.20. 

breakers were fitted with heavier main and arcing contacts and heavier closing 

springs. In order to keep the breaker in proper balance, this required adding 

a second set of opening springs as well. However, the additional opening 

springs, along with some other improvements, made the breakers more sensitive 

to the various adjustments and variable parameters being at or near nominal 

design values in order to latch reliably. According to the GE SBP design 

engineer, a second prop spring was added in 1972 as a means of simplifying the 

process of set up and adjustment of the closely balanced, high-momentary 

breakers in the factory. 

sensitive to all adjustable parameters being close to nominal in order to 

achieve reliable latching. 

adjustments needed only to be within tolerance without the concern that random 

tolerance stackup might cause the breaker occasiona1;y to fail to latch 

closed. The occasional failure to latch could also occur in those 350-MVA 

capacity, AM-4.16 breakers (Model AM-4.16-350-1-H) all of which have two 

For example, breaker 4-34 fai 1 ed to 1 atch numerous 

In order to achieve the 75-kA closc latch rating, the 

The second prop spring made the action much less 

With the faster prop action, the various 

6 

17 

I 
opening springs, but equipped with only one prop spring. The problem 

apparently does not affect newer breakers that have two prop springs. .Over 

the years, during the overhwling of the breakers, GE retrofitted some 

breakers with two prop springs. 

After some problems with prop spring end hook breakage, SBP issued SAL 073- 

(SBP prefix) 348.1 in December 1990, which provided information for obtaining 

an improved (gold colored) main prop spring, less subject to breakage, and an 

improved design for the mounting bracket (chamfering or radiussing the hole 

edges) that would cause less wear on spring hooks. SAL 348.1 recomnended that 

all breakers’ be modified as described in the SAL during a scheduled outage 

with special attention being applied to breakers approaching 2000 operations. 

However, a3 though SAL 348.1 addressed breakers already fitted with the second, 

auxiliary prop spring, it did not mention the potential problems associated 

with only one prop spring under some conditions, nor did it recommend adding 

the second prop spring. The inspectors noted that SBP had issued SAL 351.1, 

dated June 23, 1994, which provided information and recommendations (including 

ordering information for the modification kit) for installation of the second 

prop spring in ML-13 and ML-13A Magne-Blast breaker mechanisms. In addition, 

in August 1995, SBP issued SAL 354.1 which filled a long-standing need for 

definitive lubrication guidance for Magne-Blast breakers. SAL 354.1 also 

included information on the failure to latch problems in high-momentary 

breakers with one prop spring engendered by replacing the TUF-LOC prop 

bushings with aluminum-bronze in conjunction with overhaul and relubrication 

which require installation of a second prop spring. 

spring installation, is referenced. 

SAL 351.1, on second prop 

Analysis of the failures at Maine Yankee (and some additional similar failures 

at Millstone) , determined that the problem began to occur intermittently 

following about 35-50 operations of affected breakers after the original “TUFLOC” prop bushings were replaced with the new aluminum-bronze bushings in 

breakers which have (1) two opening springs and one prop spring and (2) which 

already had the rest of the mechanism TUF-LOC sleeve bearings or bushings 

replaced with aluminum-bronze bushings in accordance with SAL 318.1. After 

all bushings, including the prop bushings, are replaced, and the breaker is 

cleaned and lubricated, and, as experience has shown, after the wearing-in 

peridd mentioned above, the timing and force balancer shift enough to make a 

high-momentary breaker with one prop spring highly sensitive to the 

combinations of tolerances in adjustments such that a second prop spring is 

mandatory to ensure reliable latching. Because of the observed delayed onset, 

the problem may not present itself until after the completion of post overhaul 

testing and receipt inspection and/or post installation/pre-operational 

testing. However, this testing can involve sufficient operations or cycles of 

the breaker to render it susceptible to the latch failure shortly after its 

re1 ease for pl ant operati on. 

18 

Under the conditions in question, breakers sometimes fail to latch closed 

because the prop does not move fast enough to be in the proper position under 

the prop pins of the mechanism linkage as they descend during the closing 

cycle (whether manually or electrically initiated). When the prop moves too 

slowly relative to the motion of the descending prop pins and the pins miss 

the prop - fall in front of it - instead of landing on it, the linkage 

7 
collapses and the main contacts fall open again. 

"going trip-free. I' 

This event is also called 

3.2.4 Tests and Inspections of Breaker 4-34 at 6E's Phi 1 adel phi a 

Service Shop 

On April 5, 1994, ASD technicians performed tests on the breaker 4-34 under 

the supervision of the Lead Electrical Engineer from the PDS. NRC inspectors 

and licensee staff observed these inspections and tests. During the testing, 

the breaker exhibited erratic, unreliable latching (only latched 6 times out 

of twenty attempts. When it did latch, it had inadequate prop wipe. During 

the course of the inspection and testing phase, ASD installed the second prop 

spring modification using a standard kit. After installing the second prop 

spring, the breaker latched 8 out of ten times; much better, but still not 

fully reliable. Upon further inspection, the GE NE PDS field engineer (based 

in King of Prussia, PA) noted that at least one of the opening springs had 

excessive preload on it. 

(deliberately not adjusted to nominal) and the breaker latched approximately 

20 out of twenty tries, with prop wipe now well within tolerance, a stronc 

predictor of continued reliable latcking. 

adjustment back within tolerance, the breaker was tested further with its 

second prop spring removed. Predictably, it 1 atched re1 i ably, but with 

marginal prop wipe. On the basis of these observations, it was predicted that 

Maine Yankee would likely find excessive preload on the opening springs (or too little preload or weakened closing springs) of breaker 6-53 that had 

required a stronger second prop spring (of the gold-colored type intended to 

be used only as a main prop spring) to make it latch reliably. The inspectors 

concluded that the technical problems with Maine Yankee's Magne-Blast breakers 

have been adequately resolved. GE's deficiencies in QA, workmanship, and 

commercial grade dedication revealed by the problems at Maine Yankee are 

addressed 1 ater in thi s report. 

3.3 Hasne-Blast Desicrn Reconciliation Project 

This was adjusted to be just within tolerance 

As an experiment, with spring 

The lack of a clear recommendation from GE SBP for adding a second prop 

spring, while arguably not necessary at the time the modification was 

inst'tuted in production breakers, pointed up the need for a review of the 

design change history of the Magne-Blast line to identify any other changes 

about which it would be prudent to notify NRC licensees. 

discussions of these concerns, GE PDS and SBP undertook a joint, comprehensive 

design reconciliation project for Magne-Blast breakers, intended also to 

confirm that all design changes (including materials and process changes) and 

their cumulative effects did not invalidate design basis considerations such 

as seismic qualification. 

Following 

The result of this project was SAL 352.1, "Latest Design Configuration: GE AM 

Type Circuit Breakers and Medium Voltage Switchgear," issued by SBP on July 7, 

1995. SAL 352.1 describes and explains 33 modifications and improvements made 

to Magne-Blast breakers over the years and provides references (e.g., SALs, 

manuals, etc.), ordering information and applicability data. All nuclear 

plants with GE medium-voltage switchgear are on the standard distribution list 

for SBP SALs, unlike most other GE product department SALs that go only to 

a 
customers of record (most of whom were/are distributors or architectengineering firms). 

project documents to identify any modifications or process changes about which 

licensees should be notified. 

review was comprehensive and thorough. 

During this inspection, the inspectors reviewed the 

The inspectors found that the joint PDS/SBP 

In order to ensure that future design changes in Magne-Blast breakers get a 

contemporaneous review for impact on safety-re1 ated appl ications , performance 

and qualification, GE NE and GE SBP have instituted a process in which GE NE 

is given an opportunity to review proposed changes before they are 

implemented. The inspectors reviewed the PDS procedure SBO-DP-02 (Revision 

dated January 10, 1995), "Desktop Instruction for Design Change Review at the 

Specialty Breaker Operations." The procedure called for review Magne-Blast 

(Type AM and AMH) design change notices (DCNs) for impact on safety related 

functional performance (including qualification), dedication and maintenance, 

requiring the maintenance activities to be notified. Although not stated in 

the procedure, PDS stated that affected customers would also be notified of 

changes via SALS, SILs or other correspondence. 

3.4 Incorrect TriD Units Installed in Pilqrim Breakers 

In addition to the ASD workmanship and PDS QA concerns involving work on Maine 

Yankee breakers cited above, the inspectors investigated the circumstances 

surrounding a 10 CFR Part 21 notification (NRC Part 21 Log No. 94-265) from 

Boston Edison Company (BECo) regarding incorrect trip units installed by ASD 

in two spare GE AK-2A-50 low-voltage, metalclad circuit breakers. Under PDS 

QA controls and supervision, ASD had overhauled, upgraded, and dedicated the 

spare breakers for safety-re1 ated service at BECo's Pi 1 grim Nuclear Power 

Station (Pilgrim) per BECo purchase order (PO) No. RRR001864 to PDS, dated 

February 22, 1994. The PO specified that RMS-9 trip units should be installed 

in Pilgrim's AK-2A-50 breakers with an "LST1" trip characteristic indicating 

that they would have long-time and short-time trip functions. During shop 

testing of the breakers at Pilgrim prior to installation, BECo discovered that 

ASD had installed RMS9 trip units with an "LSIT1" trip characteristic, 

indicating 1 ong-time, short-time and instantaneous trip functions. These 

breakers were to be used as feeder breakers from 480-Vac bus "MCC B2" to "MCC 

86." 

been installed in this application, the result could have been a loss of 

breaker trip coordination. An electrical fault in a load powered from a 

switchboard fed by one of the affected feeder breakers could have caused the 

feeder breaker to trip before the affected load breaker, resulting in the 

unnecessary loss of other loads, including safety-related equipment, on that 

switchboard. 

Had any of the affected breakers with an instantaneous trip function 

The inspectors found that ASD received and inspected the breakers (serial 

numbers 256A9428-207-3EL and 256A9428-217-5EL) on March 15, 1994. On June 26, 

1994, PDS issued a work authorizing document called an apparatus requisition 

(AR) to ASD for the overhaul and upgrade of the breakers. This AR included a 

summary of the technical and QA requirements and correctly reflected BECo's PO 

requirement for the LSTl trip characteristic in the RMS9 trip units to be 

installed in the breakers. PDS issued another AR for the replacement parts to 

ED&C, the GE product department for low-voltage breakers, on July 7, 1994, but 

9 
the parts AR incorrectly specified control (catalog) number TK503T1604 for an 

LSITl RMS9 with the instantaneous trip function instead of the correct control 

number, TK503T1606, corresponding to an RMS9 with the LSTl trip function. PDS 

QA personnel at ASD received and inspected the incorrect LSITl RMS-9 kits on 

August 15, 1994, but did not detect the error because of the erroneous control 

number on the parts AR, against which PDS receipt-inspected the trip unit 

kits. ASD installed the incorrect LSITl RMS9 trip unit kits and then tested 

them satisfactorily as if they were supposed to have instantaneous trip 

functions. 

certifications on August 31 and September 1, 1994 respectively. 

discovered the error on September 19th, it returned the breakers to ASD for 

rework. 

after sat i sfactory completion of post-i nstall at i onldedicat i on testing, shipped 

the breakers back to Pilgrim on September 23, 1994. 

PDS shipped the two breakers to Pilgrim with GE NE product quality 

When BECo 

PDS had ASD install the proper replacement LSTl RMS9 trip units and, 

PDS initiated nonconformance report number 1EH4M-05 on September 19, 1994, to 

document the deficiency. The cause of the nonconformance was identified as 

the Ifl" in "LSTl" being mistaken by the person ordering parts (and by the ASD 

shop technicians) as an "I" for instantaneous, even though the I is supposed 

to come before the T in the standard trip characteristic designation format 

for a long-time, short-time, instantaneous trip unit, i .e., LSITl. The 

disposition was to rework the breakers by installing the LSTl RMS9 trip units 

after dedicating them to the applicable dedication specification. The action 

to prevent recurrence was to establish the practice of ensuring that ASD 

personnel received legible copies of the actual customer POs. In addition, 

PDS completed a 2-hour training session on October 26, 1994, which included a 

discussion of attention to detail, verbatim compliance, and a review of the 

applicable sections of the QA manual. 

BECo Supplier Finding Report 94-47, issued to PDS, addressed this issue to 

assure its resolution. In response to this finding, PDS committed to have the 

QC Supervisor verify both parts requisitions and work authorizing requisitions 

in the future to assure conformance to the purchase order requirements. In 

addition, the QC Supervisor would verify that parts conform to the PO 

requirements at receipt inspection. Also, the work authorizing documents 

would be more specific to detail the circuit breaker overhaul process. The 

work authorizing document is the basis from which the work controlling 

documents (travellers) are generated. Finally, the Project Manager and QC 

Supervisor's review of the travellers would require the documentation to be of 

sufficient detai 1 to insure the purchase order requirements are appropriately 

delineated with a requirement for verification. PDS did not change any 

procedures to document this process because it was already described in 

existing procedures. 

The inspectors reviewed several recent purchase orders and their associated 

support documentation (i .e. parts requisitions, work authorizing requisitions, 

travellers, and PQCs) and found them to be of adequate detail and to correctly 

reflect the purchase order requirements. The inspectors concluded that PDS 

had taken adequate corrective action and preventive measures, particularly in 

view of the training conducted on the lessons lear,,cd. 

vendor's action is considered satisfactory for closing out the Part 21 Report. 

Accordingly, the 

10 
3.5 QA Corrective Action 

During the inspection of this area, the inspectors toured the nuclear shop 

area, segregated from the general shop floor and dedicated to work on nuclear 

safety-related low and medium-voltage breakers. While looking at jobs in 

progress, examining parts, equipment, and documentation, and interviewing 

technicians, the inspectors observed a Magne-B1 ast breaker frame and mechanism 

assembly (without its arc chutes) being lifted by an overhead travelling hoist 

and moved about the shop floor. The inspectors noted that the hooks on the 

ends of the chain lifting bridle were not engaged with the lifting holes 

provided in the upper horizontal frame members, but rather had simply been 

hooked under the edges of those frame members. When asked, the technician 

performing this operation stated that the assembly was out of balance without 

its arc chutes so that using the designed lifting points would not lift the 

assembly straight. The inspectors noted that if the otherwise unrestrained 

hooks were to slide along the frame members should the assembly tip during 

transit, they could trip at the ends, come loose, dropping the assembly or at 

least loosing control of its attitude. In either case, the breaker may be 

damaged and cause injury to personnel. No attempt had been made to use a 

third lifting point, for example with lifting strap(s) in conjunction with the 

designed safe main lifting points to compensate for the unbalanced condition 

while maintaining positive control over the assembly. This situation was 

brought to the attention of the shop supervisor and management. 

The inspectors asked to see the procedural guidance governing this type of 

operation and were shown the only such written ASD guidance said to be 

available which was GE NE Plant Services & Projects Department Procedure SBO002, "Shipping, Handling, & Storage" (Revision 1, June 7, 1993). Paragraph 

4.2.1 of this procedure required that items be handled "...in a manner which 

will prevent physical damage and preserve the quality of the item and the 

container." Paragraph 4.2.4 stated, in part: "The weight, lifting point or 

center of gravity indicated on the crate, skid or package by the shipper shall 

be utilized to ensure proper handling during unload, transfer between 

carriers, and loading. A1 though the general admonishment of Paragraph 4.2.1 

to handle items in a safe manner could be thought to apply to moving heavy 

objects around in the shop, the more specific requirement in 4.2.4 on use of 

proper lifting points clearly applies to other handling situations. 

no other guidance offered that addressed the unsafe practice observed. 

Therefore, the inspectors concluded that contrary to the requirements of 

Criterion XIII, "Handling, Storage and Shipping," of 10 CFR Part 50, 

Appendix B, GE NE PDS had not established adequate procedures, nor, in the 

absence of procedures specifically governing intra-shop handl ing practices, 

had PDS ensured proper interpretation of and compliance with the intent of 

existing handling, storage, and shipping procedure SBO-002, which could be 

deemed to be imp1 icitly applicable, to control handl ing of Magne-Blast circuit 

breakers being moved about the service shop in a manner that would ensure that 

they would not be damaged. 

practices observed were also an occupational safety and health concern. 

Accordingly, the unsafe handling practice is cited as Nonconformance 95-01-01. 

There was 

The inspectors noted that the unsafe handling 

11 

22 
3.6 Failures of Maane-Blast Trip Coil Cranks 

Magne Blast breaker trip cranks had failed on several occasions at the 

Tennessee Valley Authority's (TVA's) Watts Bar Nuclear Plant (WBN) and one 

additional failure had been reported at the Shoreham plant. 

reports to the NRC pursuant to 10 CFR Part 21 were being tracked under NRC 

Part 21 Report Log No. 93-304. The trip crank is a piece part of the MagneBlast ML-13 operating mechanism, GE Part Number 105C9316GOOl. 

stamped, formed and drilled steel plate with a steel pin socket-welded into it 

(See Figure 1 of Appendix A to this report). The function of the trip crank 

is to convert the linear force of the trip coil armature or plunger to the 

torque needed to rotate the breaker trip shaft. Failure of the trip crank 

could prevent any electrical trip of the breaker. The known failure mode is 

fracture of the weld that fastens the trip coil armature link pin to the plate 

portion of the crank under shock loading shear reaction stress on the pin. 

The failure 

The crank is a 

In response to the original failures, GE SRB changed the manufacturing process 

for the crank by requiring some reinforcement (on the weld fastening the pin 

to the plate) to be left on the back side of the plate and not be ground off. 

Now specified on the ML-13 drawing (SBP Drawing No. 0105C9316/DCN 083-88-003, 

May 18, 1988, View "A" weld detail for Group 001 Trip Crank) is: "1/32 TO 1/16 

WELD TO REMAIN AFTER GRINDING FLAT SURFACE, DO NOT GRIND FLUSH." The 

inspectors examined a crank that was in stock at ASD, but tagged for use in 

non-safety-related applications only. The crank plate was bent at the end and 

the pin was cocked about 10 to 15 degrees. It appeared that this crank would 

very likely fail in service, albeit presumably in non-safety-related service, 

but fail nonetheless. 

The inspectors toured the SBP production floor, including the area where the 

pins are welded to the trip cranks. 

interviewed one of the factory workers who performed the welding on this 

particular part (among others). The factory worker described and partially 

demonstrated how he did the welding in terms of setup, but admitted that he 

had no fixture to ensure that proper penetration and alignment of the pin was 

maintained and it was not clear that he did any inspection of the pins before 

use, but welded them into the cranks as they were obtained from a parts bin. 

He intimated that the person who likely welded most -.f the failed trip cranks 

and ground the reinforcement flush had allowed some "cold welds," i.e. lack of 

fusion, and probably lack of adequate penetration as well, depending largely 

on the conformance of the pins to the drawing. 

The inspectors examined the setup and 

During the tour of the ASD nuclear work area, the inspectors noted that the 

ML-13 mechanisms of two AM-4.16 breakers undergoing initial inspection and 

disassembly had trip cranks that were susceptible to failure with their weld 

reinforcements ground off and the transition or interface between the pin and 

the plate clearly visible. 

the inspection and teardown had not documented this condition in the 

inspection record, but PDS stated that the technicians were aware of the 

problem and that in this case, the technician had not come to that portion of 

his inspection yet. PDS confirmed subseqGent to the inspection that the trip 

cranks in question had been replaced and the customer notified with 

recommendations for inspection of other breakers. 

The inspector noted that the technician performing 

12 
To ensure that faulty trip cranks are not used in safety-related breakers, PDS 

is now relying on the 1988 design change and it< dedication process to screen 

out improperly fabricated cranks, as evidenced by, for example, the fact that 

the unsatisfactory crank found by the inspector had been rejected for nuclear 

use. Although the amount of chamfer on the small end of the pin that is 

inserted into the plate and the penetration depth allowed by the depth to 

which the shoulder is machined into the pin both significantly affect the 

amount of pin surface area in contact with the weld filler material, these 

attributes are not specified to be checked during manufacture, as they cannot 

be seen afterward. 

solely on a 100-lbf static load test and the presence of 1/32 to 1/16 inch of 

pin weld reinforcement on the back side of the plate. With respect to faulty 

cranks in the field, SBP is considering issuing a SAL on this problem and in 

the mean time PDS has put out instructions to the service shops to inspect the 

cranks as breakers come through for service and replace the cranks as 

necessary. 

satisfactory in resolving this issue, but that it would be complete only with 

the issuance of a SAL. 

The dedication at the time of this inspection relied 

The inspectors concluded that the vendor's action thus far was 

3.7 Evaluation of 6E NE PDS Program at SBP for Dedication of Commercial 

Grade Items 

GE NE PDS reported upgrading its process for dedicating the commercial grade 

components used in manufacturing and overhauling safety-related breakers. 

During this inspection, the inspectors addressed the problems with switch 

dedication by the PDS dedication facility at SBP. 

3.7.1 Program Procedures Review 

The program and process of dedication of commercial grade items (CGIs) at all 

GE NE dedication facilities in general is described and prescribed by a 

hierarchy of procedures under the GE NE nuclear quality assurance program. 

The highest tier GE NE procedure dedicated to dedication is Engineering 

Operating Procedure (EOP) 65.2.20, "Dedication of Commercial Grade Items. The 

Vendor Inspection Branch had reviewed this procedure extensively in the past 

during inspections in San Jose. Accordingly, during this inspection it was 

givx a cursory review that confirmed no substantive changes since the last 

formal review by the NRC. Nevertheless, the inspectors reviewed the latest 

effective revision of EOP 65-2.20, Revision 5, dated July 27, 1994. 

In the system in use at GE NE'S dedication facility at SBP, the next tier of 

dedication procedures are called dedication specifications. The inspectors 

reviewed those pertinent to CGIs of particular focus during this inspection; 

namely (1) No. 24A1113, "Dedication of Magne-Blast Switchgear Parts" (Revision 

2, dated November 30, 1994), (2) No. 24A1114, "Dedication Specification for 

Hardware and Materials" (Revision 0, dated June 4, 1993). In addition, local 

procedures for general dedication guidance were reviewed: (3) SBO-DP-01 , "Desktop Instruction for Dedication of Material at the Specialty Breaker 

Operations" (revision dated May 3, 1993) , (4) SBO-DP-03, "Desktop Instruction 

For Dedication/Inspection Tolerances at the Specialty Breaker Operation" 

(revision dated January 17, 1995), and (5) No. 24A1115, "Dedication Specification for ED&C Low Voltage Switchgear Parts," Revision 0, April 25, 1992. 

13 
3.7.2 Dedication package review 

The inspectors reviewed several dedication packages at SBP for rep1 acement 

parts to be supplied to ASD for installation in overhauled/modified circuit 

breakers or to GENE for direct shipment as spare/replacement parts to the 

licensees. They included switches, shafts, clutches, trip cranks, and 

secondary couplers. 

properly identified and verified (with the few exceptions noted below) to 

assure that the components would perform their intended safety function under 

all design basis conditions and not fail in a manner adverse to safety. In 

particular, the inspectors reviewed the dedication specifications, critical 

characteristics matrices and verification records for recently dedicated 

CR2940 contact blocks (both the NO and NC types), Magne-Blast breaker trip 

cranks , and secondary coup1 ers . The veri f i cat ion records i ncl uded : the 

initiating AR, and for each line item; a traveler, CC forms, dedication check 

sheets, instrument use and calibration log sheets, etc.. In addition, the 

dedication specification for ED&C low-vol tago switchgear parts, No. 24A1115 

(Revision 0, dated April 25, 1992) was reviewed for technical adequacy even 

though the recent problems with incorrect AK breaker trip units (Pilgrim) 

discussed above were related to identification of correct material during 

procurement as opposed to adequacy of critical characteristic verification 

which was not in question at Pilgrim. 

In general , the critical characteristics were found to be 

3.7.2.1 Critical Characteristics Matrix Form No. 00181, Revision 1 (no 

revision date), for "microswitch (N.C.)," part no. Q0456A0866P006, was being 

used to specify and document the verification of critical characteristics for 

the dedication of CR2940U301 contact blocks (which are normally closed (NC) in 

their shelf state) for use in Magne-Blast ML-13 operating mechanisms. 

Magne-Blast breakers, this NC contact block is used solely for power switch 

52/SM-LS, contacts 5 and 6 discussed above. 

selection of critical characteristics appeared to be reasonably complete, but 

pointed out several weaknesses that could reduce the effectiveness or 

usefulness of the matrix. 

In 

The inspector noted that the 

0 The description block described the item as a normally open control 

switch which it is not instead of normally closed which it is (also 

inconsistent with the brief part description ir the header of the form). 

0 The function was described as "part of the breaker control circuitry" 

instead of, for example, "contacts 5 and 6 of power switch assembly 52/SM-LS as shown on "Typical Wiring Diagram," in Instruction Book GEI88671. Returns to its normal closed state when its plunger is released 

when closing spring is fully charged and enables closing spring release 

circuit. 

fully charged condition. 

Opens as soon as closing spring charging mechanism leaves 

0 The failure mode and effect was characterized as "loss of control 

circuitry" instead of circuit function; or more specifically, an open 

causing inability to close the circuit breaker remotely, a short or 

ground possibly causing loss of 125-Vdc controi power. 

14 

25 
0 Under the critical characteristic or inspection attribute of "General 

ADpearance, I' the requirement was to "Verify [vi sua1 ly] the existence of 

internal threads and thread sizes; I' a1 though the fasteners or screws 

were not mentioned. 

0 Under "Check Number of Contacts-1," the method of examination was 

"visual;" yet the contacts are hidden from view. The only visible 

attribute indicating the number of contact (sets) is that the block 

should have only two terminals. 

0 Under "Operation-Continuity,'' the method of examination/test was given 

as "Operate switch and verify open/close function," but it was not 

stated how this was to be done. 

0 Dielectric strength was listed as a critical characteristic, but the 

inspector questioned the appropriateness or usefulness of a 1500-volt 

high potential test for one minute on a tiny contact block only designed 

for 120 Vac or 125 Vdc when insulation resistance measurement at 500 or 

250 Vdc would be more than adequate. Also, testing terminals to ground 

and line to load while open (i.e., across open contacts) was not 

specified. In general, the acceptance criteria were given as part of 

the critical characteristic, but for dielectric strength, the acceptance 

criterion (typically given in terms of maximum allowable leakage current 

or stated as no breakdown) was not stated. 

3.7.2.2 Critical Characteristics Matrix Form No. 00296, Revision 1 (no 

revision date), for "Crank," Piece Part of ML-13 Mechanism, GE Part Number 

105C9316G001, was being used to specify and document the verification of 

critical characteristics for the dedication of CR2940U310 contact blocks for 

use in Magne-B1 ast ML-13 operat i ng mechani sms . 

0 The failure effect was described as "could prevent any electrical trip 

of the breaker." A known failure mode, fracture of the pin weld under 

shock loading shear reaction stress on the pin, was not mentioned. 

0 No NDE except visual. 

0 Weld proof test was specified as a 100-lbf minimum static ("dead") load 

applied "normal to axis." The inspector questioned whether this test 

effectively simulated or was reasonably equivalent to the severe shock 

loading reaction stresses generated during operation. 

design engineer stated that he had also tested production trip cranks by 

operation on a shop test breaker and would discuss this method with the 

onsite PDS staff for dedication purposes. 

The breaker 

0 Also, the amount of chamfer on the small end of the pin that is inserted 

into the plate and the penetration depth allowed by the depth to which 

the shoulder is machined into the pin both significantly affect the 

amount of pin surface area in contact with the weld filler material, yet 

these attributes were not specified to be checked during manufacture, 

and,they cannot be seen afterward. As stated above, the dedication 

15 

26 
relies solely on the pull test and the presence of 1/32 to 1/16 of weld 

reinforcement on the back side of the plate. 

3.7.2.3 Critical Characteristics Matrix Form No. 00315, Revision 0 (no 

revision date) , for "Secondary Disconnect , 'I Piece Part of ML-13 Mechani sm, GE 

Part Number 108B1931G001, was being used to specify and document the verification of critical characteristics for the dedication of secondary coupler 

blocks for connecting the control cable from Magne-Blast breaker cubicle to 

the ML-13 Magne-Blast operating mechanisms. The only comment on this matrix 

form was the lack of insulation resistance measurements or dielectric 

withstand test. PDS explained that at the time the revision of this matrix 

form was in use, PDS used material verification and surface finish as 

attributes to ensure that the electrical characteristics would be acceptable 

as an alternative to insulation resistance or dielectric withstand testing. 

Current matrices, however, have added the dielectric withstand or Hi-Pot test. 

The inspectors concluded that with the improvements recently implemented (such 

as revising sample sizes), the commercial grade dedication activities of PDS 

at SBP (with the minor exceptions discussed above) were in general being 

conducted consistent with the intevt 3f 10 CFR Part 50, Appendix B, NRC 

Generic Letters 89-02 and 91-05, EPRI Report NP-5652, and GE NE procedures. 

3.8 Review of GE NE Part 21 Procedures 

The inspectors asked to see the procedures adopted by PDS pursuant 10 CFR 

21.21(a) and were given a copy of GE Nuclear Energy Policy and Procedures 

Manual , NEDE-31746, Procedure 70-42, Revision Dated August 1994, "Reporting of 

Defects and Noncompliances Under 10 CFR Part 21." The inspectors reviewed 

this procedure with the following comments: 

3.8.1 

the same procedure that had been the subject of considerable discussion during 

a previous NRC inspection AT GE NE, San Jose, and had been substantially 

revised to address NRC concerns. A1 though these particular inspectors had not 

seen this procedure before, they were familiar with the issues previously 

discussed and noted that the procedure appeared to have been adequately 

revixd to address the concerns raised in previous NRC inspections of GE NE at 

San Jose. However, the inspectors did identify certain sections in which the 

language of the procedure was not consistent with the applicable requirements 

of Part 21. 

After completing their review, the inspectors learned that this was 

3.8.2 

aware of deviations that may be potential safety concerns (PSCs) to advise the 

Safety Evaluation Program (SEP) Project Manager in writing. However, 921.21 

requires that deviations in basic components delivered to or offered for use 

at NRC-licensed facilities be evaluated to identify defects, not just 

deviations that may be potential safety concerns. That is, all deviations are 

supposed to be evaluated to determine if they are safety concerns. To add the 

qualification, 'I.. .that may be potential safety concerns," to this provision 

of the procedure may act effectively to screen out deviations from consideration by SEP to determine even if they are PSCs because it calls for a concluParagraph 3.1, "Discovery of a PSC," required employees who become 

16 

27 
sion by the employee that he or she may not be qualified or empowered to draw. 

3.8.3 

10 CFR Part 21, 521.3, and as distinguished from deviations) which must also 

be evaluated per 521.21(a) to determine if they are associated with a 

substantial safety hazard, or reported to affected 1 icensees or purchasers per 

521.21(b). A basic component delivered to or offered for use at an NRClicensed facility by GE NE PDS could fail to comply in some respect with the 

Atomic Energy Act of 1954, as amended, or any rule, regulation, license, or 

order of the Commission (NRC) without being a deviation, per se, Le., a 

departure from a technical procurement specification, yet still require 

eval uat i on under 521.21 (a) or reporting to customers under 521.21 (b) . 

3.8.4 Paragraph 4.3, "Transfer of Information, I' provided for the customer 

notification discussed in 921.21(b), but the language used in Paragraph 4.3 

was not fully consistent with 521.21(b) and, in the judgement of the 

inspector, would not ensure that the intent of 521.21(b) would be met in all 

cases. 

(meaning all affected licensees or purchasers of whom the vendor should have 

reason to be aware) be notified of deviations and failures to comply that the 

vendor determines it is unable to evaluate per 521.21(a). 

language of Paragraph 4.3: 

of the application of the deficient product," instead of simp?y, "affected 

1 icensees or purchasers," as stated in 521.21(b) effectively restricted the 

disbursement of the information to a narrower scope than that intended by 

521.21(b). Although this provision (521.21(b)) is not currently required by 

Part 21 to be part of the procedures adopted pursuant to the regulation, in 

order to be consistent with the intent of 921.21(b), the procedure would need 

to require the vendor to identify from its records all licensees or purchasers 

who could be affected by the deviation or failure to comply, and notify all of 

them within the five working days of discovery prescribed in the regulation. 

Procedure 70-42 did not address failures to comply (as defined in 

Section 21.21(b) requires that affected 1 icensees or purchasers 

However, the 

"...to the licensee cir purchaser who has knowledge 

3.8.5 Paragraph 4.7, "Reportable Condition," required that if the evaluation 

concludes that a defect exists, the Responsible Officer be notified within 

five working days, but it did not address notifying the responsible officer of 

a failure to comply that an evaluation might have determined was associated 

with a substantial safety hazard. 

Previously this procedure addressed conditions characterized as those that 

could create a substantial safety hazard or lead to exceeding a technical 

specification safety limit, "i .e. , a defect." Although failures to comply 

were not explicitly mentioned per se, the description of the conditions of 

interest could be deemed to include failures to comply in that the failures to 

comply that ultimately are to be reported to the NRC under Part 21 (within the 

scope of provisions required to be part of procedures by §21.21(a)) are those 

that could be associated with a substantial safety hazard. However, the 

inclusive nature of the language of the procedure was negated in Paragraph 4.7 

by departing from the inclusive language used previously in the procedure and 

instead only mentioning defects. 

following the procedure as written verbatim without reference to Part 21, 

could lead to a situation in which a condition (PSC or even PRC) that could 

The problem that could arise is that 

17 

28 
create a substantial safety hazard might not be considered, let alone reported 

to the responsible officer, because it did not start out as a deviation, i.e., 

a departure from a technical procurement specification, but might still have 

been in some respect in noncompliance with the Atomic Energy Act of 1954, as 

amended, or any rule, regulation, order, or license of the Commission. 

In addition, the procedure then stated, in part: "If the PRC is determined to 

be a reportable defect by the Responsible Officer, the SEP Project Manager 

shall notify the NRC ....'I The language here implied that even after the 

evaluation has concluded that the PRC is a defect, and hence is by definition 

reportable to the NRC, the responsible officer would make a determination of 

reportability separate from the conclusion of the evaluation that the PRC was 

a defect. 

associated with substantial safety hazards) as identified on the basis of an 

evaluation, be reported to a director or responsible officer who then must 

cause them to be reported, without need for further evaluation, to the NRC. 

However, Part 21 requires that defects (and failures to comply 

The inspector discussed his concerns with the language used in Procedure 70-42 

with PDS representatives, with GE NE'S Manager of NS & PO Quality, and with 

the staff of GE NE SEP cognizant over thiz procedure. The inspector explained 

that the observations discussed above were based on the NRC staff's positions 

on Part 21 requirements, and interpretations of those requirements where 

necessary by the NiiC Office of General Counsel. 

identified were considered minor violations and as such would not be cited as 

violations in accordance with the NRC's Enforcement Pol icy, NUREG 1600. 

The procedural deficiencies 

4.0 PERSONNEL CONTACTED 

Char1 es Tayl or , Project Manager , GE-ASD 

Edward Dugan, Design Engineer, GE-SBP (215-726-2316) 

Brian Kennedy, QA Supervisor, GE NE 

Joseph La Clair, QA Engineer, GE NE, PDS 

George Wetsell, Manager, Plant Services and PM&C Quality, GE NE 

Forrest Hatch, Manager, NS&PO Quality, GE NE 

George Sanders, Lead Engineer, GE NE PDS 

Charlie Vickers, Project Manager, GE NE (ASD ext. 657/8) 

Charles Rodgers, Field Representative, GE NE PDS 

Brian Forrest, ASD Project Manager (215-289-0400) 

Richard A. Thielking, ASD Switchgear Specialist 

Tom Connolly, Manager, Northeast Region, GE NE PDS 

ADDendi x 

A. Trip Crank for 6E Magne-Blast Circuit Breakers 

18 
APPENDIX A 

~~~ 

TRIP CRANK for 

Circuit Breakers 

with ML-13type 

mechanisms 

GE Magne-Blast Q IQ 

3 NOTE: This weld detail on 

1988 revised drawing slates: 

"DO NOT GRIND FLUSH." 

GE Part No. 105C9316G001 

A 1 

Weld ground flush 

in ade q u ate ch amte r Ion ain 

SECTION A-A' I 

30 

Figure 1 

19 
November 22, 1995 

Mr. Richard G. Knoblock, President 

Pacific Scientific Company HTL/Kin-Tech Division 

22715 Savi Ranch Parkway 

P.O. Box 87019 

Yorba Linda, CA 92687-8719 

SUBJECT: NRC INSPECTION NO. 99900255/95-01 

Dear Mr. Knoblock: 

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

(NRC) inspection of Pacific Scientific Company, HTL Kin-Tech Division, 

conducted by Robert L. Pettis, Jr. and Billy Rogers, of this office, on 

September 25-28, 1995. 

Pacific Scientific quality program and the implementation of that program as 

it relates to the manufacture of mechanical shock suppressors and spare parts 

supplied to the nuclear industry as safety-related. 

The NRC inspection team conducted an evaluation of the 

The NRC inspection team reviewed documentat ion, procedures, and representative 

records, conducted interviews and held discussions with members of your staff. 

On the basis of this inspection, the inspection team determined that the 

implementation of the Pacific Scientific quality assurance program failed to 

meet certain requirements of Appendix B to Part 50 of Title 10 of the Code of 

Federal Requlations and 10 CFR Part 21. 

The enclosed i nspecti on report contains a detai 1 ed di scussion of the areas 

exami ned. 

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

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

Should you have any questions concerning this report, we will be pleased to 

discuss them with you. Thank you for your cooperation. 

Sincerely, 

Robert M. Gallo, Chief 

Speci a1 Inspect ion Branch 

Division of Inspection and Support Programs 

Off i ce of Nucl ear Reactor Regul at i on 

ORIGINAL, SIGNED BY 

Docket No.: 99900255 

Enclosures: 1. Notice of Nonconformance 

2. Inspection Report No. 99900255/95-01 
NOTICE OF NONCONFORMANCE 

Paci f i c Sci ent i f i c Company 

Yorba Linda, California 

Docket No.: 99900255 

Based on the results of an NRC inspection conducted at the Yorba Linda, 

California, facility of Pacific Scientific on September 25 through 28, 1995, 

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

with NRC requirements. 

I. 

11. 

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

and Drawings," states, in part, that activities affecting quality shall 

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

type appropriate to the circumstances and shall be accomplished in 

accordance with these instructions, procedures, or drawings. 

Contrary to the above, Pacific Scientific Company (PSC) did not have a 

procedure in place for the replacement of commercial grade parts with 

safety-related. 

process in which surplus snubbers, purchased from material brokers, were 

refurbished and sold as new to 1 icensees. 

Such parts were routinely changed out during the 

(95-01-01) 

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

and Drawings," states, in part, that activities affecting quality shall 

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

type appropriate to the circumstances and shall be accomplished in 

accordance with these instructions, procedures, or drawings. 

Pacific Scientific Standard Operating Procedure 07.104, "Supplier 

Approval," dated May 18, 1995, stated that all suppliers who furnish 

material or supplies for use in a saleable product shall be listed on 

the appl icable PSC Approved Suppliers List (ASL) . 

Contrary to the above, Walden Industrial Supply Company and Vabcor Inc., 

companies from which PSC purchased surplus market snubbers for resale to 

licensees as safety-related, were not listed on the PSC ASL. (95-01-02) 

111. Criterion I of Appendix B to 10 CFR Part 50, "Organization," states, in 

part, that the authorities and duties of persons performing quality 

assurance activities be clearly established and delineated in writing. 

These activities include the quality assurance functions of assuring 

that an appropriate qual i ty assurance program is establ i shed and 

effectively executed and verifying that activities affecting safetyre1 ated functions have been correctly performed. 

Enclosure 1 

32 
PSC Quality Assurance Manual Section 0.20, "Management Statement, 

states that the Manager of Qual i ty Assurance is given the responsi bil i ty 

and authority to maintain the Quality Assurance program and its 

implementation. Further, the Manager of Quality Assurance is given the 

freedom to identify quality assurance problems, initiate actions which 

results in solutions, verify implementation of solutions to those 

problems, and the authority to stop any work which violates any 

provisions of the Quality Assurance Manual. 

Contrary to the above, during the period of approximately 1992 through 

1994, the PSC activities involving the purchasing, receiving inspection, 

disassembly, parts repl acement, reassembly, 1 abel plate repl acement, and 

testing of snubbers (activities affecting safety-related functions) 

purchased on the surplus market and resold to licensees as safetyrelated, an activity affecting quality, occurred without the quality 

oversight or knowledge of PSC's Quality Assurance Manager and, as a 

result, the Quality Assurance Manager was not able assure that an 

appropriate qual i ty assurance program had been establ i shed and 

effectively executed or verify that activities affecting safety-related 

functions have been correctly performed. (95-01-03) 

IV. Criterion XVI of Appendix B to 10 CFR Part 50, "Corrective Action," 

states, in part, that the .identification of the significant condition 

adverse to quality, the cause of the condition, and the corrective 

action taken shall be documented and reported to appropriate levels of 

management. 

Pacific Scientific QAM, Section 11 "Nonconformance and Corrective 

Action," Revision 6, dated December 15, 1993, requires the documentation 

of corrective action on a Failure Analysis and Corrective Action Report. 

Contrary to the above, PSC did not complete a Failure Analysis and 

Corrective Action Report, as required by PSC procedure, for a PSA Model 

100 snubber which failed an activation test. The snubber, purchased as 

surplus material from Vabcor Inc., was part of an order for Pennsylvania 

Power & Light's Susquehanna Steam Electric Station. (95-01-04) 

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

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

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

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

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

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

include for each nonconformance: 

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

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

corrective actions and preventive measures were or will be completed. 

(1) a description of the steps that have 

Dated at Rockvi 11 e, Mary1 and 

this day of November, 1995 
REPORT NO.: 

ORGAN I ZATION : 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY: 

INSPECTION DATES: 

LEAD INSPECTOR: 

OTHER INSPECTORS: 

REVIEWED BY: 

APPROVED BY: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF INSPECTIGN AND SUPPORT PROGRAMS 

99900255/95-01 

Richard G. Knoblock, President 

Pacific Scientific Company 

HTL/Kin-Tech Division 

22715 Savi Ranch Parkway 

P.O. Box 87019 

Yorba Linda, Cal ifornia 92687-8719 

Steve Palm 

Quality Assurance Manager 

Manufacturer of various model mechanical shock 

suppressors supplied as safety-related. 

September 25-28, 1995 

Robert L. Pettis, Jr., P.E. Date 

Vendor Inspection Section (VIS) 

Special Inspection Branch (SIB) 

Division of Inspection 

and Support Programs (DISP) 

Office of Nuclear Reactor Regulation (NRR) 

Billy H. Rogers, VIS/SIB/DISP/NRR 

Gregoryd.&wal ina, Chief 

VIS/SIB/DISP/NRR 

Date 

Robert M. Gallo, Chief 

SIB/DISP/NRR 

Date 

Enclosure 2 
1 SUMMARY OF INSPECTION FINDINGS 

During this inspection, the NRC inspection team evaluated the implementation 

of the Pacific Scientific Company (PSC) quality assurance (QA) program related 

to the manufacture of various model mechanical shock suppressors (snubbers) 

supplied to the nuclear industry as safety-related. 

conducted to determine PSC's compliance with the requirements of Appendix B to 

Part 50 of Title 10 of the Code of Federal Requlations (Appendix B) and the 

provisions of 10 CFR Part 21 (Part 21). 

technical information, procedures and representative records, conducted 

interviews, held discussions and observed manufacturing activities. 

The inspection was 

The inspection team reviewed 

1.1 Violation 

Contrary to 10 CFR 21.21, which states that corporations subject to the 

regulations must adopt appropriate procedures to include specific requirements 

related to the length of evaluation of deviations, preparation and submittal 

of interim reports, and the informing of a director or responsible officer 

that a defect or failure to comply exists, PSC did not include the 

requirements of 10 CFR Part 21.21 in Standard Operating Procedure (SOP) 01.07, 

"Compliance With 10CFR21," dated August 10, 1993. (Non-Cited Violation) 

1.2 Nonconformances 

1.2.1 Contrary to Criterion V of Appendix 6, "Instructions, Procedures, and 

Drawings," PSC did not have a procedure in place for the replacement of 

commerci a1 grade parts with safety-re1 ated or dedicated parts used during the 

process of upgrading surplus secondary market snubbers to current production 

line standards. (95-01-01) 

1.2.2 Contrary to Criterion V of Appendix 6, "Instructions, Procedures, and 

Drawings," and PSC SOP 07.104, "Supplier Approval ," dated May 18, 1995, PSC 

had not listed Walden Industrial Supply Company and Vabcor Inc., companies 

from which PSC purchased surplus snubbers which were resold to licensees as 

safety-related, on its Approved Suppliers List (ASL). (95-01-02) 

1.2.3 Contrary to Criterion I of Appendix B, "Organization," PSC performed a 

portion of its safety-related activities without the quality oversight or 

knowledge of the PSC QA manager and these activities were solely conducted 

under the control of production quality control personnel and, as a result, 

the Quality Assurance Manager was not able assure that an appropriate quality 

assurance program had been establ ished and effectively executed or verify that 

activities affecting safety-related functions have been correctly performed. 

1.2.4 Contrary to Criterion XVI of Appendix B, "Corrective Action," and PSC 

QA Manual , Section 11 "Nonconformance and Corrective Action , I' Revi si on 6 , dated December 15, 1993, PSC did not perform a failure analysis and corrective 

action Report, as required by PSC procedure, for a PSA Model 100 snubber which 

failed an activation test. The snubber, purchased as surplus material from 

Vabcor Inc., was part of an ol*der for Pennsylvania Power & Light's Susquehanna 

Steam Electric Station. (95-01-04) 

(95-01-03) 
2 STATUS OF PREVIOUS INSPECTION FINDINGS 

None were reviewed during this inspection. 

3 INSPECTION FINDINGS AND OTHER COMMENTS 

3.1 Bac kqround 

PSC manufactures snubbers which are designed to restrain piping systems and 

associated components from damage resulting from earthquakes and other shocks 

while enabling unrestricted movement for thermal growth. 

on the principle of limiting the acceleration of any pipe movement to a 

threshold level of .02 times the acceleration of gravity (g), which is the 

maximum acceleration that the snubber will permit the piping system to 

experience. Should the piping system experience an acceleration in either 

direction, a braking force will be applied within the snubber necessary to 

limit the acceleration to less than .02 g’s, while not restricting thermal 

expansion. 

The snubber operates 

This braking force is achieved by a ball screw and drum which attempts to 

angularly accelerate an inertia mass. The inertial resistance of the mass 

causes a resilient capstan spring to tighten around a hardened mandrel which 

is part of a structural tube which provides a restrain force against rotation 

of the ball screw. The snubber’s performance is independent of the amount of 

force being applied and at no time does it lock to become a rigid strut. 

Each snubber is built in accordance with the American Society of Mechanical 

Engineers (ASME) Boiler and Pressure Vessel Code, Subsection NF, and to an 

ASME Section I I I, Di vi si on 1, NCA-3800, approved Qual i ty Assurance Program 

which complies with the requirements of Appendix B, American National 

Standards Institute (ANSI) ANSI N45.2, and the reporting requirements Part 21. 

Each snubber is production tested by PSC in accordance with established 

procedures and ASME Code rules to establish rated capacities and are designed 

and built to ASME Class 1 Code requirements regardless of customer 

specifications. PSC has served the power industry for more than 30 years and 

has supplied snubbers to over 100 nuclear power plants. 

ASME Quality System Certificate (QSC-527) as a material supplier. 

PSC also maictains an 

3.2 Entrance and Exit Meetinqs 

During the entrance meeting, held on September 25, 1995, the NRC inspection 

team met with members of PSC management and staff, discussed the scope of the 

inspection, and established organizational contacts. During the exit meeting, 

held on September 28, 1995, the inspection team summarized its findings with 

PSC management and staff. 

contacted during the inspection. 

The Appendix of this report lists the persons 

3.3 10 CFR Part 21 Proqram 

The inspectors reviewed PSC’s 10 CFR Part 21 program including procedures and 

implementation. PSC had identified several potential deviations in products 

that they had supplied and consequently had performed evaluations in 

accordance with its Part 21 program. 

3 

36 
I In August of 1992, PSC discovered a group of PSA Model 110 pinion gears (part 

number 1801418-011, which did not appear to have been heat treated. PSC 

determined that one batch of gears had not been tested for hardness during the 

initial production of PSA 100 snubbers. 

determined that a PSA 100 snubber using a non-heat treated pinion gear would 

still meet 100% rated load and that the use was not a defect. 

In June of 1993, Arizona Public Service (APS) reported to PSC that snubbers 

which had passed PSC testing had subsequently failed during APS testing. PSC 

performed additional testing on the snubbers, with the oversight of APS, and 

concluded that the APS test failure (high drag) was caused by set up error due 

to a loose adaptor, and that the test error was conservative in that it would 

fail a good snubber but would not pass a bad snubber. PSC concluded that was 

not a deviation or a defect. The inspectors determined that PSC had performed 

adequate evaluations in accordance with their Part 21 program and did not 

identify any concerns in this area. 

The inspectors reviewed PSC's Part 21 implementing procedure, SOP 01.07, 

"Compliance With lOCFR21," dated August 10, 1993, and determined that it did 

not include several items required to be proceduralized by Part 21. Section 

10 CFR 21.21 requires that corporations subject to the regulations adopt 

appropriate procedures to: (1) evaluate deviations and failures to comply to 

identify defects and failures to comply associated with substantial safety 

hazards as soon as practicable, and, except as provided in (2) , in a1 1 cases 

within 60 days of discovery; (2) ensure that if an evaluation of an identified 

deviation or failure to comply potentially associated with a substantial 

hazard cannot be completed within 60 days of discovery of the deviation or 

failure to comply, an interim report is prepared and submitted to the 

Commission through a director or responsible officer or designated person; and 

(3) ensure that a director or responsible officer is informed as soon as 

practicable, and, in all cases, within the 5 working days after completion of 

the evaluation in (1) or (2) if a defect or failure to comply associated with a substantial safety hazard exists. 

The inspectors concluded, based on a review of the Part 21 evaluations 

performed by PSC, that although PSC had failed to proceduralize the 10 CFR 

21.21 requirements, there had not been an occurrence where PSC's actions were 

not in accordance with 10 CFR 21. 

requirements specified in 10 CFR 21.21 constitutes a violation of minor 

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

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

In addition, SOP 01.07 also contained several instances where the terms 

deviation and defect were inappropriately interchanged. 

terms, and their definitions in Part 21, were discussed with PSC who indicated 

that the procedure would be modified to correctly use these two terms. 

inspectors also reviewed PSC's posting as required by 10 CFR 21.6 and 

determined it to be in accordance with the regulation. 

3.4 Processins SurPlus Snubbers 

Discussion with PSC personnel indicated that they had bought surplus snubbers 

(referred to as buybacks) from two material brokers, Walden Industrial Supply 

Company (WISC) and Vabcor Inc. (VI) , performed certain modifications and tests 

and resold them to licensees as safety-related. 

PSC performed an evaluation and 

The failure to proceduralize the 

The use of these 

The 

PSC purchased the surplus 

4 

37 
snubbers, which had been originally sold to licensees as safety-related and 

had never been installed, from WISC and VI. 

inspectors, various PSC personnel stated that this practice had been ongoing 

for approximately ten years. 

During conversation with the 

The inspectors reviewed the records associated with the two companies 

supplying the surplus snubbers. PSC provided purchase orders it had made to 

WISC and VI since 1992 but did not have any purchase orders to either company 

prior to 1992. 

earliest either company had been used for surplus snubbers. PSC was unable to 

provide information on any other companies which might have supplied surplus 

snubbers prior to 1992 (although PSC indicated it had purchased surplus 

snubbers since approximately 1985). 

In further discussion, PSC indicated that this was the 

PSC indicated that the process used to prepare the surplus snubbers for 

safety-related sales to licensees had evolved over the period from 1992 to 

1995. Initially the surplus snubbers were only visually inspected, tested and 

shipped without documentation of the activities. 

current practice which included receiving inspection, disassembly, replacement 

of commercial grade parts with safety-related parts, ASME code reconciliation, 

reassembly, and testing. In addition, PSC replaced the original label plate 

on the surplus snubbers with a new label plate indicating the current 

production information including a new serial number and manufacturing date. 

Each model of snubber was processed to a manufacturing order (MO) specific to 

the model and the model specific MOs were in various states of r e v i s i o n w i t h 

some models of snubbers requiring less work to be prepared for shipment than 

other model s. 

This had evolved to the 

For the current production line, PSC had evaluated all the snubber designs, 

determined that the majority of components should be safety-related, and had 

established a policy and revised the applicable MOs for the various snubber 

models to dedicate the appropriate commercial grade parts used in the 

production of new snubbers. During the initial production of the surplus 

snubbers, PSC indicated that some commercial grade parts had possibly been 

used in the manufacture of the snubbers (if manufactured prior to the 

implementation of PCS's dedication policy). 

processing the surplus snubbers included replacing all commercial grade parts 

with new parts which were safety-related (dedicated) to bring the surplus 

snubbers up to current production line standards. However, PSC was unable to 

provide procedures requiring the replacement of commercial grade parts or 

documentation that this had occurred for various models and orders of surplus 

snubbers from 1992 through 1994. 

PSC's current practice in 

PSC's failure to have a procedure in place for replacement of commercial grade 

parts with safety-related parts, or documentation that this had occurred, 

during the process of upgrading the snubbers to current production line 

standards, in the 1992 through 1995 time period, was identified as a 

Nonconformance with Criterion V of Appendix B (95-01-01). 

3.5 Suppl i er Approval 

The inspectors reviewed SOP 07.104, "Supplier Approval ,I' dated May 18, 1995, 

which established the procedure for the qualification of PSC suppliers. 

Paragraph 3.1.1 stated that all suppliers who provide items for use in the 

nuclear product line would be qualified through surveys or audits. In 

5 

38 
addition, paragraph 3.4 stated that all suppliers who furnish material or 

supplies for use in saleable product shall be listed on the applicable PSC 

ASL. Discussion with PSC personnel and review of various purchase orders 

(POs) indicated that PSC had purchased numerous snubbers from two companies, WISC and VI. The snubbers were originally manufactured by PSC and sold to 

licensees who, in turn, sold them on the surplus market to material brokers 

such as WISC and VI. A review of PSC’s ASL, Revision 3, dated August 17, 

1995, did not identify either company as an approved supplier as required by 

SOP 07.104. The QA manager indicated that neither company had ever been on 

the ASL or had ever been qualified through survey or audit. 

PSC’s failure to place WISC and VI on the ASL, as required by SOP 07.104, was 

identified as a Nonconformance to Criterion V of Appendix B (95-01-02). 

3.6 Oualitv Assurance Orqanization 

The inspectors reviewed PSC’s organization in the current QA manual which 

established that the QA Manager reported directly to the division President. 

Additional quality control personnel reported directly to the Commercial and 

Nuclear Director who was responsible for nuclear production. The QA Manager 

indicated that the functions of the position included administrative and QA 

aspects, pol icy and procedure, internal audits, document control, receiving 

inspection for new material (not the surplus snubbers) and calibration. 

Qual i ty control personnel, who reported to the Commercial and Nuclear 

Director, responsibilities included supplier control, material review board, 

corrective actions, in-process inspection, final inspection, non-destructive 

testing, receiving inspection (for surplus snubbers), change control for 

drawings, contract review, and PO review. 

PSC Quality Assurance Manual Section 0.20, “Management Statement ,’I stated that 

the Manager of Quality Assurance is given the responsibility and authority to 

maintain the Quality Assurance program and its implementation. 

inspectors determined that the arrangement of quality personnel’s 

responsibilities, quality control personnel reporting directly to the 

Commercial and Nuclear Director with little oversight by the QA Manager, could 

effect the QA Manager’s ability to assure that an appropriate quality 

assurance program was established and effectively executed or to verify that 

activities affecting safety-related functions were correctly performed. 

example of this was determined during discussion with the PSC QA Manager, who 

indicated that for a particular period of time, approximately 1992 through 

1994, the purchasing of surplus snubbers, receiving inspection of surplus 

snubbers, disassembly, parts repl acement, reassembly, 1 abel plate repl acement, 

and testing had been solely conducted under the control of production quality 

control personnel without the quality oversight or knowledge of the PSC QA 

Manager. 

However, the 

An 

The inspectors concluded that the activities related to the purchasing, 

processing, and testing of surplus snubbers were activities affecting quality, 

and therefore should have been performed with the quality oversight and 

knowledge of the QA Manager to assure that an appropriate quality assurance 

program had been established and effectively executed and to verify that 

activities affecting safety-related functions were correctly performed. 

was identified as Nonconformance to Criterion I of Appendix B (95-01-03). 

This 
3.7 Purchase Order Review 

The NRC inspectors reviewed several customer POs to PSC for various safetyrelated snubbers. One such order, Pennsylvania Power & Light (PP&L) PO 3- 

50439-1, dated October 11, 1993 (PSC Order No. 0410003), ordered 11-PSA Model 

100 snubbers for general use at the Susquehanna Steam Electric Station. The 

PO required that the documentation package include test reports and a 

certificate of conformance to the PO requirements. The PO also required that 

the items supplied be new and that refurbished and modified items would not be 

accepted without the approval of PP&L prior to shipment. The technical and 

quality requirements required that the units be designed and fabricated in 

accordance with the PP&L approved PSC quality program, which complies to 

Appendix B and Part 21. 

PSC purchased all 11 units from WISC and VI as buyback inventory as follows: 

A total of nine units were purchased from VI under PSC POs 1124040 and 1110250 

and the remaining 2 units were purchased from WlSC under PSC PO 1106560. 

review of the POs to both companies did not identify any quality or technical 

requirements. 

A 

Following receipt by PSC, all snubbers were tested under PSC IT 533, 

"Acceptance Test for 1801119 Shock Arrestor, PSA-100," dated January 29, 1975. 

Paragraph 6.0 of IT 533 identifies the following tests to be performed: 

Activation, Lost Motion and Drag/Breakaway. Design Report (DR) 1319, 

"Mechanical Shock Arrestors Standard Design Specification, " dated April 17, 

1975, requires in Paragraph 8.1 that each production unit shall be subjected 

to acceptance tests to verify functional compliance with applicable drawings 

and specification requirements. 

PSC policy is that all snubbers, regardless of application, are designed and 

manufactured as Class 1 Linear Standard supports and are classified as 

material in accordance with the provisions of ASME 111, Subsection NF-1214, as 

documented in DR 1319. In addition, all snubbers are designed in accordance 

with ASME Section 111, Division 1, Subsection NF, Sub-article 3200 or 3300. A 

review of the documentation package identified that PSC did not complete a 

failure analysis and corrective action report for one of the four snubbers, 

purchased under PO 1110250, which failed the activation test on June 17, 1994. 

The unit was retested satisfactorily on October 28, 1994, however the package 

did not give a reason for the failure. The NRC inspectors verified the 

existence of a Certificate of Compliance, dated October 27, 1994, which 

certified compliance to the PO requirements for snubber serial numbers 2489 

through 2499. The final inspection checklist for the failed snubber (serial 

number 2492) was reviewed and appeared satisfactory. PSC's failure to 

complete the failure analysis and corrective action report for snubber number 

2492 was identified as a Nonconformance to Criterion XVI of Appendix B 

(95-01-04). 

Several other POs were reviewed during the inspection. 

Commonwealth Edison PO 4W5147, dated January 19, 1995, for one Model PSA 35, 

and PP&L PO 20581-1, dated April 10, 1990, for 25-Model PSA 100s. The 

documentation package for these appeared satisfactory. 

They included 

7 

40 
APPEND I X 

PERSONS CONTACTED 

The following persons were contacted during the inspection and except as 

noted, attended both the entrance and exit meetings. 

Pacific Scient i f i c : 

R. Knoblock 

J. Dowdy 

S. Palm 

A. Camacho 

N. Hergenreder 

w. Scott 

L. Wright 

Division President 

Nuclear Sales and Product Manager 

Quality Assurance Manager 

Senior Product Engineer 

Field Services Manager 

Qual i ty Control Engineer 

Quality Assurance Engineer 

U.S. Nuclear Regulatory Commission: 

R. Pettis 

B. Rogers 

Team Leader 

Reactor Engineer 

Did not attended exit meeting 

41 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-oOOl 

October 6, 1995 

Mr. Nicholas J. Liparulo 

Nuclear Safety and Regulatory Activities 

Westinghouse Electric Corporation 

P.O. Box 355 

Pittsburgh, Pennsylvania 15230 

SUBJECT: NRC INSPECTION NO. 99900404/95-02 

Dear Mr. Liparulo: 

This letter addresses the inspection at the Ente per le Nuove Tecnologie, 

L'Energia e L'Ambiewte (ENEA) Valve and Pressurizer Operating Related 

Experiments (VAPORE) test facility in Casaccia, Italy, conducted by Richard P. 

McIntyre of the Nuclear Regulatory Commission's (NRC's) Special Inspection 

Branch, Alan E. Levin of the Reactor Systems Brancn, Juan 3. Peralta of the 

Quality Assurance and Maintenance Branch, Andrzej Drozd of the Containment 

Systems and Severe Accident Branch, and James H. Wilson of the Standardization 

Project Directorate, on July 24 through 26, 1995. The details of the 

inspection were discussed with your staff members during the inspection and at 

the exit meeting on July 26, 1995. 

The purpose of the inspection was to determine if automatic depressurization 

system (ADS) testing activities performed at the VAPORE test facility to 

support design certification of the Westinghouse AP600 advanced reactor design 

were conducted under the appropriate provisions of WCAP-8370, Revision 12A, 

the most recent Westinghouse Quality Assurance Plan (topical report) that has 

been approved by the NRC. 

implemented at YAPORE by ENEA document AP600-GQ9402, "Qual i ty Assurance P1 an 

Description: AP600 Test Program Conducted at the VAPORE Plant in ENEA Cassacia 

(Phase B)," Revision 2. 

The pertinent provisions of WCAP-8370 were 

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. 

The inspection consisted of an examination of 

The results of the inspection indicate that Westinghouse/ENEA, in general , were adequately implementing the AP600 Project quality assurance program 

requirements with the exception of one finding and an unresolved item. 

Specifically, the team identified a Nonconformance with program implementation 

with respect to the generation of facility as-built drawings for AP600 ADS 

Phase B testing at VAPORE. 

concerning the fact that the ENEA quality assurance program does not include 

adequate measures to effectively control the calibratiort status of the 

reference instruments or standards used for instrument calibration. No 

provisions were in place to require re-calibration at the requisite intervals. 

Also, the team identified an Unresolved Item 

42 
N. Liparulo -2- 

The responses requested by this letter and the enclosed Notice of 

Nonconformance 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 Part 2.790 of the NRC's "Rules of Practice," a copy 

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

Room. 

Should you have any questions concerning this inspection, we will be pleased 

to discuss them with you. 

Sincerely, 

n 

Robert lzk-zk- M. Gallo. C ief 

Speci a1 Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Docket No.: 52-003 

Enclosures : 

1. Notice of Nonconformance 

2. Inspection Report No. 99900404/95-02 

cc w/encls: See Next Page 

43 
Mr. Nicholas J. Liparulo 

Westinghouse Electric Corporation 

cc: 

Mr. B.A. McIntyre 

Advanced Plant Safety & Licensing 

Westinghouse Electric Corporation 

Energy Systems Business Unit 

P.O. Box 355 

Pittsburgh, PA 15230 

Mr. John C. Butler 

Advanced Plant Safety & Licensing 

Westinghouse Electric Corporation 

Energy Systems Business Unit 

Box 355 

Pittsburgh, PA 15230 

Mr. M.D. Beaumont 

Nuclear and Advanced Techno1 ogy 

Division GE Nuclear Energy 

Westinghouse Electric Corporation 

One Montrose Metro 

11921 Rockville Pike 

Suite 350 

Rockville, MD 20852 

Mr. Sterling Franks 

U.S. Department of Energy 

Washington, DC 20585 

NE-42 

Mr. S.M. Modro 

EG&G Idaho Inc. 

Post Office Box 1625 

Idaho Falls, ID 83415 

Mr. Frank A. Ross 

U.S. Department of Energy, NE-42 

Office of LWR Safety and Technology 

19901 Germantown Road 

Germantown, MD 20874 

Docket No.: 52-003 

AP600 

Mr. Ronald Simard, Director 

Advanced Reactor Programs 

Nuclear Energy Institute 

1776 Eye Street, N.W. 

Suite 300 

Washington, DC 20006-3706 

STS, Inc. 

ATTN: Lynn Connor 

Suite 610 

3 Metro Center 

Bethesda, MD 20814 

Mr. James E. Quinn, Projects Manager 

LMR and SBWR Programs 

GE Nuclsar Eoergy 

175 Curtner Avenue, M/C 165 

San Jose, CA 95125 

Mr. John E. Leatherman, Manager 

SBWR Design Certification 

GE Nuclear Energy, M/C 81 

San Jose, CA 95125 

Barton Z. Cowan, Esq. 

Eckert Seamans Cherin & 

600 Grant Street 42nd F 

Pittsburgh, PA 15219 

Me1 1 ott 

oor 

Mr. Ed Rodwell , Manager 

PWR Design Certification 

Electric Power Research Institute 

3412 Hillview Avenue 

Palo Alto, CA 94303 

Mr. Charles Thompson, Nuclear Engineer 

U.S. Department of Energy 

Washington, DC 20585 

AP600 Cert i f i cation 

NE-451 

44 
NOTICE OF NONCONFORMANCE 

Westinghouse Electric Corporation 

Pittsburgh, Pennsylvania 

Docket No. : 52-003 

99900404 

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

Westinghouse Electric Corporation's AP600 Automatic Depressurization System (ADS) 

Phase B test program at the Ente Nazionale Energia e Ambiente (ENEA) VAPORE Test 

Facility, conducted on July 24 through July 26, 1995, it was determined that activities supporting Westinghouse Electric Corporation's AP600 design certification 

program, were not conducted in accordance with NRC requirements as identified below: 

A. Section 9.0, "Quality Assurance Requirements" of WCAP-14112, "Automatic 

Depressurization System Test Specification (Phase B1) ," Revision 2, and 

Section 7.0 , "As-Bui 1 t Records" of ENEA document AP600-GQ9402, "Qual i ty 

Assurance Plan Description: AP600 Test Program Conducted at the VAPORE Plant in 

ENEA Cassacia (Phase B)," Revision 2, provide for and require the preparation 

and maintenance of VdPORE test facility as-built drawings which pertain to the 

ADS Phase B tests that characterize +.he features which influence 

thermal-hydraul ic and structural parameters for code Val idation and calculation 

met hodol ogy veri f i cat i on efforts . 

Contrary to the above, as-built drawings, as defined and stipulated in 

WCAP-14112 and AP600-GQ9402, had not been generated for AP600 ADS Phase B 

testing at VAPORE. (95-02-01). 

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

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

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

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

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

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

include for each nonconformance: (1) a description of the steps that were or 

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

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

actions and preventative measures were or will be completed. 

Dated at Rockville, Maryland 

This day of , 1995 

Enclosure 1 

45 
ORGANIZATION : 

REPORT NO.: 

CORRESPONDENCE 

ADDRESS : 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY: 

INSPECTION 

CONDUCTED: 

TEAM LEADER: 

OTHER INSPECTORS: 

REVIEWED: 

APPROVED : 

INSPECTION BASES: 

INSPECTION SCOPE: 

PLANT SITE 

APPLICABILITY: 

Westinghouse Electric Corporation 

Pittsburgh, Pennsylvania 

99900404/95-02 

Mr. Nicholas J. Liparulo 

Nuclear Safety and Regulatory Activities 

Westinghouse Electric Corporation 

P.O. Box 355 

Pittsburgh, Pennsylvania 15230 

John Butler, Principal Engineer 

Advanced P1 ant Safety and Licensing 

(412) 374-5268 

Nuclear steam supply system components and services 

July 24 through 26, 1995 

*jL?$k&&-& 

Richard P. McIntvre 

Vendor Inspection Section (VIS)' 

Special Inspection Branch (PSIB) 

Juan D. Peral ta, HQMB 

Alan E. Levin, SRXB 

Andrzej Drozd, SCSB 

James H. Wilson, PDST 

Cwal ina, Section Chief, VIS -v 

Robert M. Gallo, Chief'; PSIB 

Date 

10 CFR Part 50, Appendix B and 10 CFR Part 21 

To determine if activities performed to support the 

design of AP600 and, specifically, the automatic 

depressurization system (ADS) Phase B test program at 

the Ente Nazionale Energia e Ambiente (ENEA) VAPORE 

Test Facility in Casaccia, Italy, were conducted under 

the appropriate provisions of WCAP-8370, Revision 12A, 

the most recent Westinghouse Quality Assurance Plan 

that has been approved by the NRC. 

None 

Enclosure 2 

46 
1 INSPECTION SUMMARY 

1.1 Nonconformance 

Nonconformance 99900404/95-02-01 was identified and is discussed in 

Section 3.3 of this report. 

1.2 Unresolved Item 

Unresolved Item 99900404/95-02-02 was identified and is discussed in 

Section 3.4 of this report. 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

No previous inspection has been conducted at the VAPORE test facility. 

3 INSPECTION FINDINGS AND OTHER COMMENTS 

3.1 AP600 Quality Assurance Proqram 

Chapter 17 of the AP600 Standard Safety Analysis Report (SSAR) describes the 

Westinghouse Electric Corporation (g) quality assurance (QA) program for the 

design phase of the AP600 Advanced Light Water Reactor (ALWR) Plant Program. 

The QA program is identified as Westinghouse Topical Report WCAP-8370, "Energy 

System Business Unit-Power Generation Business Unit Quality Assurance Plan," 

Revision 12A, dated April 1992. WCAP-8370 applies to all M activities 

affecting quality of items and services, including the design certification 

process for AP600. Accordingly, WCAP-8370 establishes Y's commitments to 

meeting the requirements of 10 CFR 50 Appendix By ASME NQA-1 and NQA-2, and 

Regul atory Guide 1.28, Revi si on 3, "Qual i ty Assurance Program Requirements 

(Design and Construct i on) It 

Section 17.3 of the AP600 SSAR states that activities supporting the dasign 

and design certification phase of the project are performed in accordance with 

M topical report WCAP-8370 as suppl emented by a project-spec1 f i c Qual i ty P1 an. 

WCAP-12600, "AP600 Advanced Light Water Reactor Design - Quality Assurance 

Program Plan (QAPP) ," dated December 1993, the project-specific QA plan, was 

developed by y to enhance WCAP-8370 in specific areas and to establ ish 

additional commitments needed to support the AP600 program. 

that the AP600 design certification test programs and related analyses are 

within the scope of this QAPP. 

It also states 

WCAP-12601, "AP600 Program Operating Procedures," Revision 13, dated July 8, 

1994, was developed by W to establish requirements and responsibilities for 

developing, approving , imp1 ementi ng, revising and maintaining operating 

procedures to meet the QA and administrative requirements of the AP600 

program. WCAP-12601 includes an "AP600 Program Procedure Matrix, " Revision 

15, dated April 4, 1995, which identifies the correlation between the y 

commitments to the QA requirements of (1) ANSI/ASME NQA-1, "Quality Assurance 

Program Requirements for Nuclear Faci 1 i ti es , 1983 Edition (as endorsed by 
Regulatory Guide 1.28, Revision 3) and (2) ANSI/ASME NQA-1 1989 Edition 

through NQA-lb-1991 Addenda, and the corresponding implementing guidance 

embodied in WCAP-9565, "Nuclear and Advanced Technology Division (NATD) 

Quality Assurance Program," Revision 34, dated May 2, 1994, and in WCAP-12601. 

WCAP-9565 governs the implementation of all NATD activities related to areas 

within the scope of WCAP-8370. 

During the inspection, the team assessed the implementation of the applicable 

QA criteria essential to support the AP600 Design Certification application, 

including Design Certification testing. Specifically, the team evaluated the 

effectiveness of the QA program and controls, as described above, including 

the soundness of the data obtained during the AP600 VAPORE ADS test program. 

3.1.1 Ente Nazionale Energia e Ambiente (ENEA) VAPORE Facility QA Program 

In the AP600 design, the automatic depressurization system (ADS) ensures that 

the reactor coolant system (RCS) is depressurized during pertinent transient 

and accident conditions, thereby initiating and maintaining long-term gravity 

injection. Under a technical cooperation agreement, W, ENEA, and ANSALDO 

S.p.A., combined resources to conduct testing at the ENEA VAPORE test facility 

in Cassacia, Italy, with two major objectives: (1) advance knowledge and 

understanding of passive safety system operations, and (2) conduct testing of 

the ADS to provide both design information and data for computer code 

Val idation efforts needed to support design certification. 

During phase A of testing, information on the performance of a prototypic 

sparger was gathered. The ADS phase B tests comprised full-sized simulation 

of one of the two AP600 ADS flowpaths from upstream of the ADS valves to a 

sparger and was intended to simulate and/or conservatively bound the operating 

conditions of the AP600 ADS system configuration. 

WCAP-14112, "Automatic Depressurization System Test Specification (Phase B1) ,'I 

Revision 2, provides that testing, designed to demonstrate overall automatic 

depressurization system (ADS) performance verification, be conducted under a 

QA program that conforms to the requirements of the American Society of 

Mechanical Engineers (ASME) NQA-1, "Quality Assurance Program Requirements for 

Nuclear Facilities, " 1989 Edition through NQA-1-1991 Addenda. 

To this effect, ENEA developed AP600-GQ9402 (AP600 Document No. RCS-T1H-001), 

"Qual i ty Assurance P1 an Description: AP600 Test Program Conducted at the 

VAPORE Plant in ENEA Cassacia (Phase B)," Revision 2 (QAPD), which defines 

responsibilities, prescriptions and recommendations to govern the AP600 ADS 

Test Program/Phase B according to W requirements and pertinent ENEA 

procedures. 

During the inspection, the team reviewed the pertinent documents to determine 

if design certification testing activities associated with the AP600 program 

and performed at the ENEA VAPORE test facility during the ADS test program 

were conducted in accordance with the appropriate provisions of W's 10 CFR 50, 

Appendix By QA program (WCAP-8370). 

activities in specific areas within the scope of AP600-GQ9402, i.e. test 

control, test instrument calibration, facility and records configuration 

The team examined the performance of 

-3- 
control to confirm that activities in these areas were conducted and 

accomplished under suitably controlled conditions by properly trained 

personnel and that the resultant test data collected during such activities 

was appropriately recorded and maintained. 

Based on reviews of these areas, including the QA implementation audit at 

VAPORE conducted by on June 6-9, 1995, and documented in y Audit Report No. 

QLA/ENA0006, dated June 19, 1995, the team concluded that the QA program set 

forth-in AP600-GQ9402, in conjunction with the Y's implementation of the 

pertinent criteria of WCAP-12601, provided sufficient evidence of overall QA 

implementation appropriate to design certification testing, except for certain 

areas as identified below. 

3.2 Test Control 

The team reviewed aspects of test control for the AP600 VAPORE Phase B1 test 

program. These tests involved prototype Stage 1, 2, and 3 automatic 

depressurization system valves in a full-scale piping network, to acquire 

thermal-hydraul ic performance data for Val idation of computer models for 

accident and transient analyses of the AP600 plant. The test program was 

completed in 1994. Documentation reviewed included the ADS test 

specification, test procedures, test logs, test checkl ists, and test results 

as reflected by selected data and Day of Test reports. WCAP-14112, Revision 

2, was developed by u, and the detailed test procedures were developed by ENEA 

and approved by H. 

The test procedures were comprehensive and easy to follow, and included 

checklists for each type of test that was performed in the Phase 61 program. 

The procedures implement the test specification requirements. The test 

procedure included requirements for the development of a test diary as well as 

the checkl ist. Documentation for each test, including completed and signedoff checklists, data plots, Day of Test reports, and other relevant 

information, was contained in a separate divider in the Design Record File 

(DRF). 

engineer who was assigned to the VAPORE faci ity to oversee and participate in 

the conduct of the test program. 

Each test was performed by a team of 2-4 individuals and a 4? resident 

The Day of Test reports were prepared by N's on-site resident engineer for 

transmittal to y Test Engineering in Monroev lle, Pennsylvania, and, as was 

found in the inspection of the SPES-E program at SIET, provide an excellent 

contemporaneous record of important test results, "unexpected events, " 

instrumentation condition and potential problems, and other relevant 

discussion. A list of critical instrumentation was also included as part of 

the test procedures, but functionality was not verified prior to performing 

each test. When the team questioned W and ENEA personnel about verification 

of critical instrumentation, they were informed that the verification was 

performed after the test, rather than before. It was noted that, although the 

primary objective of the testing was the acquisition of thermal-hydraul ic 

data, a substantial amount of mechanical -re1 ated instrumentation was a1 so 

included on the loop, including strain gages and accelerometers. These were 

not considered to be critical instruments for the Phase B1 tests. In general, 

-4- 

49 
functionality of thermal-hydraulic instrumentation (thermocouples, flowmeters, 

pressure transducers) was adequate. 

The test data records were maintained in a well organized manner. 

acquisition system (DAS) software was devel oped according to British standards 

and checked independently by W. The test data was recorded on PROSIG system 

using CD ROM disks and 4? is custodian of all the data. As has been the case 

in other W AP600 test programs performed by outside organizations, the final 

review of the data to determine if test acceptance criteria were met, was 

performed by y personnel in Monroeville. As a result, there is little 

documentation of the actual data review itself, other than the initial 

assessments as contained in the Day of Test reports. 

during the tests requiring disposition are also not recorded in ENEA's DRF. 

The team was informed that this documentation is in Y's files in Monroeville 

and could not be verified during this inspection. Nevertheless, the design 

record file does contain correspondence between 4? and ENEA, specifically 

addressing acceptability of the VAPORE Phase B1 tests. In most cases, tests 

were considered to have met acceptance criteria; however, several tests were 

re-run as a result of W's determination that they had not met acceptance 

criteria. 

to be appropriate and adequate to ensure proper test control. 

The data 

In addition, deviations 

The team considered the documentation of N's review in this manner 

3.3 As-Built Drawinas and Confiquration Control 

WCAP-14112, Section 9.0, "Qual ity Assurance Requirements," and AP600-GQ9402, 

Section 7.0, "As-Built Records," provide for the preparation and maintenance 

of the VAPORE test facility as-built drawings which pertain to the ADS Phase B 

tests that characterize the features which influence thermal-hydraul ic and 

structural parameters. The thermal-hydraul ic parameters are needed to a1 low 

adequate mode11 ing of the facility for code Val idation efforts. 

also provides that specific as-built features to be recorded shall be as 

specified by W. Additional configuration control provisions in AP600-GQg402 

are specified under Section 7.2, "Features to be Recorded and Format," Section 

7.3, "Method and Responsibilities," and 7.4, "Changes". 

AP600-GQ9402 

Modifications to the VAPORE test facility, necessary to support AP600 ADS 

design certification testing, were performed by ANSALDO S.p.A. under contract 

to W. On November 29, 1994, y placed a contract with ANSALDO (MB21177S Change 

Notice) to provide as-built documentation of the ADS test loop at the ENEA's 

VAPORE test facility. 4? stipulated that ANSALDO provide one full set of 

as-built drawings (comprising P&ID, line list of principal flow paths, valve 

list, ADS loop layout drawings, ADS loop isometric drawings, ADS loop 

platform, and ADS loop support drawings) covering both ADS Phases B1 and 62 

configurations. 

records package for AP600 VAPORE Phase B testing. 

During the inspection, however, the team found that as-built drawings, as 

defined and stipulated in WCAP-14112, and in AP600-GQ9402, had not been 

generated for AP600 ADS Phase B testing at VAPORE. This issue was identified 

as Nonconformance 95-02-01. 

y intended to include these drawings as part of the as-built 

50 

-5- 
3.4 Procurement and Cali bration of Test Instrumentation 

WCAP-14112 requires, in part, under Section 9.0, "Quality Assurance 

Requirements," that the following measures be taken in the detailed test 

procedure(s): (1) Provisions for ensuring that calibration of test equipment 

is traceable to recognized national standards, and (2) Verification and 

documentation, to be submitted to W, by the testing organization that the 

instrumentation calibrations have been performed prior to testing. 

6.0, "Instrumentation Management and Control ," of the ENEA QAPD Document, 

AP600-GQ9402, implemented these requirements. 

Section 

During the inspection, the team confirmed that all test instruments used in 

the ENEA VAPORE test facility had been calibrated, prior and after testing, 

using standards or reference instruments traceable to the Servizio di Taratura 

in Italia or, Italian calibration System (SIT). 

of the Western €uropean Calibration Cooperation (WECC), national calibration 

standards equivalent to NIST are established and maintained by SIT. 

In Italy, under the auspices 

During the inspection, the team reviewed the VAPORE test facility calibration 

records which provided evidence of i;*-.ceabi 1 i ty to the appropriate ENEA 

control led SIT-certified standards. This review a1 so provided evidence of the 

adequacy of the facility instrumentation calibration status during each 

testing phase. The team found, however, that the ENEA QA program does not 

include adequate measures to effectively control the calibration status of 

reference instruments or standards used for instrument calibration, as no 

provisions were in place to require re-calibration by SIT at the requisite 

intervals. This may have resulted in the introduction of uncertainties in the 

adequacy of calibration of test facility instrumentation which relied on these 

standards to establish and maintain their accuracy. 

Pending confirmation by W that this lapse in the SIT-certified calibration 

interval for the ENEA standards did not undermine or adversely impact the 

VAPORE ADS test results, this issue will remain unresolved. This issue has 

been i dent i f i ed as Unresol ved 1 tem 95-02-02. 

3.5 Quality Assurance Records 

QA records and documents associated with the W AP600 ADS testing at the VAPORE 

facility were processed in accordance with the provisions in the ENEA QAPD 

(AP600-GQ9402). 

implementation of QAPD requirements. 

QA department, one individual , the designated QA Responsible, has 

responsibility for all QA activities at the facility. 

charged with maintaining all QA records, including, QAPD and audits, 

procedures, test specifications and test matrix modifications, instrument 

calibration records, data acquisition, test results, training and informal 

meetings, ENEA surveys, and nonconformance reports. 

The QAPD lists all of the applicable ENEA procedures for 

Although ENEA does not have a separate 

The QA Responsible is 

ENEA maintains a single "dossier" document file of instrument calibration 

records for each instrument , 2s required by ENEA Document EIHE-94021, 

"Instrument Management." Each dossier contains an instrument card, 

-6- 
calibration certificates and/or calibration control certificates, and any 

existing instruction manuals. 

As part of its contract with 41, ENEA has developed an overall design record 

file that archives all documents associated with AP600 ADS testing at the 

VAPORE facility. This design record file, conforming to the ENEA QAPD, is a 

deliverable from ENEA to 4! and serves as the official record of testing 

activities at VAPORE. 

3.6 Inst ructions. Procedures and Drawi nas 

Procedures and drawings were processed and maintained in accordance with the 

provisions of Section 5.0, "Documentation Management," of the ENEA QAPD 

Document, AP600-GQ9402. The QAPD requires that a copy of the procedures 

applicable to test activities be sent to the QA Responsible for approval and 

distribution to the relevant departments. Drawings and as-built records 

pertaining to the AP600 ADS tests at VAPORE are maintained to document the 

features that influence thermal hydraulic and structural parameters of the 

tests. 

ENEA maintains all testing-related procedures and drawings in the design 

record file that archives all documents associated with AP600 ADS testing at 

the VAPORE facility. 

appropriate procedural controls had been developed and implemented to 

the conduct of AP600 quality-related test activities at VAPORE. 

3.7 Audlts 

Based on the team review, it was determined that 

The ENEA QAPD does not include any requirements for conducting interna 

overn 

audits 

of work performed at the VAPORE facility. 

responsibility under their QA program and conducted a series of readiness 

assessments (in 1993 and 1994) and implementation audits (in 1991 and 1995) of 

the activities associated with the AP600 ADS development. In these 

assessments and audits, 41 treats ENEA as an approved supplier. 

recent 4! audit, WES 95-243, was conducted on June 6-9, 1995, and evaluated 

ENEA's implementation of the activities described under Criterion 7, 

"Procurement," of WCAP-12601. The audit report, QLA/ENA0006, dated June 19, 

1995, conta;ned 7 findings and 2 recommendations and requested a response from 

ENEA by July 20, 1995. The audit report was distributed to ENEA management 

and department heads. In a July 21, 1995, memorandum, the !$ lead auditor 

concluded that the actions taken by ENEA, as described in the July 20, 1995 

response, were acceptable for each of the findings and the recommendations, 

and that the findings were closed. 

However, !$ has assumed this 

The most 

The staff concluded that the W procedures for the conduct of internal audits 

were appropriately followed in the evaluation of QA activities at the VAPORE 

test facility and were effective in evaluating activities associated with 

AP600 ADS testing. 

-7- 
4 PERSONNEL CONTACTED 

Westinahouse Electric CorDoration 

Eugene Piplica, AP600 Test Manager 

David Alsing, AP600 Quality Assurance Manager 

Robin Nydes, AP600 Senior Project Engineer 

Robert Tupper, AP600 Project Engineer 

Tim Bueter, AP600 Test Engineering 

Ente Der le Nuove Tecnoloqie. L’Enerqia e L’Ambiente (EN€ A). 

C. Kropp, Division Head, VAPORE Operating Group 

Piero Incalcaterra, Test Responsible 

Bruno Mi 1 ana, Qual i ty Assurance Responsible 

L. Solaro, Instrumentation Responsible 

G. Serafini, Maintenance Mechanical Responsible 

-8- 

53 
Ident i f i er 

Selected Generic Correspondence on the Adequacy of 

Vendor Audits and the Quality of Vendor Products 

Information Notice 95-45 

- Title 

American Power Service Falsification of American 

Society for Nondestructive Testing (ANST) 

Cert i f i cates 
CORRESPONDENCE RELATED TO VENDOR ISSUES 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 205!%-0001 

December 5, 1995 

Mr. David Z. Hathcock, Quality Assurance Manager 

Cardinal Industri a1 Products , Limited Partnership 

3873 West Oquendo 

Las Vegas, NV 89118 

SUBJECT: CARDINAL'S AUGUST 30, 1995 REPLY TO NOTICE OF NONCONFORMANCE 

Dear Mr. Hathcock: 

99901076/94-01-03 (NRC INSPECTION REPORT NO. 999001076/94-01) 

Thank you for your letter of August 30, 1995, which supplemented the 

information contained in your May 30, 1995, and January 30, 1995, letters in 

response to Notice of Nonconformance 99901076/94-01-03, 

We have reviewed the sampling rationale discussed your letter and noted that, 

while this approach can be expected to improve the overall assurance of 

product integrity, it places heavy reliance on visual and dimensional 

inspection to support the verification of lot homogeneity. We agree that 100% 

visual inspection can be a significant factor in the identification of mixed 

product lots by noting non-uniform markings, coloration differences, or other 

visual discontinuities. However, visual inspection can not assure that all 

items in the same product lot were manufactured from the same heat of material 

or were heat treated under the same conditions. 

With respect to your reference to the use of ASTM A-325 shipping lot sampling plan, please review the comments contained in our April, 24, 1995 letter to 

Mr. Scott Akers, Jr. 

sap?ling is used as the basis for qualifying an item for unrestricted safety 

application in a nuclear power plant, the NRC staff has generally accepted a 

confidence level of 90-95% that no more than 5-10% of the sampled items are 

nonconforming. The sample size and rationale described in your August 30, 

1995 letter does not appear to provide this level of confidence for random 

lots (unverified traceability) of material , especially for the verification of 

critical characteristics related to the physical properties of material. 

Recognizing the need for more specific guidance in this area, the NRC staff 

has initiated a generic study of industry practices and existing guidance 

related to material sampling. This study, when completed, may result in NRC's 

endorsement of specific sampl ing approaches. 

The latest revision of 10 CFR Part 21 (Federal Resister 48369, September 19, 

1995 - enclosed) defines dedication as "an acceptance process undertaken to 

provide reasonable assurance that a commercial grade item to be used as a 

basic component will perform its intended safety function and, in this 

respect, is deemed equivalent to an item designed and manufactured under a 

10 CFR 50, Appendix B, quality assurance program." The regulation further 

states that reasonable assurance is achieved by identifying the critical 

In this letter, we also stated that, when product 

56 
Mr. David Z. Hathcock -2- 

characteristics of the item and verifying their acceptability by inspections, 

tests, or analyses. Critical characteristics of an item are defined as "those 

important design, material, and performance characteristics of a commercial 

grade item that, once verified, will provide reasonable assurance that the 

item will perform its intended safety function." 

Determining the intended safety function of the item being dedicated is an 

integral part of the dedication process. The item's intended safety function 

is considered in both the identification and verification of critical 

characteristics, including the selection of an appropriate sampling plan. The 

safety function of the item is typically controlled by the end user of the 

item (licensee). If the licensee elects to procure such items as basic 

components, the licensee must invoke 10 CFR 50, Appendix B in the procurement 

documents. The supplier has the responsibility of assuring that it is 

supplying either an item which has been designed/manufactured under the 

applicable provisions of Part 50, Appendix B or of dedicating a commercial 

grade item under its own Part 50, Appendix B program. In either case, the 

licensee is responsible for reviewing and approving the suppliers quality 

assurance program controlling these activities. 

Appendix B equivalency when dedicating items such as material, a supplier must 

consider all of the applicable specification requirements to be critical and 

verify that these specification requirements have been met using a sampling 

plan with a confidence level such as discussed above since, typically, the 

supplier does not know the end application (safety function) of the item. 

In order to achieve Part 50, 

The licensee may also opt to procure an item which has been subjected to 

specified inspections and tests by the supplier, but is being sold by the 

supplier as a commercial grade item. Under these conditions, the licensee is 

ultimately responsible for the item's dedication which would include 

identification of the critical characteristics of the item and verifying their 

acceptability. If the inspections and tests performed by the supplier are to 

be ccnsidered a part of the dedication process, the i,spections and tests must 

be controlled in accordance with a quality assurance program that is 

consistent with the applicable portions of Part 50, Appendix B that has been 

audited or surveyed and approved by the 1 icensee. The procured i tern would be 

considered a commercial grade item until completion of the dedication process 

by the licensee. 

would certify that the item has been designed/manufactured in accordance with 

its 1 i censee-approved qual i ty assurance program but would not certify that the 

item complies with Part 50, Appendix 6. Considering that a significant number 

of items supplied to nuclear plants may have a specialized or limited safety 

application, suppliers with quality assurance programs that do not demonstrate 

specification compliance with a confidence level as discussed above may, since 

the recent revision of Part 21, consider supplying such items as commercial 

grade under their quality assurance program. This option would need to be 

negotiated, of course, with the end user of the item (licensee). The licensee 

as part of its dedication of the item, would determine the acceptability of 

sampling procedures and the need for additional inspections or tests based on 

the safety related application of the item. 

When supplying items under these conditions, the supplier 
Mr. David Z. Hathcock -3- 

In accordance with Section 2.790 of 10 CFR, a copy of this letter will be 

placed in the NRC's Public Document Room. 

No response to this letter is required. 

Sincerely, 

Robert M. Gallo, Chief 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Docket No.: 99901076 

58 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-0301 

December 4, 1995 

Mr. Mark T. Capallo, President 

Energy & Process Corporation 

2146-8 Flintstone Drive 

Tucker, GA 30084-5000 

SUBJECT: YOUR AUGUST 25, 1995 RESPONSE TO NOTICE OF NONCONFORMANCE 

(NRC INSPECTION REPORT NO. 99900866/95-01) 

Dear Mr. Capallo: 

Thank you for your letter of August 25, 1995, in response to our June 14, 

1995, letter and the enclosed Notice of Nonconformance. 

We have reviewed your letter and find the proposed corrective actions for 

nonconformances 99900866/95-01-01 and 99900866/95-01-03 satisfactory. 

review the implementation of these actions during a future inspection. 

We will 

We find that your response to nonconformance 99900866/95-01-02 does not 

provide an adequate basis for substantiating that your material sampling plans 

for verifying critical characteristics provide reasonable assurance that the 

dedicated items meet all of the applicable specification requirements. 

When sampling is used to qualify items for unrestricted safety applications in 

nuclear plants, the NRC staff generally expects such sampling to provide a 

confidence level of 90-95% that no more than 5-10% of the items sampled are 

nonconforming. Your sampl ing program does not provide a confidence level 

consistent with these expectations. 

With respect to your comments concerning the lack of established guidelines 

for sampling of commercial grade items, the NRC staff has initiated a generic 

study of industry practices and existing guidance related to material 

sampling. 

specific sampl ing approaches. 

This study, when completed, may result in NRC's endorsement of 

The latest revision of 10 CFR Part 21 (Federal Reqister 48369, September 19, 

1995 - enclosed) defines dedication as "an acceptance process undertaken to 

provide reasonable assurance that a commercial grade item to be used as a 

basic component will perform its intended safety function and, in this 

respect, is deemed equivalent to an item designed and manufactured under a 

10 CFR 50, Appendix By quality assurance program." The regulation further 

states that reasonable assurance is achieved by identifying the critical 

characteristics of the item and verifying their acceptability by inspections, 

tests, or analyses. Critical characteristics of an item are defined as "those 

important design, material , and performance characteristics of a commercial 

grade item that, once verified, will provide reasonable assurance that the 

item will ptrform its intendr-' safety function." 

59 
Mr. M.T. Capallo -2- 

I 

Determining the intended safety function of the item being dedicated is an 

integral part of the dedication process. 

is considered in both the identification and verification of critical 

characteristics, including the selection of an appropriate sampling plan. The 

safety function of the item is typically controlled by the end user of the 

item (licensee). If the licensee elects to procure such items as basic 

components, the licensee must invoke 10 CFR 50, Appendix B in the procurement 

documents. 

supplying either an item which has been designed/manufactured under the 

applicable provisions of Part 50, Appendix B or of dedicating a commercial 

grade item under its own Part 50, Appendix B program. In either case, the 

licensee is responsible for reviewing and approving the suppliers quality 

assurance program controlling these activities. 

Appendix B equivalency when dedicating items such as material, a supplier must 

consider all of the applicable specification requirements to be critical and 

verify that these specification requirements have been met using a sampling 

plan with a confidence level such as discussed above since, typically, the 

supplier does not know the end application (safety function) of the item. 

The item’s intended safety function 

The supplier has the responsibility of assuring that it is 

In order to achieve Part 50, 

The licensee may also opt to procure an item which has been subjected to 

specified inspections and tests by the supplier, but is being sold by the 

supplier as a commercial grade item. Under these conditions, the licensee is 

ultimately responsible for the item’s dedication which would include 

identification of the critical characteristics of the item and verifying their 

acceptability. If the inspections and tests performed by the supplier are to 

be considered a part of the dedication process, the inspections and tests must 

be controlled in accordance with a quality assurance program that is 

consistent with the applicable portions of Part 50, Appendix B that has been 

audited or surveyed and approved by the licensee. The procured item would be 

considered a commercial grade item until completion of the dedication process 

by the licensee. When supplying items under these conditions, the supplier 

would certify that the item has been designed/manufactured in accordance with 

its 1 icensee-approved quality assurance program but w-rlld not certify that the 

iten complies with Part 50, Appendix B. Considering that a significant number 

of items supplied to nuclear plants may have a specialized or limited safety 

application, suppliers with quality assurance programs that do not demonstrate 

specification compliance with a confidence level as discussed above may, since 

the recent revision of Part 21, consider supplying such items as commercial 

grade under their quality assurance program. This option would need to be 

negotiated, of course, with the end user of the item (licensee). The licensee 

as part of its dedication of the item, would determine the acceptability of 

sampling procedures and the need for additional inspections or tests based on 

the safety related application of the item. 

60 
Mr. M.T. Capallo -3- 

In accordance with Section 2.790 of 10 CFR, a copy of this letter will be 

placed in the NRC’s Public Document Room. 

No response to this letter is required. 

Sincerely, 

n 

Robert M. Gallo, Chief 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nucl ear Reactor Regul at i on 

Docket No.: 99900866 

Enclosure: Federal Register 

61 
' +**** ' 

Mr. William B 

Mackson, Inc. 

2346 Southway 

Rock Hill, SC 

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20-1 

ac kwe 

Drive 

29730 

December 5, 1995 

1, President 

SUBJECT: MACKSON, INC's AUGUST 24, 1995 REPLY TO NOTICE OF NONCONFORMANCE 

99901179/95-01-01 (NRC INSPECTION REPORT NO. 99901179/95-01) 

Dear Mr, B1 ackwel 1 : 

Thank you for your letter of August 24, 1995, which supplemented the 

information contained in your April 75. 1995, letter in response to Notice of 

Nonconformance 99901179/95-01-01. 

We have reviewed the additional information provided in your letter and agree 

that the increased sampling level, combined with the other qualitative 

factors, together with augmented visual inspection, should increase the level 

of confidence when the verification of critical characteristics is 

accomplished. These changes represent a significant improvement to your 

commercial grade dedication program. 

However, as stated in our letter of July 19, 1995, when qualifying an item for 

unrestricted safety application in a nuclear plant, the NRC staff would 

generally expect sampling with a confidence level of 90-95% that no more than 

5-10% of the items sampled are nonconforming. Your revised program does not 

provide a confidence level consistent with these expectations for destructive 

testing of commercial grade material obtained from unqualified suppliers or 

distributors. Recognizing the need for more specific guidance in this area, 

the NRC staff, as discussed in our July 19, 1995 letter, has initiated a 

gene,ic study of industry practices and existing guidance related to material 

sampling. 

specific sampling approaches. 

This study, when completed, may result in NRC's endorsement of 

The latest revision of 10 CFR Part 21 (Federal Reqister 48369, September 19, 

1995 - enclosed) defines dedication as "an acceptance process undertaken to 

provide reasonable assurance that a commercial grade item to be used as a 

basic component will perform its intended safety function and, in this 

respect, is deemed equivalent to an item designed and manufactured under a 

10 CFR 50, Appendix By quality assurance program." The regulation further 

states that reasonable assurance is achieved by identifying the critical 

characteristics of the item and verifying their acceptability by inspections, 

tests, or analyses. Critical characteristics of an item are defined as "those 

important design, material, and performance characteristics of a commercial 

grade item that, once verified, will provide reasonable assurance that the 

item will perform its intended safety function." 

62 
Mr. Wi 11 i am B1 ackwell -2- 

Determining the intended safety function of the item being dedicated is an 

integral part of the dedication process. The item’s intended safety function 

is considered in both the identification and verification of critical 

characteristics, including the selection of an appropriate sampling plan. The 

safety function of the item is typically controlled by the end user of the 

item (licensee). If the licensee elects to procure such items as basic 

components, the licensee must invoke 10 CFR 50, Appendix B in the procurement 

documents. The supplier has the responsibility of assuring that it is 

supplying either an item which has been designed/manufactured under the 

applicable provisions of Part 50, Appendix B or of dedicating a commercial 

grade item under its own Part 50, Appendix B program. In either case, the 

1 icensee i s responsible for reviewing and approving the suppl i ers qual i ty 

assurance program controlling these activities. 

Appendix B equivalency when dedicating i tems such as material, a suppl ier must 

consider all of the applicable specification requirements to be critical and 

verify that these specification requirements have been met using a sampling 

plan with a confidence level such as discussed abdve since, typically, the 

supplier does not know the end application (safety function) of the item. 

In order to achieve Part 50, 

The licensee may also opt to procure an item which has been subjected to 

specified inspections and tests by the supplier, but is being sold by the 

supplier as a commercial grade item. Under these conditions, the licensee is 

ultimately responsible for the item’s dedication which would include 

identification of the critical characteristics of the item and verifying their 

acceptability. If the inspections and tests performed by the supplier are to 

be considered a part of the dedication process, the inspections and tests must 

be controlled in accordance with a quality assurance program that is 

consistent with the applicable portions of Part 50, Appendix B that has been 

audited or surveyed and approved by the licensee. The procured item would be 

considered a commercial grade item until completion of the dedication process 

by the 1 icensee. 

would certify that the item has been designed/manufactured in accordance with 

its iicensee-appruved quality assurance program but would not certify that the 

item complies with Part 50, Appendix B. Considering that a significant number 

of items supplied to nuclear plants may have a specialized or limited safety 

application, suppliers with quality assurance programs that do not demonstrate 

specification compliance with a confidence level as discussed above may, since 

the recent revision of Part 21, consider supplying such items as commercial 

grade under their quality assurance program. This option would need to be 

negotiated, of course, with the end user of the item (licensee). The licensee 

as part of its dedication of the item, would determine the acceptability of 

sampling procedures and the need for additional inspections or tests based on 

the safety related application of the item. 

When supplying items under these conditions, the supplier 
Mr. William Blackwell -3- 

In accordance with Section 2.790 of 10 CFR, a copy of this letter will be 

placed in the NRC’s Pub1 ic Document Room. 

No response to this letter is required. 

Sincerely, 

Robert M. Gallo, Chief 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Docket No.: 99901179 

64 
IRC FORM 335 US. NUCLEAR REGULATORY COMMISSION 

2-89) 

IRCM 1102, 

201.3202 BIBLIOGRAPHIC DATA SHEET 

/See instructions on the reverse) 

'. TITLE AND SUBTITLE 

Licensee Contractor and Vendor Inspection 

Status Report 

Quarterly Report 

October - December 1995 

8. AUTHOR61 

1. REPORT NUMBER 

(Aulgnnsd by NRC. Add Vol., SUPP.. Rev., and Addendum Numbers, if any.) 

NUREG-0040 

Vol. 19, No. 4 

3. DATE REPORT PUBLISHED 

MONTH YEAR 

I 

February 1996 

4. FIN OR GRANT NUMBER 

6. TYPE OF REPORT 

Quarterly 

October - December 1995 

7. PERIOD COVERED (Inclusive Dates) 

I, PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC.provide Division, OfficeorRegion, U.S. Nuclear Regulatory Commission. andmai1ingaddress;if contractor, provide 

name and mailing address.) 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

U.S. Nuclear Regulatory Codssion 

Washington, D.C. 20555-0001 

I. SPONSORING ORGANIZATION - NAME AND ADDRESS {If NRC. type "same asabove'Jifcontracror.provide NRC Division, Office or Region, U.S. Nuclear Regulatory Commission, 

and mailing addred 

Same as above 

0. SUPPLEMENTARY NOTES 

1. ABSTRACT 1200 words or led 

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 1995. 

2. KEY WORDSIDESCRIPTORS (List words orphrases that willarsistresearchers in locating thereport.1 

Vendor Inspection 

RC FORM 335 (2-891 

13. AVAILABILITY STATEMENT 

Unlimited 

14. SECURITY CLASSIFICATION 

(This Pagel 

Unclassified 

Unclassified 

(This Report1 

15. NUMBER OF PAGES 

16. PRICE