NUREG-0040 

Vol. 20, No. 1 

Licensee Contractor 

and Vendor Inspection RECEWED 

Status Report 

Quarterly Report 

January - March 1996 

U.S. Nuclear Regulatory Commission 

Office of Nuclear Reactor Regulation 

M 
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2. 

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NUREG-0040 

Vol. 20, No. 1 

Licensee Contractor 

and Vendor Inspection 

Status Report 

Quarterly Report 

January - March 1996 

Manuscript Completed: May 1996 

Date Published: May 1996 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 
DISCLAIMER 

Portions of this document may be illegible 

in electronic image products. Imagec; are 

produced from the best available original 

document. 
DISCLAIMER 

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

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

any of their employees, make any warran@, expms or implied, or assumes any legal liability or responsibility for the acuuacy, completeness, or usefulness of any infommtion, apparatus, product, or process disdosed, or represents that its use would not infringe privately 

owned rights. Reference herein to any specific commercial product, process, or service by 

trade name, trademark, manufacturer, or otherwise does I@ necessarily constitute or 

imply its endorsement, recommedab 'on, or favoring by the United States Government or 

any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof. 
ABSTRACT 

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

Special Inspection Branch, Vendor Inspection Section, that have been 

distributed to the inspected organizations during the period from January 1996 

through March 1996. 

iii 

CONTENTS 

- PAGE 

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

In troduct ion ............................................................. vii 

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

Amer Industrial Technologies, Inc. (99901292/96-01) .............. 2 

Wilmington, DE 

Asea Brown Boveri (99900538/95-02) .............. 37 

/Combustion Engineering, Inc. 

Windsor, CT 

Brand Fire Protection Services, Inc. (99901020/95-01) .............. 52 

Addison, IL 

Modumend, Inc. 

Newbury Park, CA 

(99901291/95-01) .............. 60 

Rochester Gas & Electric Corporation (50-244/96-201) ............... 73 

Rochester, NY 

V 

Select Generic Correspondence on the Adequacy of Vendor .................. 82 

Audits and the Quality of Vendor Products 

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 Federa7 

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

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

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

industry within the framework of quality verification programs. 

The Federal government and nuclear industry have established a system 

The licensee is responsible for developing and maintaining a detailed quality 

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

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

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

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

codes, and standards. 

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

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

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

laboratories performing equipment qualification tests, and holders of NRC 

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

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

coordi nat i on between 1 i censees and vendors. 

These 

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

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

and review of equipment problems found during operation and their corrective 

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

the inspected organization is required to take appropriate corrective action 

and to institute preventive measures to preclude recurrence. 

implications are found, NRC ensures that affected licensees are informed 

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

not ices and bull et i ns . 

When generic 

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

during the calendar quarter for which it is published. 

report lists the nuclear facilities inspected. This information will also 

alert affected regional offices to any significant problem areas that may 

require special attention. Appendices list selected bulletins, generic 

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

correspondence involving vendor issues. 

Each vendor inspection 

viii 
INSPECTION REPORTS 

1 
. ***+* ' 

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-0001 

March 21, 1996 

Mr. Ahmad E. Amer, President, 

Amer Industrial Technologies, Inc. 

1000 south Madison Street 

Wilmington, DE 19801 

SUBJECT: NRC INSPECTION NO. 99901292/96-01 

Dear Mr. Amer: 

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

of your facility at Wilmington, Delaware, conducted by Messrs. U. Potapovs, 

R. McIntyre and S. Matthews of this office on January 29 through February 2, 

1996, and the discussions of their findings with you and members of your staff 

at the conclusion of the inspection. The inspection was conducted to evaluate 

your quality assurance program and its implementation in selected areas 

including (1) design control, (2) control of the manufacturing process, 

procurement and upgrading of stock material, (3) internal and external audits, 

and (4) indoctrination and training of personnel. Your program for compliance 

with the requirements of Part 21 of Title 10 of the Code of Federal 

Requlations (10 CFR Part 21) was also reviewed during this inspection. 

The inspection was accomplished through objective evaluation of selected 

procedures and records, discussions, and observations of ongoing activities by 

the inspectors. The specific areas examined during the NRC inspection and the 

findings are discussed in the enclosed inspection report. 

Based on the results of this inspection, we determined that certain of your 

activities appeared to be in violation of NRC requirements. 

(1) the document describing your policy for compliance with 10 CFR Part 21 did 

not provide for evaluating deviations and reporting defects in accordance with 

10 CFR 21.21(a), and (2) the evaluations conducted to determine if a 

reportable condition existed were inadequate. 

Specifically, 

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

specified in the enclosed Notice of Violation when preparing your response. 

In your response you should document the specific actions taken and any 

additional actions you plan to prevent recurrence. 

Additionally, during this inspection, we identified several instances where 

the implementation of your quality assurance proc,ram failed to assure that the 

products supplied by your company complied with the applicable technical or 

quality requirements specified in the applicable procurement documents. The 

specific findings and references to the pertinent requirements are identified 

in the enclosures to this letter. 

Please provide us within 30 days of this letter a written statement in 

accordance with the instructions specified in the enclosed Notice of 

2 
A.E. her -2- 

Nonconformance. 

good cause for us to do so. 

We will consider extending the response time if you can show 

The response requested by this letter and the enclosed Notices are not subject 

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

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

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

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

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

discuss them with you. 

Si ncerel y , 

&egory@.htfal ina, Acting Chief 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Docket No.: 99901292 

Enclosures: 1. Notice of Violation 

2. Notice of Nonconformance 

3. Inspection Report 99901292/96-01 

3 
NOTICE OF VIOLATION 

Amer Industrial Technologies, Inc. 

Wilmington, Delaware 

Docket No. : 99901292 

Report No.: 96-01 

During an NRC inspection conducted at your Wilmington, Delaware facility on 

January 29 through February 2, 1996, a violation of NRC requirements was 

identified. In accordance with the "General Statement of Policy and Procedure 

for NRC Enforcement Actions," NUREG-1600, the violation is listed below: 

10 CFR 21.21(a) requires, in part, that each corporation subject to the 

regul at ions shall adopt appropri ate procedures to ensure the eval uation of 

deviations within 60 days of discovery, the submittal to the NRC of an interim 

report if the evaluation cannot be completed within 60 days, and the reporting 

to a responsible official of a defect or a failure to comply related to a 

substantial safety hazard within 5 working days of completing the evaluation. 

Contrary to the above, (1) AMER I;: ' IC' pial Technologies, Inc. (AIT) document, 

"Procedures for Compliance With 10 CFR Part 21," dated April 27, 1993, which 

described the AIT policy for compliance with 10 CFR Part 21 did not provide 

for evaluation and reporting in accordance with 10 CFR 21.21(a), and (2) AIT 

performed inadequate and incomplete evaluations to determine if a defect or 

failure to comply associated with a substantial safety hazard existed. The 

evaluations consisted of three Interoffice Memos from the QA Manager to the 

AIT President that stated that none of the findings from the December 5-7, 

1994, and the June 26-28, 1995, ASME surveys contained issues that should have 

been reported to the NRC per 10 CFR Part 2:. (99901219/96-01-01) 

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

Pursuant to the provisions of 10 CFR 2.201, AMER Industrial Technologies, Inc. 

is hereby required to submit a written statement or explanation to the U.S. 

Nuclear Regulatory Comnission, ATTN: Document Control Desk, Washington, D.C. 

20555 with a copy to the Chief, Special Inspection Branch, Division of 

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

date of the letter transmitting this Notice of Violation (Notice). This reply 

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

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

contested, the basis for disputing the violation, (2) the corrective steps 

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

will be taken to avoid further violations, and (4) the date when full 

compl i ance wi 11 be achieved. 

docketed correspondence, if the correspondence adequately addresses the 

required response. 

extending the response time. 

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

shall be submitted under oath or affirmation. 

Your response may reference or i ncl ude previous 

Where good cause is shown, consideration will be given to 

Enclosure 1 

4 
Because your response wi 11 be placed in the NRC Pub1 ic Document Room (PDR) , to 

the extent possible, it should not include any personal privacy, proprietary, 

or safeguards information so that it can be placed in the PDR without 

redaction. 

should clearly indicate the specific information that you desire not to be 

placed in the PDR, and provide the legal basis to support your request for 

withholding the information from the public. 

However, if you find it necessary to include such information, you 

Dated at Roc kv i 1 1 e, Mary1 and 

this &+day of March 1996 
NOTICE OF NONCONFORMANCE 

MER Industrial Technologies, Inc. 

Wilmington, Del aware 

Docket No.: 99901292 

Report No.: 96-01 

Based on the results of an NRC inspection conducted on January 29 through 

February 2, 1996, it appears that certairi of your activities were not 

conducted in accordance with NRC requirements. 

A. Criterion 111, "Design Control" of Appendix B to Part 50 of Title 10 of 

Code of Federal Requlations, (10 CFR Part 50) requires that measures 

shall be established to assure that applicable requirements are 

correctly translated into specifications, drawings, and instructions. 

Criterion I11 also requires the establishment of interfaces between 

participating design organizations for the review, approval , and 

revision of design documents as well as for checking the adequacy of 

design. 

Paragraph NCA 3260(a) of Section I11 of the ASME Code states that the 

Design Report which the Certificate Holder or Designer provides, shall 

be reviewed by the Owner or his designee. 

Paragraph NCA 3554 of Section I11 of the ASME Code states that any 

modification of any document used for construction, from the 

corresponding document used for the design analysis, shall be reconciled 

with the design report. 

Paragraph ND 3362 of Section I11 of the ASME Code states that flanges 

designed to standards other than B 16.5 are acceptable provided they 

have been designed in accordance with the rules of ASME Code, Section 

111, Appendix XI. 

1. Contrary to the above, the Dcsign Report for Job 392 did not contain 

documentation of the Owner's review. The report also did not 

include the latest revisions of the construction drawings and, 

therefore, did not accurately reconcile the design changes with the 

design report. 

2. Contrary to the above, her Industrial Technologies, Inc. (AIT) 

dispositioned Nonconformance Report (NCR) 392-1 "use as is" without 

demonstrating that the design would meet the applicable ASME Code 

requirements. (99901292/96-01-02) 

B. Criterion VI I "Control of Purchased Materi a1 , Equipment , and Services" 

of 10 CFR Part 50, Appendix B states, in part, "Measures shall be 

established to assure that purchased material , equipment, and services, 

whether purchased directly or through contractors and subcontractors, 

conform to the procurement documents." 

Enclosure 2 

6 
Criterion I11 "Design Control" of 10 CFR Part 50, Appendix B states, in 

part, "Measures shall be established for the selection and review for 

suitability of application of materials, parts, equipment, and processes 

that are essential to the safety-related functions of the structures, 

systems, and components. " 

Paragraph NCA 3867.4 of Section I11 ot the ASME Code states that an ASME 

Certificate Holder who elects to upgrade unqualified stock material may 

accept certification of the requirements of the material specification 

which must be performed during melting and of the heat analysis, 

providing that the Certificate Holder performs (or subcontracts) all 

other requirements of the material specification on each piece of the 

stock material. 

1. 

2. 

3. 

4. 

5. 

Contrary to the above, AIT elected to upgrade stock material for Job 

392 inlet and outlet pipe nozzles but failed to perform all testing 

required by the applicable material specification (SA 106, Grade B). 

Specifically, no documentation was available to indicate that 

flattening test and hydrostatic test were performed on this 

materi a1 . 

Contrary to the above, AIT elected to upgrade stock material for 

SA-249 heat exchanger tubing for Job 442 but failed to provide 

sufficient documentation to demonstrate that all of the testing 

required by the material specification was performed on each of the 

36 tubes purchased from an unqual i f i ed suppl i er. 

Contrary to the above, AIT elected to upgrade explosively clad 

SA 516 Grade 70 heat exchanger tube sheets for Job 331 but failed to 

demonstrate that this material conformed with the applicable 

speci f i cat i on requi rement s . showed Charpy v lateral expansion lower than permitted by paragraph 

NC 2330 of Section I11 of the ASME Code or by the AIT purchase 

specification for this material. 

documentation regarding the disposition of the nonconforming 

condition. 

Speci f i call y , 1 aboratory test resul t s 

The file contained no 

Contrary to the above, AIT elected to upgrade SA 516, Grade 70 plate 

material for Job 331, but failed to provide sufficient documentation 

to demonstrate that the required testing had been performed on each 

piece of the stock material. Specifically, three separate pieces of 

this plate were identified with the same heat code number (M-2501) 

and documentation in the Job file showed only one sample with this 

heat code sent for laboratory testing. 

Contrary to the above, AIT elected to upgrade SA 516, Grade 70 plate 

material for fuel oil filter body and for inlet and outlet slip-on 

flanges (both items for Job 392) but failed to perform all testing 

required by the material specification and to provide sufficient 

information to the test laboratory to assbre that the testing would 

be performed to the specification requirements. (99901292/96-01-03) 

2 
C. 

D. 

Criterion IX "Control of Special Processes" of 10 CFR Part 50, Appendix 

B states: "Measures shall be established to assure that special 

processes, including welding, heat treating, and nondestructive testing, 

are control 1 ed and accompl i shed by qual if i ed personnel using qual if i ed 

procedures in accordance with applicable codes, standards, 

specificat ions, criteria, and other speci a1 requirements. 'I 

Paragraph ND-4622.7 states, in part, that welds in certain materials are 

exempt from mandatory postweld heat treatment provided that a 200 OF 

minimum heat is maintained during welding. Paragraph ND-2400 states, in 

part, that required test shall be conducted for each heat of bare 

electrodes for use with the gas Tungsten arc welding (GTAW) processes. 

Paragraph 3.1 of Section 3.0, "Welding/Brazing and Fabrication 

Requirements," of Bechtel's design specification required, in part, that 

all welded joints of category D, as defined in paragraph ND-3351, shall 

be in accordance with subparagraph ND-3352.4. ND-3352.4(b), "Full 

Penetration Corner Welded Attachments," required, in part, that nozzles 

shall meet the fabrication requirements of ND-4244(b), "Corner Welded 

Nozzles and Branch Piping Connections," that required, in part, that 

when complete joint penetration cannot be verified by visual examination 

or other means permitted, backing strips or equivalent shall be used 

with full penetration welds deposited from only one side. 

1. Contrary to the above, a minimum preheat temperature of 200 OF was 

not specified in either the welding procedure specification (WPS) WT-713 or PQR 713 and may not have been performed since none of AITs 

records document the actual preheat. Additionally, the weld metal 

qualification test (required by ND-2400 and performed by Amer's 

supplier for its PO 20537, dated October 27, 1993) did not qualify 

the SFA-5.17, EM12K filler metal for use in the GTAW process. 

2. Contrary to the above, the 0.0185-inch root gap provided by AIT's 

nozzle penetraltion machining dimensions specified on AIT Drawing 

392-2, "Body for Oil Filter - Machining Detail - Item lAll and 1A12 

(2 units)," Revision 1, dated December 16, 1993, did not ensure that 

a full penetration weld was achieved. The WPSs specified a root gap 

for grove welds of 1/16-inch to 3/16-inch (0.0625- to 0.1875-inch). 

However, the hole size for the nozzle penetration was specified on 

Drawing 392-2 as 3.535-inches diameter (k 0.002-inch). Given a 

maximum hole size of 3.537-inches diameter and a 3-inch, schedule 40 

pipe with an outside diameter of 3.5-inches, with the pipe nozzle 

inserted in the hole (forming a category D welded joint), the 

resulting maximum root gap would be 0.0185-inch, not the 0.0625-inch 

desired root gap described in the WPS. (99901292/96-01-04) 

Criterion VI1 "Control of Purchased Material, Equipment, and Services'' 

of 10 CFR Part 50, Appendix B states, in part, "Measures shall be 

establ i shed to assure that purchased materi a1 , equipment , and services, 

whether purchased directly or through contractors and subcontractors, 

conform to the procurement documents. These measures shall incl ude 

provisions, as appropriate, for source evaluation and selection, 

8 
E. 

F. 

provisions, as appropriate, for source evaluation and selection, 

objective evidence of quality furnished by the contractor or 

subcontractor . . . 'I 

Criterion IV, "Procurement Document Control" of 10 CFR Part 50, Appendix 

B states, in part, "TO the extent necessary, procurement documents shall 

require contractors or subcontractors to provide a quality assurance 

program consistent with the pertinent provisions of this appendix." 

Bechtel's PO CCDG0767 for 10 filter cartridges (Job 523) invoked the 

quality requirements of ASME Code, Section 111, NCA 4000 for pressure 

retaining parts and American National Standards Institute (ANSI) 

standard N45.2 for other parts determined to be safety related. 

Bechtel 's procurement specification a1 so stated that for safety-re1 ated 

non-Code parts AIT shall either provide a QA supplement to control the 

step-by-step processing of these items, or provide a QA program 

supplement which specifies that AIT's ASME Code QA program shall be used 

to process non-Code parts. The performance requirements for these 

cartridges were specified in Bechtel specification SP-760. 

Contrary to the above, AIT procured the filter cartridges from a 

supplier which had not been audited or otherwise qualified and did not 

verify by either inspections, tests, or analyses that the design, 

material, and performance characteristics of the commercial grade 

cartridges complied with the specification requirements. AIT, without 

any basis, certified that the filter cartridges complied with ASME Code 

Section 111. (99901292/96-01-05} 

Criterion XVII, "Quality Assurance Records" of 10 CFR Part 50, Appendix 

B states, in part, "Sufficient records shall be maintained to furnish 

evidence of activities affecting quality. The records shall include at 

least the following: ... inspections, tests ... I' 

Paragraph NCA 3867.2 of ASME Code, Section I11 states: "All 

characteristics required to be reported by the material specifications 

and by this section shall be verified and the results recorded." 

Paragraph 7.7 of AIT's Qual ity Assurance Manual (QAM) states that "The 

Hydro Test Record will be prepared by the Project Engineer." 

Contrary to the above, AIT could not produce a record of hydrostatic 

testing of 35 tubes for Job 442. AIT did produce a reconstructed 

Hydrostatic Test Record of the 35 tubes, dated after completion of Job 

442, however, the fabrication sequence and signoffs indicated on Job 442 

route sheet do not support the basis of this report. (99901292/96-01-06} 

Criterion V, "Instructions, Procedures, and Drawings" of 10 CFR Part 50, 

Appendix B states, in part that "Activities affecting quality shall be 

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

type appropriate to the circumstances and shdi 1 be accompl ished in 

accordance with these instructions, procedures, or drawings." 

4 

9 
G. 

Paragraph 7.3.2 of AIT's QAM states that "Each Route Sheet shall contain 

the manufacturing, testing, examination and inspections in their proper 

sequence and ... reference applicable procedure by number and revision 

1 eve1 . " 

Paragraph 7.3.4 of AIT's QAM states that "If work is required as a 

result of nonconformity, a revised route sheet shall be issued." 

1. 

2. 

3. 

Contrary to the above, step 14 on the route sheet for Job 442 

specified "Roll tubes on LH and RH tubesheets" but failed to 

identify applicable procedure or any parameters to control the 

rolling operation. 

for this operation and no procedure for this operation was in the 

Job file. 

There were no signatures in the sign-off blocks 

Contrary to the above, step 10 on the route sheet for Job 4102 

specified "Clean/Prepare for shipment" but failed to identify 

applicable cleaning procedure and contained no signatures in the 

sign-off blocks for this operation. 

found in the Job file. The customer specification for this item 

imposed a maximum Chloride limit for the cleaning solution and 

required the cleaning procedure to be available upon request. 

Cleaning procedure was not 

Contrary to the above, substantial work, including welding, was 

performed on Job 331 heat exchanger baffle segments to repair a 

nonconformity without a revised route sheet. 

apparently performed in accordance with a sketch which was attached 

to the nonconformance report. 

heat code number for material to be used or any nondestructive 

examination of the repair weld. (99901292/96-01-07) 

The work was 

This sketch failed to specify the 

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

of Appendix B to 10 CFR Part 50 states, in part, "Measures shall be 

established to assure that purchased material, equipment, and services, 

whether purchased directly or through contractors and subcontractors, 

conform to the procurement documents. These measures shall i ncl ude 

provisions, as appropriate, for source evaluation and selection, 

objective evidence of quality furnished by the contractor or 

subcontractor, inspection at the contractor or subcontractor source and 

examination of products upon delivery." 

Supplement 18S-1, "Supplementary Requirements For Audits," of ASME 

NQA-1-1989 requires in Section 4, "Performance," that objective evidence 

shall be examined to the depth necessary to determine if these elements 

are being implemented effectively. Audit results shall be documented by 

auditing personnel and shall be reviewed by management having 

responsibility for the area audited. 

Section 10.0 "Audits," of AIT's QAM states, in part, in Paragraph 10.2.3 

that elements that have been selected for audit shall be evaluated 

5 
H. 

I. 

against specified requirements. Objective evidence shall be examined to 

the depth necessary to determine if these elements are being implemented 

effectively . 

Contrary to the above, the inspection identified that the reports for 

both internal audits and external vendor evaluations did not provide 

adequate documented objective evidence for the areas reviewed and the 

activities conducted during these audits. 

an adequate definition of the audit scope and contained limited overall 

depth. (99901292/96-01-08) 

The audit reports also lacked 

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

of 10 CFR Part 50, Appendix B states, in part, "Measures shall be 

established to assure that purchased material, equipment, and services, 

whether purchased directly or through contractors and subcontractors, 

conform to the procurement documents. It further states that the 

effectiveness of the control of quality by contractors and 

subcontractors shall be assessed at intervals consistent with the 

importance, complexity, and quantity. 

1. Contrary to the above, no provisions existed in AIT's QAM requiring 

that AIT perform implementation audits or conduct some other 

activity to verify that ASME certificate holders are effectively 

implementing their QA program prior to supplying material for use in 

products to be supplied to nuclear plants by AIT as meeting 10 CFR 

Part 50, Appendix B. 

2. Contrary to the above, the Registered Professional Engineer (RPE) 

who is currently used by AIT for ASME Code design work was not 

listed on the current Approved Vendors List as qualified to 

providing engineering services. 

3. Contrary to the above, AIT procured material and services for 

Job 331 from a vendor (Trinity Industries, Navasota, TX) without 

verifying the effectiveness of the control of quality at the 

location where these services were being performed. (99901292/96-01- 

09 1 

Criterion 11, "Quality Assurance Program" of 10 CFR Part 50, Appendix B 

states, in part, "The program shall provide for indoctrinati'on and 

training of personnel performing activities affecting quality as 

necessary to assure that suitable proficiency is achieved and 

maintained. " 

Criterion XVII, "Quality Assurance Records" of 10 CFR Part 50, Appendix 

8 states, in part, "Sufficient records shall be maintained to furnish 

evidence of activities affecting quality. The records shall include at 

least the following: Operating logs ... and material analyses. The 

records shall also include closely-related data such as qualifications 

of personnel , procedures, and equipment . " 

6 
Section 4.0 "Personnel Training," of the AIT QAM, states, in part, in 

Paragraph 4.3.7 that the QA Manager has the responsibility to maintain 

the indoctrination and training records for employees. 

Contrary to the above, AIT could not provide the indoctrination and 

training records for current AIT employees. It also appeared that AIT 

failed to conduct appropriate training activities of certain personnel 

required as part of the corrective action to several findings from the 

two most recent ASME Surveys, conducted in December 1994 and June 1995. 

(99901292/96-01-10) 

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) the reason for the nonconformance, or if 

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

steps that have been taken and the rp ults 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 atl$ockville, Maryland 

this 21 day of March, 1996 

7 

12 
REPORT NO.: 

ORGANIZATION: 

ORGAN I ZAT IONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY : 

INSPECTION DATES: 

LEAD INSPECTOR: 

OTHER INSPECTORS: 

REVIEWED BY: 

APPROVED BY: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF INSPECTION AND SUPPORT PROGRAMS 

1. : 

9990 1292 /96-01 

Amer Industri a1 Techno1 ogi es , I nc. 

1000 South Madison Street 

Wilmington, DE 19801 

Ahmad E. Amer, President 

Amer Industrial technologies, Inc. manufactures and 

supplies vessels, components, parts, piping 

subassemblies, and component supports to the nuclear 

i cdustry . 

January 29, 1996 through February 02, 1996 

-3-Zl-C1L 

Date 

lL&L {TG;-- Uldis Potapovs 

Vendor Inspection Section (VIS) 

Special Inspection Branch (PSIB) 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regul at i on 

Richard P. McIntyre, VIS/PSIB 

Steven M. Matthews, VIS/PSIB 

Grkyorv C. Cwalina. Chief 

Vendor Inspection Section 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

1 Speci a1 Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Enclsoure 3 

13 
1 $WRY OF INSPECTION FINDINGS 

During this inspection, the NRC inspectors evaluated the implementation of 

her Industrial Technology Inc.'s (AIT) quality assurance (QA) program by 

examining records for safety related components manufactured by AIT and 

shipped to licensed nuclear power plants over the last three years. 

inspectior, focused on procurement and upgrading of stock material, control of 

manufacturing processes, indoctrination and training of personnel, internal 

and external audits, and the implementatinn of Title 10 Code of Federal 

Requlations, Part 21 (Part 21) requirements. 

The 

The inspection basis consisted of the following: 

Appendix By "Quality assurance Criteria for Nuclear Power Plants and 

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

Federal Requl ations (10 CFR Part 50). 

10 CFR Part 21. 

American Society of Mechanical Fngineers (ASME) Boiler and Pressure 

Vessel Code (Code). 

AIT's Quality Assurance Manual (QAM), Rev. 9, dated June 28, 1995. 

1.1 Violations 

Contrary to 10 CFR 21.21, which states that corporations subject to this 

regulation must adopt appropriate procedures ensure evaluation of deviations 

for reportability under this rule, (1) AIT did not include the essential 

requirements of 10 CFR 21.21 in the AIT document titled "Procedures for 

Compliance with 10 CFR Part 21," dated April 27, 1993, and (2) AIT performed 

inadequate and incomplete evaluations to determine if a defect or failure to 

comply associated with a substantial safety hazard existed. 

identified as a Severity Level IV Violation (Supplement VII) and is discussed 

in Section 2.1 of the report. (99901292/96-01-01) 

This issue was 

1.2 Nonconformances 

Nonconformance 99901292/96-01-02 was 

Section 3.4.1.1.1 of this report. 

Nonconformance 99901292/96-01-03 was 

Sections 3.4.1.1.2(3), 3.4.1.3.1, 3.4 

3.4.1.1.2(2) of this report. 

dent 

dent 

1.4. 

fied and is discussed 

fied and is discussed 

(l), 3.4.1.4.1(2), and 

Nonconformance 99901292/96-01-04 was identified and is discussed in 

Sections 3.4.1.1.3 and 3.4.1.1.4 of this report. 

Nonconformance 99901292/96-01-05 was identified and is discussed in 

Section 3.4.1.2 of this report. 

2 

14 
2 

Th i 

3 

3.1 

Nonconformance 99901292/96-01-06 was identified and is discussed in 

Section 3.4.1.3.1 of this report. 

Nonconformance 99901292/96-01-07 was identified and is discussed in 

sections 3.4.1.3.2, 3.4.1.5, and 3.4.1.4.2 of this report. 

Nonconformance 99901292/96-01-08 was identified and is discussed in 

Sections 3.4.4.1 and 3.4.4.2 of this report. 

Nonconformance 99901292/96-01-09 was identified and is discussed in 

Sections 3.4.4.2.and 3.4.1.4.1(3) of this report. 

Nonconformance 99901292/96-01-10 was identified and 

section 3.4.5 of this report. 

STATUS OF PREVIOUS INSPECTION FINDINGS 

was the first NRC inspection at AIT 

INSPECTION FINDINGS AND OBSERVATIONS 

Entrance and Exit Meetinss 

s discussed in 

During the entrance meeting on January 29, 1996, the inspectors discussed the 

inspection scope and logistics and developed general information about AIT’s 

products and activities. A preliminary exit meeting was held on February 2, 

1996 to summarize the inspection findings and observations and to identify 

areas where additional information was needed. AIT provided additional 

information after the preliminary exit meeting in the areas of tube rolling 

process qualification, in-process inspection, material subdividing, and design 

reports. This information was considered by the inspection team and is 

included in the inspection report. The final exit meeting was conducted on 

March 15, 1996, at NRC headquarters in Rockville, Maryland to discuss the 

inspection findings and observations with AIT management. 

3.2 DescriDtion of Facilities 

Up until March 1995, AIT held ASME Certificates of Authorization to 

manufacture ASME Code Section 111, Class 1, 2, and 3 vessels, components, and 

piping subassemblies and to apply the Code symbol stamp to these items. ASME 

allowed the Certificates to expire after two unsuccessful resurveys of AIT’s 

manufacturing activities by the ASME survey team. AIT president indicated 

that they have applied to ASME for re-accreditation and expected a survey to 

be scheduled by April 1996. AIT holds current ASME U, U2, and PP stamps for 

the manufacture of ASME Section VI11 and B31.1 components. 

Manufacturing capability includes material cutting, bending, welding, and 

limited machining. Nondestructive test capability is limited to liquid 

penetrant and magnetic particle inspection and hydrostatic testing. 

analyses, radiography, ultrasonic testing, mechanicai testing, and heat 

treatment are subcontracted. 

Chemical 

3 

15 
At the present time, no work was being done on items destined for domestic 

licensed nuclear power plants. 

such items was shipped during February 1995. 

According to available records, the last of 

3.3 10 CFR Part 21 Proaram Review 

The inspectors reviewed the AIT document, "Procedures for Compliance With 

10 CFR Part 21," dated April 27, 1993, which describes AIT's policy for 

compliance with 10 CFR Part 21. This document was in the form of a memo from 

the company president t o a l l employees requesting them to notify him of any 

deficiencies as described in the memo so that he or his designee could notify 

the customer and the NRC. The memo did not contain the essential elements to 

assure compliance with 10 CFR Part 21. 

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 all cases 

within 60 days of discovery; (2) ensure that if an evaluation of an identified 

deviation or failure to comply potezb !ally 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. 

To review 10 CFR Part 21 implementation, the inspection team requested 

documented evaluations for Part 21 reportability that had been performed to 

date. AIT only had documented records of three instances, in December 1994 

where review for Part 21 reportability had been performed. In these three 

cases, the QA Manager sent a memorandum to the President, at his request, 

stating that he had reviewed the findings from the December 5-7, 1994, and the 

June 26-28, 1995, ASME surveys and none of the ASME findings contain issues 

that should have been reported to the NRC per 10 CFR Part 21. 

Interoffice Memos to the President did not contain any documentation of the 

basis for reaching this conclusion, therefore, the inspectors could not 

determine the extent of the evaluation performed. 

"Procedures for Compliance With 10 CFR Part 21" does not define the evaluation 

process or documentation needed to respond to the Part 21 review request by 

the President. 

These 

The AIT document 

The failure to proceduralize the requirements specified in 10 CFR 21.21(a) and 

the failure to perform an adequate and complete review to determine if a 

defect or failure to comply associated with a substantial safety hazard 

existed constitutes a violation of more than minor significance and is 

identified as a Severity Level IV Violation. (99901292/96-01-01) 

4 
3.4 Qual itv Assurance Proqram ImDlementation 

AIT's quality assurance program is described in their QAM. Revision 9 of the 

QAM, dated June 28, 1995, was in effect at the time of this inspection. 

According to the manual pol icy statement, this manual describes the control 1 ed 

manufacturing system which has been established by management to effectively 

maintain the quality of product design, construction, and installation in 

accordance with the requirements of the ASME Boiler and Pressure Vessel Code, 

Section 111, Division 1, Class 1, 2, 3, MC, and Materials Organization for 

supplying material. AIT does not have a separate QA program for supplying 

safety related non-Code items under the rules of 10 CFR Part 50, Appendix B 

and there are no references in their ASME manual that would indicate its 

applicability to Appendix B items. 

by reviewing records for previously shipped orders and observing work in 

progress. 

Implementation of AIT's QAM was examined 

3.4.1 Review of Fabrication Records 

The NRC inspection team reviewed available fabrication records for safety 

related products shipped to licensed nuclear power plants over the last three 

years to assess the effectiveness of AIT's QA program implementation in 

assuring that these products conformed to customer specifications and 

referenced industry standards. 

stated policy of shipping all permanent records (Code Data Reports, Design 

Reports, Certified Material Test Reports, etc.) to the customer after 

completion of a project. 

files. 

The review was somewhat hampered by AIT's 

Nonpermanent QA records were maintained in project 

3.4.1.1 Job 392 - Fuel Oil Filter 

Bechtel Power Corporation (Bechtel) issued purchase order (PO) CCDGOO65, to 

AIT for two safety-related fuel oil filters for the Calvert Cliffs Nuclear 

Power Plant. The filters were for the fuel oil transfer line between the 

storage tank and the transfer pump installed in the new diesel generator 

buildings at Calvert Cliffs. According to AIT records the fuel oil filters 

were completed and shipped to Calvert Cliffs in February 1994. 

Bechtel ' s PO imposed procurement speci f i cat i on DG-80382, "Procurement 

Specification for Calvert C1 iffs Nuclear Power Plant Diesel Generator 

Project," Revision 1, dated September 21, 1993. The procurement specification 

identified the safety-related fuel oil filters as "basic components" and 

imposed the requirements of 10 CFR Part 21. 

also required that AIT have a QA program that conforms with NCA-4000 of the 

ASME Code for pressure retaining parts, and American National Standards 

Institute (ANSI) standard N45.2 for other parts determined to be safety 

related. The procurement specification stated that for safety-related nonCode parts AIT shall either provide a QA program supplement to control the 

step-by-step processing of these items, or provide a QA program supplement 

which specifies that AIT's ASME Code QA program shall be used to process nonLode parts. 

The procurement specification 

5 

17 
The PO also imposed the requirements of Bechtel's design specification SP-760, 

"Safety-related Fuel Oil Filters," Revision 1, dated September 21, 1993. The 

design specification prescribed that the fuel oil filters be designed, 

constructed, examined, tested, N-stamped, cleaned, and shipped in accordance 

with the ASME Code, 1986 edition, Section 111, "Rules of Construction of 

Nuclear Power Plant Components," Division 1. rlass 3 Components. The design 

specification listed the loading conditions to be considered in the design of 

the fuel oil filters as including, but not limited to, internal pressure and 

seismic accelerations. The specified design pressure was -14.7 psig (vacuum) 

to 15 psig at a design temperature of 15°F to 120°F. 

3.4.1.1.1 Design 

The Design Report for Job 392 was not provided to the team during the course 

of this inspection. However, with its transmittal letter to the NRC dated 

February 13, 1996, AIT sent a copy of the Design Report for job 392 for the 

team's review. The team determined that the Design Report consisted of a 

document titled "Seismic Analysis of the (2) Fuel Oil Transfer Filter-Item 

lAll and 1A12," originally dated November 1993 (Revision 1, dated April 1994). 

The chronology of AIT's design activities is summarized below: 

December 10, 1993, the report was signed by AIT's subcontracted 

registered professional engineer (RPE). 

February 10, 1994, the ASME data report for the two fuel oil 

transfer filters were signed by AIT and the ASME Code Authorized 

Nuclear Inspector (ANI). 

May 3, 1994, the RPE, analysis engineer, and project manager 

resigned the Design Report for Revision 1 to the Design Report to 

include the drawing revision level of the four drawings used to 

manufacture the fuel oil filters. 

January 19, 1995, the RPE signed the Design Report Certification 

statement. 

March 27, 1995, the RPE signed a statement reconciling the revised 

design requirements for the filter cartridge elements with the 

seismic analysis. 

June 6, 1995, the RPE signed the Reconciliation Statement for 

nonconformances reports (NCRs) 392-1, -2, -3, and design change 

order 392-1. 

The team determined that on February 10, 1994 (when the ASME data reports were 

signed by AIT and the ANI indicating that the fuel oil filters conformed to 

the rules of construction of the ASME Code), the Design Report for Job 392 did 

not comply with the following requirements of the ASME Code: 

The Design Report did not include documentation that the Owner's 

review, required by NCA-3260(a), had been conducted, as required by 

NCA-3260(b). The Design Report, as submitted to the NRC on February 

6 
13, 1996, still did not include documentation that the Owner, or 

designee, had reviewed the Design Report to determine that all 

design and service loadings as stated in the design specification 

had been evaluated and that the acceptance criteria expl ici tly 

provided for in the ASME Code had been considered. 

The Design Report did not include the latest revisions of the 

drawings and therefore did not accurately indicate the 

reconciliation of design changes with the Design Report, as required 

by NCA-3250(d)(3) and NCA-3554. 

The Design Report did not include reconciliation 392-1, -2, -3, and 

design change order 392-1, as required by NCA-3554. 

Faiiure to verify adequacy of the design as required by Criterion I11 10 CFR 

Part 50, Appendix B and the above referenced ASME Code paragraphs was 

identified as an example of Nonconformance 99901292/96-01-02. 

In its transmittal to NRC dated February 7, 1996, AIT supplied additional 

documentation related to NCR 392-1. NCR 392-1, issued February 8, 1994, 

identified that the inlet and outlet nozzle flanged end connections were made 

by AIT from SA-516, Grade 70 plate material. Bechtel's design specification 

required the flanged end connections to be 150-lb. rated raised face slip-on 

flanges that comply with ANSI standard 816.5, "Pipe Flanges and Flanged 

Fittings." 

Pipe flanges that comply with ANSI B16.5 for the application specified in 

Bechtel 's design specification are forged carbon steel, normally manufactured 

to material specification SA-105, "Specification for Forgings, Carbon Steel, 

for Piping Components." 

This standard does not permit the use of SA-516 plate material. 

The NCR was dispositioned "use as is" with a technical justification that "it 

is not possible to upgrade a forged flange from commercial grade to nuclear 

grade in the time frame (schedule) available for this fast-track project." 

The NCR was approved by the QA Manager and the ANI on February 10, 1994. 

The team determined that AIT's disposition of NCR 392-1 did not conform with 

ASME Code requirements because neither the Design Report nor NCR 392-1 

addressed the requirements of paragraph ND-3362, "Bolted Flange and Studded 

Connections," that requires that flanges to other standards (in this case, 

other than B16.5) are acceptable provided they have been designed in 

accordance with the rules of ASME Code Section 111, Appendix XI, "Rules for 

Bolted Flange Connections for Class 2 and 3 Components and Class MC Vessels." 

This issue was identified as an example of Nonconformance 99901292/96-01-02. 

3.4.1.1.2 Materials 

The team reviewed the procurement and qualification of the following materials 

used in the construction of the fuel oil filters: 

7 

19 
(1) Fuel Oil Filter Body 

According to AIT Drawing 392-2, "Body for Oil Filter - Machining 

Detail - Item lAll and 1A12 (2 units)," Revision 1, dated December 16, 

1993, the pressure retaining body of the fuel oil filters was machined 

from 6-inch thick SA-516, Grade 70 plate. The finished rectangular-body 

dimensions were 7 1/2-inches high x 6-inches wide x 10-inches long. 

In its PO 20548, dated November 11, 1993, to American Alloy Steel (AAS), 

AIT ordered one piece of SA-516, Grade 70 plate, 6-inches thick x 11- 

inches wide x 18-inches long. The PO required the supplier to provide a 

certified material test report (CMTR), perform Charpy v impact tests, 

certify that no welding was performed, and imposed the reporting 

requirements of 10 CFR Part 21. The team determined that AIT's PO to 

AAS did not specify the impact-test requirements (i .e. , specimen 

orientation, testing temperature, and acceptance criteria) in its 

ordering information as required by ASME Code specification SA-20, as 

referenced by specification SA-516. 

Since AIT had not audited AAS t.o qualify it as a supplier, the material 

had to be upgraded for use in ASHE Code application. 

ASME Code Subsection NCA-3800, material upgrade in this case is 

performed by (a) assuring that no welding with filler metal had been 

performed on the unqualified source material, (b) performing or 

subcontracting a product analysis to verify the chemical composition of 

each piece of unqualified source material, and (c) performing or 

subcontracting all other requirements of the material specification on 

each piece of unqual if i ed source materi a1 . 

In accordance with 

AIT sent a sample of the plate material to a laboratory (PO 20563, dated 

November 30, 1993, to Ramball Test Lab, Inc.) and requested that 

chemical analysis, tensile-test, and impact-test be performed, and 

imposed the reporting requirements of 10 CFR Part 21. 

determined that AIT's PO to its laboratory did not specify the following 

parameters necessary to verify all of the requirements of the SA-516 

material specification in order to upgrade this material : 

The team 

a. the impact-test requirements (i.e., specimen orientation, 

testing temperature, and acceptance criteria) 

b. the final rolling direction of the plate specimen 

c. verification that the material conformed to the fine austenitic 

grain size requirements of specification SA-20, as referenced by 

SA-516 

AIT's CHTR for the SA-516 body material supplied to Calvert Cliffs was 

not in the job records or provided for the team's review. However, the 

lab report from Ramball TestLab, Inc. did not include the McQuaid-Ehn 

test to determine the austenitic grain size number, or, alternately, an 

analysis for aluminum contents to verify that the steel was made to using 

fine grain practice as required by SA-516. The lab's report also did not 

8 

20 
describe the tension-test and impact-test specimen orientation as 

transverse to the final rolling direction of the plate. 

The team also noted that, since AIT had not audited the supplier or 

performed specific examination for evidence of welding, there was no 

apparent basis for accepting the supplier’s certification that no welding 

had been performed on this material. 

The team concluded that AIT’s upgrade of the unqualified source material 

used for the fuel oil filter bodies supplied to Calvert Cliffs did not 

meet the requirements of NCA-3800 of the ASME Code. Failure to 

demonstrate compliance with the applicable technical and quality 

requirements for this item was identified as an example of Nonconformance 

99901292/96-01-03. 

(2) Inlet and Outlet Slip-on Flanges 

As described above, NCR 392-1, issued February 8, 1994, identified that 

the inlet and outlet nozzle flanged end connections were made by AIT from 

SA-516, Grade 70 plate material. 

AIT sent a sample of the plate material to d laboratory (PO 20579, to 

Ramball Test Lab, Inc.) and requested that chemical analysis, teiisiletest, and impact test be performed, and imposed the reporting 

requirements of 10 CFR Part 21. The team determined that AIT 

laboratory did not specify the foll-owing parameters necessary 

all of the requirements of the SA-516 material specification 

upgrade this material : 

s PO to its 

to verify 

n order to 

a. Mechanical testing requirements (i.e.? specimen orientation with 

respect to coupon configuration, testing temperature, and 

acceptance criteria) 

b. Verification that the material conformed to the fine austenitic 

grain size requirements of specification SA-20, as referenced by 

SA-516 

AIT’s CMTR for the SA-516 material supplied to Calvert Cliffs as pipe 

flanges was not in the job records reviewed at AIT. 

February 7, 1996, transmittal to the NRC, AIT’s CMTR was provided. AIT 

attached to the CMTR its lab reports for the test performed. 

However, in its 

The team determined that the lab reported impact test values did not 

comply with the requirements of 80-2331. For the three specimens tested, 

the lab reported an average lateral expansion of 15 mils, with the lowest 

of the three specimens at 8 mils. Paragraph ND-2331 requires for plates 

1 3/8 inches thick an average lateral expansion of 20 mils, with the 

lowest of the three specimens at 15 mils. Although it was not clear from 

the available procurement specification documents that impact testing was 

required for these flanges, the nonconforming results should have been 

9 

21 
recognized and dispositioned as a part of the receiving inspection 

process. 

reported impact values did not meet ASME Code requirements. 

In this case, AIT apparently failed to observe that the 

The team also determined that the lab reported chemistry values were 

significantly different than those rovrted in the original heat 

analysis. 

that reported in the heat analysis and Phosphorus, as reported by AIT's 

lab, was 25% lower than that reported in the heat analysis. 

For instance the lab reported Manganese at 25.4% lower than 

The team also noted that on the copy of the heat analysis (performed by 

the material manufacturer and supplied to AIT by its supplier AAS), AAS 

had applied a stamp and written in its customer (AIT), AIT's PO number, 

and the date mailed (December 9, 1993). 

and Accepted" stamp made by AIT with the initials "HM" and was dated 

October 25, 1993, leaving the appearance that AIT reviewed and accepted 

the heat analysis before AAS mailed to AIT. The same AIT stamp appeared 

on all three pages of the heat analysis. The team also noted the AAS's 

certification that no welding was performed on the plate material was 

dated February 8, 1994, after the date of AIT's CMTR (February 7, 1994), 

which, when certified, meant that all ASME Code requirements had been 

met. 

The same copy has a "Reviewed 

The lab report did not include the McQuaid-Ehn test to determine the 

austenitic grain size number, or, alternately, an analysis for aluminum 

contents to verify that the steel was made to fine grain practice as 

required by SA-516. 

test and impact-test specimen orientation as transverse to the final 

rolling direction of the plate. 

The lab's report also did not describe the tensionThe team also noted that, since AIT had not audited the supplier or 

performed specific examination for evidence of welding, there was no 

apparent basis for accepting the supplier's certification that no welding 

had been performed on this material. 

The team concluded that AIT's upgrade of the unqualified source material 

used for the pipe flanges supplied to Calvert Cliffs did not meet the 

requirements of NCA-3800 of the ASME Code. 

conformance with the applicable technical and quality requirements for 

this item was identified as an example of Nonconformance 99901292/96-01- 

03. 

Failure to demonstrate 

(3) Inlet and Outlet Pipe Nozzles 

In its PO 20568, dated December 1, 1993, to BBL Company, AIT ordered 

3-ft. of 3-inch, schedule 40, seamless plain-end carbon steel pipe, to 

material specification SA-106, Grade B. The PO required that BBL Company 

provide a CMTR, certify that no welding was performed, and imposed the 

reporting requirements of 10 CFR Part 21. 

Since the material was procured from an unqualified supplier, AIT elected 

to upgrade this material in accordance with the requirements of NCA 3800. 

10 

22 
AIT sent a sample of the pipe material to its laboratory (PO 20604, 

dated January 28, 1994, to Ramball Test Lab, Inc.) and requested that 

chemical analysis and tensile-test be performed. The team determined 

that AIT's PO to its laboratory did not specify the following tests 

and requirements necessary to meet all the requirements of SA-106, 

Grade B material specification: 

a. flattening test in accordance with SA-530, as referenced by 

specification SA- 106 

b. hydrostatic test. 

Additionally, AIT's PO to the test lab did not impose the reporting 

requirements of 10 CFR Part 21. AIT's CMTR for the SA-106, Grade B pipe 

nozzles supplied to Calvert Cliffs was not in the job records 

for the team's review. The lab report from Ramball Test Lab, Inc. did 

not include the flattening test to determine the ductility and soundness 

of the pipe; and the lab's report also did not include the hydrostatic 

test. 

for other pipe material upgrades, however, no documentation of a 

hydrostatic test for this pipe material was in the job records or 

provided for the team's review. The team concluded that AIT's upgrade of 

the unqualified source material used for the fuel oil filter pipe nozzles 

supplied to Calvert Cliffs did not meet the requirements of NCA-3800 of 

the ASME Code. 

technical and quality requirements for this item was identified as an 

example of Nnonconformance 99901292/96-01-03. 

provided 

The team noted that AIT had performed in-house hydrostatic tests 

Failure to demonstrate conformance with the applicable 

The team noted that the job record NCR log included NCR 392-2, issued 

December 7, 1994, that addressed discrepancy report 7 from an earlier ASME 

accreditation survey. The discrepancy report identified certain issues with 

AIT's upgrading unqualified source material for use in ASME Code applications. 

The NCR was dispositioned ''use as is" and the actions taken were that "all 

material were checked to comply with Code and QA Manual for upgrading." The 

NCR was completed and approved by the AIT QA Manager on June 20, 1995. The 

team concluded that AIT's disposition of NCR 392-2 was not supported by the 

team's findings described in the preceding paragraphs. 

3.4.1.1.3 We1 ding 

From its review of AIT's job records, the team determined that the primary 

welding procedure specification (WPS) used to fabricate the fuel oil filters 

was WPS WT-713, Revision 4, dated February 1, 1994, with its supporting 

procedure qualification record (PQR) 713, Revision 4. This WPS documented a 

gas tungsten-arc welding (GTAW) process that, according to the PQR, was 

qual ified in the as-welded condition (i .e., no postweld heat treatment) and 

minimum preheat of 50°F. 

class E70S-3, SFA-5.18, "Specification for Carbon Steel Electrodes and Rods 

for Gas Shielded Arc Welding." 

The WPS identified the filler metal specification as 

According to AIT's project engineer and production superintendent, the welding 

of the fuel oil filter assemblies was performed in the GTAW process. However, 

11 

23 
the actual filler metal used in the GTAW process was specification class 

EM12K, SFA-5.17, "Specification for Carbon Steel Electrodes and Fluxes for 

Submerged Arc Welding," normally used in the submerged arc welding (SAW) 

process. AIT did not, during the course of this inspection, produce a WPS 

that described the use of SFA-5.17, EMlEK filler metal in a GTAW process. 

team's review of the welding records identified the following concerns: 

The 

a. A minimum preheat temperature of 200°F was not specified in either 

WPS WT-713 or PQR 713 and may not have been performed, as required by 

ND-4622.7(b). The job records (traveler) did not document the actual 

preheat, if any, that was used. 

b. The weld metal qualification test (required by ND-2400 and performed 

by AIT's supplier for its PO 20537, dated October 27, 1993) did not 

qualify the SFA-5.17, EMl2K filler metal for use in the GTAW process. 

The suppliers weld metal qualification test addressed qualifying this 

filler metal for use in combination with a lot of submerged arc flux 

(F7A2) welded in the SAW process. 

Based on the above, the team concluded that, in this irIstance, AIT's weld 

metal did not comply with the applicable ASME Code requirements. 

demonstrate conformance with the technical and qual i ty requirements for 

welding material used in this process was identified as an example of 

Nonconformance 99901292/96-01-04, 

Failure to 

3.4.1.1.4 Fabrication 

The team's review of the available job records identified several concerns 

with AIT's fabrication process control for the fuel oil filters. For 

instance, the production traveler lacked specific detail to identify 

individual f i t-ups, examinations, we1 ding, and inspect ions. 

In one instance, the team determined that the guidance provided to the welder 

by the WPSs for the fit-up root gap dimension was not possible to achieve 

given the nozzle penetration machining dimensions specified on AIT Drawing 

392-2, "Body for Oil Filter - Machining Detail - Item lAll and 1A12 

(2 units)," Revision 1, dated December 16, 1993. 

two welding processes used to fabricate the fuel oil filters; WPSs WT-713 

(GTAW), described above, and WPS 701, a shielded metal-arc welding (SMAW) 

process. Both WPSs specified a root gap for groove welds of 1/16-inch to 

3/16-inch (0.0625- to 0.1875-inch). However, the hole size for the nozzle 

penetration was specified on Drawing 392-2 as 3.535-inches diameter (k 0.002- 

inch). Given a maximum hole size of 3.537-inches diameter and a 3-inchY 

schedule 40 pipe with an outside diameter of 3.5-inchesY with the pipe nozzle 

inserted in the hole, the resulting maximum root gap would be 0.0185-inchY not 

the 0.0625-inch desired root gap described in the WPS. 

The job records identified 

The team concluded that the 0.0185-inch root gap provided by AIT's design 

would not ensure that a full penetration weld was achieved, as required by 

NO-4244(b) of the ASME Code and Bechtel's design specification. Failure to 

ensure that welding was accomplished in accordance with applicable criteria 

was identified as an example of Nonconformance 99901292/96-01-04. 

12 
3.4.1.2 Job 523 - Filter Cartridges 

Bechtel issued PO CCDG0767, dated February 10, 1995, for 10 filter cartridges 

for the fuel oil filters supplied under its PO CCDG0065 (AIT Job 392 described 

above). Bechtel 's PO imposed procurement specification DG-80382, "Procurement 

Specification for Calvert C1 iffs Nuclear Power Plant Diesel Generator 

Project," Revision 2, dated December 23, 1994. The procurement specification 

identified the safety-related filter cartridges as ''basic components" and 

imposed the requirements of 10 €FR Part 21. 

also required that AIT have a QA prograni that conforms with NCA-4000 of the 

ASME Code for pressure retaining parts, and American National Standards 

Institute (ANSI) standard N45.2 for other parts determined to be safety 

related. The procurement specification also stated that for safety-related 

non-Code parts AIT shall either provide a QA program supplement to control the 

step-by-step processing of these items, or provide a QA program supplement 

which specifies that AIT's ASME Code QA program shall be used to process nonCode parts. 

The procurement specification 

AIT's job records did not indicate that any QA program supplement was 

developed for the supply of these filter cartridges. 

The PO also imposed the requirements of Bechtel's design specification SP-760, 

"Safety-re1 ated Fuel Oi 1 Fi 1 ters, 'I Revision 2, dated December 23 , 1994. 

Bechtel's procurement specification and design specification had been revised 

from requiring 3-5 micron filtration to a replaceable cartridge with 25 micron 

absolute particle size filtering capability from a fuel flow rate of 35 

gallons per minute (gpm). The filtration system was designed to have a 

maximum differential pressure of 5 5 psi across the filters at design flow and 

clean condition. The maximum differential at dirty condition and design flow 

shall be 10 psi. The cartridge shall be capable of withstanding a 

differential pressure greater that 15 psi. 

In its PO 20808, dated February 16, 1995, to Norman Ultraporous Filter 

Division of Bridgeview, Illinois, AIT ordered 10 filter cartridges, model 

588F-B2SAN-DOE for diesel fuel oil service. AIT specified that the filters 

shall be capable of filtering 25 micron absolute particle size from a flow 

rate of 35 gpm and shall be made of disposable fiberglass. AIT also specified 

that the filters shall be supplied with a certificate of conformance, the 

shelf life of the filter, the flow versus pressure drop data, and the center 

of gravity and weight. AIT's Attachment 8, "Specifications," was part of its 

PO requirements. Paragraph 3 of Attachment B required that all materials 

shall comply with ASME Code Section 11, Part A and Section 111, Class 3, 1986 

edition; paragraph 9 of Attachment B imposed the reporting requirements of 

10 CFR Part 21. 

AIT's supplier had not been audited or otherwise qualified as a supplier of 

safety-re1 ated basic components. 

Calvert Cliffs with a certificate of conformance (COC) that stated, in part, 

that the 10 filter cartridges were furnished in accordance with AIT's QA 

program and complied with the requirements of Bechtel's PO, ASME Code Section 

111, Class 3, 1986 edition, and the reporting requirements of 10 CFR Part 21. 

AIT suppl ied the f i 1 ter cartridges to 

13 

25 
'AIT, without any basis, certified that the non-ASME Code safety-related filter 

cartridges complied with ASME Code Section 111, Class 3 components. More 

importantly, AIT did not verify by either inspections, tests, or analyses that 

the design, material, and performance characteristics of the commercial grade 

filter cartridges provide reasonable assurance that the filter cartridges, 

when used as basic components, will perform their intended safety function, as 

Bechtel specified in its design specification. 

As discussed in paragraph 3.4, above, AIT does not have a QA program for 

manufacturing and supplying safety related non Code items under the rules of 

10 CFR Part 50, Appendix B. Likewise, AIT does not have a QA program or 

procedures for dedicating commercial grade material to achieve compliance with 

10 CFR Part 50, Appendix B. 

The team concluded that AIT provided an incorrect certification to Calvert 

Cliffs and that the filter cartridges supplied under this order did not comply 

with the Bechtel specification requirements. Failure to demonstrate 

conformance with the applicable technical and quality requirements for this 

item was identified as Nonconformance 99901292/96-01-05. 

3.4.1.3 Job 442 - Heat Exchangers (South Carolina Electric & Gas Co.) 

South Carolina Electric & Gas Company (SCEG) issued PO Q650395, dated May 19, 

1994 for three charging pump gear oil heat exchangers. These heat exchangers 

were to be designed and fabricated to the requirements of the 1971 edition of 

Section 111, Class 3, of the ASME Code. The PO also invoked SCEG procurement 

technical and quality requirements, 10 CFR Part 50, Appendix B, and 10 CFR 

Part 21. 

3.4.1.3.1 Materi a1 s 

AIT procured the heat exchanger tube material to the requirements of ASME 

SA 249. "SDecification for Welded Austenitic Steel Boiler, SuDerheater, HeatExchanger, ' and Condenser 

Marmon/Keystone, New Cast 

36 pieces, 20+ feet long, 

thickness. 

The Job record file conta 

ubes", TP 316. 

e, DE, an unqualified supplier. The PO specified 

.5-inch outside diameter, and .049-inch average wal 

The material was purchased from 

ned NCR 442-1 which was related to improper 

upgrading of this material. 

that, contrary to the requirements of NCA 3800 of the ASME Code, AIT had 

performed mechanical and hydrostatic tests on only one of the 36 tubes instead 

of testing each tube as required by the Code. 

noting that mechanical and hydrostatic testing has been performed on the other 

35 tubes. 

The December 1994 ASME survey had identified 

The NCR was dispositioned by 

A Certificate of Conformance from Laboratory Testing, Inc., dated April 17, 

1995, was on file stating that one tensile test, two flange tests, two 

flattening tests, one reverse bend test and one hardness test had been 

performed on each of 35 samples of tube material marKed with heat code/piece 

traceability by AIT. 

conformance with SA 249 requirements. 

The actual test results were provided and were in 

14 

26 
The files also contained two Hydrostatic lest Records. One of the records was 

dated July 15, 1994 and indicated that AIT had hydrostatically tested ''1" tube 

for Job 442 at 1000 psi with satisfactory results. 

the project engineer and by the QA manager. 

Record was dated March 13, 1995 and stated that 35 tubes with heat code 

markings M 26627-1 through M 26627-35 were hydrostatically tested at 1000 psi. 

The record was signed by 

The second Hydrostatic Test 

The inspection team reviewed the available records for the additional tests 

and identified several concerns regarding the closure of this NCR. 

(1) According to the route sheet for Job 442, after receiving inspection the 

tubes were cut to size per drawing requirements (four foot lengths). The 

route sheet signature blocks for receiving inspection and cutting 

operation were signed by the AIT QA manager on the same date (7/5/94). 

No steps or instructions for removal of test specimens or for hydrostatic 

testing of the tubes were included on the route sheet. To complete the 

fabrication process, each of the 3 heat exchangers would require 56 four 

foot lengths of tubing for a total of 168 such lengths. Subdividing the 

36 tubes received from the supplier would yield a total of 180 four foot 

segments (5 from each piece). Assuming no scrap, 12 four foot lengths 

(from no more than 12 of the 36 twenty foot long pieces purchased) would 

be left over for test specimens. 

The NRC inspection team could not determine, from the review of existing 

records, the source of material for the additional tests on each of the 

35 tubes as reported by Laboratory Testing Inc., and discussed above. 

Based on standard specimen dimensions as specified by SA 450, the 

applicable test specification, at least 20 inches of material would be 

required from each of the 36 tubes to perform these tests. 

above, there did not appear to be sufficient material available to 

perform this testing. 

were longer than 20 feet since they were ordered as 20+ lengths, but no 

receiving inspection or cutting records were available to support this. 

It is noted that SA 450 specifies a length tolerance of +1/8 inch, -0 for 

cut lengths of such tubing. 

As discussed 

It was suggested that the supplied tube pieces 

(2) With respect to hydrostatic testing, the NRC inspection team questioned 

when the testing of the 35 tubes reported in the Hydrostatic Test Record 

dated March 13, 1995 was actually performed since fabrication of the 

units was completed in 1994. According to AIT staff, all 36 tubes had 

been hydrostatically tested after delivery from the supplier and the 

March 13 record was generated based on the recollection of the test 

engineer and AIT's standard practice to test all tubes. Other than the 

previously discussed July 15, 1994 test record for 1 tube, no 

documentation was provided to support the additional tests. It was also 

noted that, based on the route sheet sign-offs (see previous paragraph), 

the July 15, 1994 hydrostatic test appears to have been performed after 

the tubes were cut into four foot segments. This could be explained by 

assuming that not all of the 36 tubes were cut to size on 7/5/94 since 

only 34 of the tubes were required for fabrication of the units. 

However, this would indicate that no more than 2 tubes were available for 

hydrostatic testing on July 15, 1994. 

15 

27 
Based on the above, the NRC inspection team concluded that the documentation 

presented during the inspection did not provide an adequate basis for the 

resolution of NCR 442-1 and did not demonstrate that the tube material was 

upgraded in accordance with the applicable ASME Code requirements. Failure to 

demonstrate conformance with the appl icable technical and quality requirements 

for this item was identified as an example of Nonconformance 99901292/96-01- 

03. 

The inspection team also concluded that failure to prepare and maintain 

records of hydrostatic testing of the tubes during the upgrading process 

violated the requirements of paragraphs 7.5.10 and 7.7 of AIT's QAM and 

paragraph NCA 3867.2 of Section I11 of the ASME Code. 

identified as Nonconformance 99901292/96-06-06. 

This issue was 

3.4.1.3.2 Process Control 

The NRC inspection team reviewed the route sheets used to control the 

fabrication of the heat exchangers and identified several concerns. 

Specifically, as discussed paragraph 3.4.1.3.1 (1) and (2), above, no 

instructions for removal of test ccli:.\ns from the source material to be used 

for upgrading were included on the route sheet. Hydrostatic testing of the 

tubes, which is an essential part of the upgrading process, was also not 

specified on the route sheets. It was noted that Step 14 of the route sheet 

for Job 442 specified "Roll tubes on LH and RH tubesheets" without specifying 

what procedure was to be used or identifying specific parameters for the 

rolling operation. 

sign-off blocks for this operation. 

It was also noted that there were no signatures in the 

The NRC inspection team determined that inspection and test activities and 

manufacturing operations for Job 442 were not sufficiently defined on the 

route sheets or referenced procedures to assure that these activities were 

performed in accordance with the applicable technical and quality 

requirements. Failure to provide procedures or instructions for activities 

affecting quality and to implement measures to assure that these activities 

are accomplished in accordance with these instructions was identified as 

Nonconformance 99901292/96-01-07. 

3.4.1.4 Job 331 - Heat Exchangers (PECO Energy Co.) 

PECO Energy Company (PECO) issued purchase order C ANE 379937, dated February 

18, 1992, for two heat exchanger tube bundles and two heat exchanger shells to 

be used as replacement components in the Limerick, Unit 1 residual heat 

removal system. 

the 1989 edition of ASME Code, Section 111, Division 1, Class 2. 

This equipment was to be fabricated to the requirements of 

3.4.1.4.1 Materi a1 s 

(1) Tube Sheets 

AIT procured the tube sheets from Explosive Fabricators, Louisville, CO, 

an unqualified vendor. 

inches diameter and 5 1/2 inches thick. 

These items were purchased as clad sheets, 60 7/8 

The tube sheets were explosively 

16 
clad with 1/2 inch AL6XN material and were to be qualified by ultrasonic 

testing in accordance with SA 578. AIT's PO to Explosive Fabricators 

also requested CMTRS for the material and required impact testing at 10°F 

to be performed in accordance with SA 370. No QA requirements were 

invoked on this vendor. 

AIT apparently upgraded the tube sheets by having chemical, mechanical, 

and ultrasonic testing performed by laboratories on their approved vendor 

list. Review of the test results from Ramball Test Labs showed Charpy v 

lateral expansion values for three tests reported as 46 mils, 36 mils, 

and 59 mils, respectively, with an average value of 47 mils. 

inspection team noted that these results do not meet the requirements of 

paragraph NC 2330 of the ASME Code or AIT purchase specification for this 

material which require minimum lateral expansion of 40 mils for each 

specimen tested. There was no documentation which would indicate retesting of this material or other basis for disposition of the 

nonconforming test result. 

The 

The NRC inspection t2am concluded AIT's upgrading of the tube sheet 

material for Job 331 did not fully comply with the applicable 

requirements of the ASME Code and that AIT had not demonstrated that the 

fracture toughness of the tube sheets met th; Code requirements. Failure 

to demonstrate conformance with the applicable technical and quality 

requirements for this item was identified as an example of Nonconformance 

99901292/96-01-03, 

(2) Shell and related material 

According to the Job records, on September 28, 1993, AIT issued PO 20497 

to American Alloy Steel Inc., an unqualified supplier, for various cut 

sizes of SA 516 Grade 70 steel plate. 

inspection report, items 3, 4, and 6 of this order were identified with 

the same material control number (M-2501). Although all of these items 

were one inch thick, they had different size configurations: 72inch by 

174inches, 2 by 240inches, and 60 by 168inches, respectively. According 

to the available records, only one material sample with this control 

number was sent to a test laboratory (Ramball) for upgrading tests. AIT 

issued PO 20550 on November 10, 1993 to Ramball requesting tensile, 

chemical, and Charpy tests on 1 inch plate, identified with material 

control number M-2501. The inspection team noted that assigning the 

same heat code number to several pieces of unqualified source material is 

contrary to the requirements of Paragraph 7.5.3.1 of AIT's QA Manual 

which states that "For material purchased as unqualified source material 

in accordance with Sec. XVII, a Material Control No. shall be assigned to 

each piece of unqualified source material". The inspection team also 

noted that to upgrade unqualified source material in accordance with the 

applicable provisions of NCA 3800 of the ASME Code, each piece of the 

material would have to be tested. 

laboratory did not identify the temperature for impact tests or indicate 

the location of the tensile and impact specimens with respect to the 

plate configuration (rolling direction). 

According to AIT's receiving 

Additionally, the PO to the test 

Test results were not in the 

17 

29 
Job record file. 

the upgrading process included verification of conformance to fine 

austenitic grain size requirements as specified in SA 516. 

Failure to demonstrate conformance with the applicable technical and 

quality requirements for this materi21 was identified as an example of 

Nonconformance 99901292/96-01-03. 

There was also no evidence which would indicate that 

(3) Semi -el 1 i pt i cal heads 

AIT issued PO 20494, dated September 28, 1993, to Trinity Industries, 

Cincinnati, Ohio, for two 52 3/4 inch diameter, 2 inch straight flange 

heads with 1 1/16 inch minimum thickness after forming, made from 

seamless 1 1/4 inch nominal thickness SA 516 Grade 70 plate. The heads 

were to be furnished in hot formed and normalized condition along with 

CMTRS and corner scrap material. 

requirements but requested authorization to audit Trinity Industries, 

Navasota, TX shop where the plates were apparently formed. 

AIT assigned material control tiumbors M-2527 and M-2528, respectively, to 

the two heads and on November 15, 1993, issued PO 20551 to Ramball 

requesting the performance of chemical analyses, tensile, and impact 

tests on samples from this material. The PO did not identify test 

temperature for the impact tests or specify the location of any of the 

test specimens with respect to material coupons. 

in the Job record file. 

process included the verification of fine austenitic grain size 

requirement as specified by SA 516. 

The PO did not invoke any QA 

Test results were not 

Also, there was no evidence that the upgrading 

Additionally, review of AIT's audit record files indicated that, although 

AIT had audited Trinity Industries, Cincinnati, Ohio facilities on 

February 6, 1992, there was no record of any audits having been performed 

at the Navasota, TX shop, where the heads were apparently fabricated. 

Contracting with an unqualified organization to supply safety related 

material and services (forming and heat treatment) without performing 

audit or surveillance at the location where these services are performed 

was identified as a nonconformance with Criterion VI1 of 10 CFR Part 50, 

Appendix B and Section 12.2.1 of AIT's QAM (Nonconformance 99901292/96- 

01-09). 

3.4.1.4.2 Manufacturing Process Control 

The NRC inspection team reviewed the disposition of selected nonconformance 

reports to assess AIT's control of the corrective action process. 

concerns were identified in this area. 

(1) 

Several 

Specifically: 

The recomended action for the disposition of NCR 331-4 was to plug holes 

using procedure APP-331-20 and to liquid penetrant test the seal welds on 

the plugs (no procedure specified). The "Action Taken" paragraph stated 

18 

30 
that tube holes had been plugged per 331-20 and that nondestructive 

testing had been performed and documented on the route sheet. 

Examination of the route sheet, however, showed no documentation of the 

such testing. 

(2) NCR 331-3 was written to address missing tube holes in baffles MK1 and 

MK3, identified after assembly. The recommended action was to cut off 

the untubed sections of the baffles per attached sketch and provide 

rep1 acement segments. 

(with drilled holes) filet welded to the existing baffles. Although the 

replacement baffle material was specified as SA 516 Grade 70, the sketch 

did not identify a material control number or heat number. The sketch 

also did not identify any nondestructive examination of the weld used to 

attach the replacement segment. 

paragraph 7.3.4 of AIT's QA Manual, which states that "If work is 

required as a result of nonconformity, a revised Route Sheet shall be 

issued," a revised route sheet was not found in the job files and there 

was no record of the repair operation on the original route sheet. 

The sketch showed over1 apping baffle segments 

Contrary to the requirements of 

The NRC inspectors identified AIT's failure to control and document repair 

activities as a nonconformance with paragraph 7.3.4 their QAM and with 

Criterion V of 10 CFR Part 50, Appendix B (Nonconformance 99901292/96-01-07). 

3.4.1.5 Job 4102 - Flow measuring Nozzles 

Union Electric Company issued PO 092710 on October 14, 1994, for two flow 

measuring nozzles for use in the Callaway plant. The nozzles were required to 

be fabricated to the requirements of Section 111, Class 3 of the ASME Code and 

to Union Electric specification M-1160 (Q) Rev 0. 

specified as type 316 stainless steel. 

The nozzle material was 

AIT purchased the nozzle material from Sandmeyer Steel Company, an unqualified 

vendor, to SA 240, TP 316 requirements and upgraded it to ASME Code by 

subcontracting the required testing. 

IMT, Inc., a vendor qualified by audit to perform these services. 

Machining services were contracted to 

One of the final operations 1 isted on the route sheet was "Clean/Prepare for 

shipment." 

each route sheet shall reference appl icabl e procedures by number and revision 

level, the route sheet did not specify a cleaning procedure nor was this step 

signed off. 

Specification for these nozzles (para. 7.1.1) specified that cleaning agents 

and rinse water shall contain less than 200 parts per million chlorides and 

that the cleaning procedures shall be made available to the buyer upon 

request . 

Contrary to Paragraph 7.3.2 of AIT's QA Manual, which states that 

The inspection team noted that the applicable Union Electric 

Failure to specify an acceptable cleaning procedure on the route sheet was 

identified as a nonconformance with Paragraph 7.3.2 of AIT's QAM and Criterion 

V of 10 CFR Part 50, Appendix 6. (Nonconformance 99901292/96-01-07} 

19 

31 
3.4.2 

During the course of this inspection, AIT had only one job for nuclear 

application in fabrication; eight suction stabilizers for Korea Electric Power 

Corporation (KEPCO). AIT’s Drawing N-391-1, “Suction Stabilizers,” Revision 

7, dated January 26, 1996, required the seismic class 1 vessels to be 

fabricated in accordance with ASME Code Section 111, Class 2, 1989 edition. 

Although these vessels were not intended for use by the U.S. nuclear industry, 

the team reviewed AIT’s fabrication activities to determine the effectiveness 

of reported recent improvements in AIT’s process control and inspections. 

Observation of Current Activities (Job 391 - Suction Stabilizer) 

The vessel shell sections for Job 391 had been rolled and the long seams buttwelded. All other activities were on hold pending the next ASME accreditation 

survey, when these items were intended to be used by AIT to demonstrate the 

adequacy of its QA program and compliance with ASME Code requirements. The 

team noted that some improvements in documentation on the Job 391 travelers 

for certain activities that affect quality had been implemented by AIT. 

However, the team determined that based on the observations identified below, 

significant improvements would be necessary to meet the quality expectations 

consistent with 10 CFR Part 50, Appendix B and ASME Code NCA-4000 

requirements. 

In general, inspection activities necessary to verify specified 

requirements for control of the process and quality were not 

documented and the traveler did not provide an adequate means to 

document these activities. For instance, the production traveler 

failed to address the identification, welding, and removal of 

temporary attachments. 

The inspection documentation did not identify characteristics, 

methods, acceptance criteria, or provide for recording objective 

evidence of inspection results. For instance, the production 

traveler failed to provide any acceptance criteria for the any of the 

QC sign-offs observed. Also, after review of the production 

travelers, the team determined QC did not inspect all production 

activities that affect quality. Instead, QC appeared to select 

certain inspections and therefore, most activities were not 

inspected. 

3.4.3 Interfaces with the Authorized Nuclear Inspector 

The authorized inspection agency of record during the fabrication and 

upgrading of materials for all previous jobs (except job 391) was one of the 

insurance company’s d/b/a Factory Mutual Engineering Association. The 

authorized nuclear inspector’s (ANI’s) records were not available for reviewed 

during the course of this inspection. 

job records described in this report, the team identified several concerns 

with the ANI’s activities that resulted in signing data reports for items that 

apparently did not comply with ASME Code requirements. 

On the basis of its review of the past 

In September 1995, AIT changed its authorized inspection agency to Commercial 

Union Insurance Company. Under this contract, the ANI is providing four hours 

20 

32 
per day, five days per week coverage of AIT's ASME Code activities. The team 

met with the ANI to review several of the past and current issues. 

determined from this inspection and its interviews with the ANI and AIT staff 

that many changes in process control have taken place in the short time the 

new ANI has been at AIT. 

monitoring AIT's present ASME Code activities. During this inspection, the 

AIT staff stated that the ANI'S four hours per day, five days per week 

contract was in place until AIT received its accreditation from ASME. After 

being accredited, the ANI services may be reduced. The team concluded that, 

based on the issues identified by reviewing past job records, significant ANI 

oversight of AIT's ASME Code activities may be necessary to ensure compliance 

with all technical requirements of the ASME Code. 

The team 

The team found the ANI knowledgeable and effectively 

3.4.4 Audits 

3.4.4.1 Internal Audits 

The inspectors reviewed QAM Section 10.0, Revision 8, "Audits,".dated June 

28, 1995, which define the purpose, use and methods of the audits conducted to 

assure adequate implementation of the controlled manufacturing system. This 

system covered audit methods, the conducting of audits, and reporting the 

results to responsible personnel to assure that QA system is .functioning in 

accordance with the AIT QA manual. 

The inspectors reviewed the results of the internal audits conducted the last 

two years and the audit schedule for 1996. Internal audits review the 

implementation of the various sections of the QA manual utilizing standard AIT 

check lists. They are performed at a frequency such that all phases of the QA 

program are audited once every twelve months. 

the inspectors included very little documented objective evidence of the areas 

reviewed. 

all the 1994 and 1995 audits (that represents 24 separate audit reports). The 

inspectors noted that QAM Section 10.3.1 states that "objective evidence of 

the area audited shall be noted on the checklist." 

The audit reports reviewed by 

In fact, only one Quality Audit Discrepancy Report was written for 

The inspectors also reviewed the annual audits of the QA department conducted 

by the QA manager and submitted to the AIT President going back to 1990 and 

each was done by a different QA manager. Again, these reports are very 

limited in documentation and rarely identified any problems with 

implementation of the QA program. Even the report to the President after the 

unsuccessful December 1994 ASME Survey did not identify any significant 

problems with QA program implementation. 

During discussions with the AIT QA Manager (who had only been hired into the 

position six weeks ago) concerning the depth and quality of the internal 

audits and the lack of documented objective evidence, he stated that he has 

already communicated to the President the need to conduct and document more 

meaningful and detailed audits. Failure to document objective evidence of 

areas reviewed and activities audited was identified as an example of 

Nonconformance 99901292/96-01-08. 

21 
3.4.4.2 External Audits (Vendor Evaluations) 

The inspectors reviewed QAM Section 6.0, Revision 9, "Procurement Control ,'I 

dated June 28, 1995, which describes the system which provides control of 

vendors of purchased items and services. This includes the preparation and 

maintenance of the Approved Vendors List (4"L) by the QA Manager. Paragraph 

6.3.1.1.1 stated that vendors holding valid ASME certifications are considered 

qualified without survey and automatically placed on the AVL within the scope 

of their certificate. 

perform audits of ASME certificate holders to verify that they are effectively 

implementing their QA program prior to purchasing material for use in products 

to be supplied to nuclear plants as meeting 10 CFR Part 50, Appendix B. 

Information Notices 86-21, Supplement 1, and 86-21, Supplement 2, "Recognition 

of ASME Accreditation Program for N Stamp Holders," provide information 

regarding the purchaser's responsibilities for verifying effective 

implementation of a supplier's QA program. 

example of Nonconformance (99901292/96-01-09). 

I 

There are no QA program provisions requiring AIT to 

NRC 

This issue is identified as an 

I Paragraph 6.3.1.1.2 did state that vendors not having ASME Section 111 

certificates are required to have tneir quality program surveyed, audited and 

qualified by AIT for placement on the AVL. 

The inspectors reviewed AIT's current AVL, as well as many of the previous 

versions to determine the qualification status of several vendors who have 

supplied material to AIT in the past for use on specific domestic nuclear 

jobs. The inspectors also reviewed external audits for several vendors who 

supplied material and services such as testing for the most recent jobs 

shipped to domestic nuclear plants including Limerick, V.C. Summer, Calvert 

Cliffs, and Callaway. During the review of the AVLs the inspectors identified 

that the Registered Professional Engineer (RPE) who is currently used by AIT 

for ASME Code design work was last audited by AIT in December 1994 and was not 

listed on the current AVL as qualified for providing engineering services. 

This issue is identified as an example of nonconformance (96-01-09). 

The inspectors reviewed approximately 12 audits performed by AIT to place 

vendors on the AVL as required by QAM Section 6.0. 

documented in an Audit Report and using a Vendor Quality Evaluation (VQE) 

document as a checklist. 

an Exhibit in either Section 6 or 10 of the QAM. 

a yeslno type of questionnaire and included very little documented objective 

evidence of the scope of the audit or the review conducted. Of the dozen or 

so audits reviewed by the inspectors, only one Quality Audit Discrepancy 

Report was ever issued. 

type of review conducted and is reflected in the depth of the objective 

evidence documented in the VQE. 

Nonconformance (99901 292/96-0 1-08) 

These audits were 

This VQE checklist was not referenced or included as 

The VQE checklist was really 

This finding appears to be a direct result of the 

This issue is considered as part of 

3.4.4.3 PhiladelDhia Electric Company (PECO) Audit of AIT 

The inspectors reviewed the results of an audit performed at AIT by PECO on 

April 26-30 and May 5, 1993. The audit was conducted utilizing the NUPIC 

Joint Audit Checklist, Rev. 4. No findings were identified, but the audit 

22 

34 
report included the identification and issuance of two recommendations. The 

Audit Summary stated, in part, "Overall, the AIT quality program relative to 

the supply of pi ping subassembl i es , pressure vessel s as nuclear service 

components and pipe shop fabrication was identified as adequate and 

effectively implemented. The audit team was particularly impressed with the 

AIT operations, even though the relative size of the company is smaller than 

most nucl ear suppl i er organizations . " 

The PECO audit appeared to review the inplementation of the appropriate areas 

of the AIT QA program, however, the results of the May 1993 PECO audit 

disagree considerably with this January 1996 NRC inspection and the December 

1994 and June 1995 ASME Surveys of AIT in the following areas: 

Design Control 

Procurement Con t r o 1 

Fabrication/Assembly/Special Proc, asses 

Test and Inspections 

Material Control 

Audits 

Document Control 

10 CFR Part 21 Implementation 

The inspectors concluded that May 1993 dudit pbrformed by PECO does not appear 

to be reflective of the level of the overall QA program implementation as 

evidenced by the NRC inspection team in January 1996 and the ASME Survey teams 

in December 1994 and June 1996. 

used this PECO audit as part of their basis for placing AIT on their approved 

suppliers list and ultimately procured charging pump gear oil heat exchangers 

for the V. C. Summer Nuclear Station. 

It was also noted that in April 1994 SCE&G 

3.4.5 Personnel Qualification and Training 

The inspectors reviewed QAM Section 4.0, Revision 9, "Personnel Training," 

dated June 28, 1995, which describes the system for the qualification 

indoctrination, training and education for personnel whose activities affect 

quality and require appropriate knowledge of the technical principles 

applicable to their duties. This includes auditor, inspection and test, and 

NDE personnel qual i f i cat i on , trai ni ng , and cert i f i cat i on. 

The inspectors attempted to review the indoctrination and training records for 

several employees currently working for AIT with different job descriptions. 

The QA Manager was not able to provide any training records going back further 

than 1995 for any of the current AIT employees. QAM Section 4.3.7 states that 

the QA Manager has the responsibility to maintain the indoctrination and 

training records for employees. 

able to locate employee training records since he began work for AIT in 

December 1994. 

empl oyee to empl oyee. 

The QA Manager stated that he has not been 

The training records that did exist were inconsistent from 

AIT, as part of&their corrective acticrn for several of the ASME Survey 

findings, committed to conduct training in several areas for certain 

empl oyees. None of these employee' s training records i ncl uded any 

23 
documentation that they had received this training. 

evidence that some of this training was performed by the fact that the 

Nonconformance Report (NCR) and corrective action documentation included a 

copy of the Training Records form for the subject identified on the NCR. 

However, these forms did not include a signature or initials of the persons 

who received the training in the Attendees portion of the form. 

that one person had completed the form. 

Nonconformance 99901292/96-01-10 was identified during this part of the 

inspection for failure to conduct appropriate indoctrination and training 

activities of personnel and to maintain quality records of these activities. 

The inspectors did see I 

It appeared I 

4 PERSONS CONTACTED 

The persons contacted during this inspection are listed below: 

Awer Industrial Technologies, Inc. 

* t Ahmed E. her, President 

* t Darrell E. Whitmer, Quality Assilra-ce Manager 

I 

t Hassan Mageid, Project Engineer 

t R. Cretella, Administrative Assistant 

Aziz Elsawy, Shop Supervisor 

t Attended the entrance and preliminary exit meetings * Attended entrance, prel imi nary and final exit meetings I 

24 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 205550001 

February 28, 1996 

Dr. Richard S. Siudek 

President, Nuclear Operations 

Asea Brown Boveri/Combustion Engineering, Inc. 

1000 Prospect Hill Road 

Windsor, CT 06095 

SUBJECT: NRC INSPECTION REPORT NO. 99900538/95-02 AND NOTICE 

OF NONCONFORMANCE 

Dear Dr. Siudek: 

This letter forwards the results of the inspection of your organization's 

activities at your Windsor, Connecticut, facility, conducted December 18-19, 

1995, by Mr. Stephen Alexander and Mr. James Isom of this office. Their 

findings have been discussed with some of the members of your staff identified 

in the enclosed report. The inspection was conducted to assess corrective 

actions implemented by your organization in response to a previously 

identified nonconformance regarding commercial grade dedication, to review 

your actions in response to the identification of deviations in reactor head 

closure studs supplied in 1990 to the Millstone Nuclear Station Unit 2, and to 

review your practices and procedures adopted pursuant to Part 21 of Title 10 

of the Code of Federal Resulations (10 CFR Part 21). 

the inspection and our findings are discussed in detail in the enclosed 

report. 

procedures and representative records, interviews with personnel , and 

observations by the inspectors. 

Areas examined during 

Within these areas, the inspection consisted of an examination of 

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

assurance program did not meet certain NRC requirements, specifically, 

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

Part 50. We found that contrary to the requirements of ABB-CE administrative 

procedures adopted pursuant to 10 CFR Part 21, ABB-CE did not (with regard to 

the 1990 Millstone Unit 2 reactor head studs), and does not, as a matter of 

stated practice or policy, document negative findings in evaluation of 

deviations in the manner prescribed by your Part 21 procedures when the 

deviations are obviously not safety significant or appear not to be safety 

significant upon initial review. However, in the instances which led us to 

this finding, you had documented technically sound dispositions of the 

particular deviations in question in accordance with other quality assurance 

procedures and had informed affected licensees or purchasers. Therefore, in 

these instances, although your organization was not intentionally following 

your Part 21 procedures, your actions met Part 21 requirements, and therefore, 

you were constructively in compliance with 10 CFR Part 21. 

findings and references to the pertinent requirements are identified in the 

enclosures to this letter. 

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

instructions specified in the enclosed Notice of Nonconformance. 

The specific 

You are requested to provide us within 30 days 

37 
Dr. R.S. Siudek -2- 

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

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

Public Document Room (PDR). 

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

it can be placed in the PDR without redaction. 

necessary to include such information, you should clearly indicate the 

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

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

the public. 

To the extent possible, your response should not 

However, if you find it 

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

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

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

The cooperation of your staff in this matter was greatly appreciated. Should 

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

them with you. 

Michael R. Johnson, Acting Chief 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Docket No.: 93900538 

Enclosures: 1. Notice of Nonconformance 

2. Inspection Report 99900538/95-02 

38 
NOTICE OF NONCONFORMANCE 

ABB-Combustion Engineering, Nuclear Operations 

Windsor, Connecticut 

Docket No. 9$6b0538 

Report No. 95-02 

Based on the results of an NRC inspection conducted on December 18-19, 1995, 

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

with NRC requirements. 

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

10 CFR Part 50 requires in part that activities affecting quality be 

prescribed by instructions, procedures, and drawings appropriate to the 

circumstances and that these instructions, procedures, and drawings be 

foll owed. 

Paragraph D.2 of ABB-CE, Nuclear Operations' "Administrative Procedure 

for Reporting Defects and Noncompliance," Revision dated September 1994, 

adopted pursuant to 10 CFR Part 21, defines deviations and failures to 

comply consistent with the defjnitions in 10 CFR 21.3. Paragraphs E.3 

and E.4 of this procedure require, consistent with 10 CFR 21.21(a), that 

deviations and failures to ~~.,,,.,ly be evaluated to identify defects or 

failures to comply associated with substantial safety hazards, and 

Paragraph E.4.b requires, consistent with 10 CFR 21.51, that such 

evaluations be documented and the records, using the form provided in 

the procedure, be retained for five years when no defects or failures to 

comply associated with substantial safety hazards are identified. 

Contrary to the above, ABB-CE did not (with regard to reactor head studs 

supplied to Millstone Nuclear Station, Unit 2, in 1990) and, as a matter 

of stated practice or policy, routinely does not document negative 

findings in evaluation of deviations or failures to comply in the manner 

prescribed by the procedures when the deviations or failures to comply 

are obviously not safety significant (i.e., could not create or are not 

related to a substantial safety hazard) or appear not to be safety 

significant upon initial review. 

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

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

with a copy to the Chief, Special Inspection Branch, Djvision 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 Zhday of February, 1996 

Enclosure 1 
REPORT NO.: 

ORGANIZATION 

INSPECTED: 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACT1 VITY : 

INSPECT ION 

CONDUCTED : 

LEAD INSPECTOR: 

OTHER INSPECTORS: 

REVIEWED BY: 

APPROVED BY: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF INSPECTION AND SUPPORT PROGRAMS 

99900538/95-02 

ABB-Combustion Engineering (ABB-CE) 

Nuclear Operations 

1000 Prospect Hill Road 

Windsor, CT 19406 

Virgil Paggen, Nuclear Licensing 

203-688-1911 

Nuclear steam supply system designer and manufacturer, 

technical services, spare and rep1 acement parts 

December 18-19, 1995, at the Millstone Nuclear Station 

dnd the ABB-CE Windsor facility 

Vendor Inspection -Section 

Special Inspection Branch 

Division of Inspection and Support Programs 

Off ice of Nuclear Reactor Regul at i on 

James Isom 

Special Inspection Section, PSI6 

A, 

H( L- m /J 

GFegoryg,//Cwal ina, Section Chief 

+hfi6 

Date 

Vendor infpection Section 

Special Inspection Branch 

Division of Inspxtion and Support Programs 

Office of Nuclear Reactor Regulation 

n 

Special Inspect&’ BranchDivision of Inspection and Support Programs 

Office of Nuc?ear Reactor Regulation 

Enclosure 2 
1 .o 

1.1 

0 

0 

0 

0 

1.2 

0 

0 

0 

1.3 

INSPECTION SUMARY 

InsDection Basis: 

Appendix By "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) 

ABB-Combustion Engineering (ABB-CE), Nuclear Operations Quality 

Assurance (QA) Program Documents and Procedures 

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

ABB-CE Nuclear Operat ions Admi ni strati ve Procedures for Reporting of 

Defects and Noncompliance in Accordance with 10 CFR Part 21 

InsDection Scope: 

Eva1 uat i on of ABB-CE Nucl ear Operat i ons procedures for dedi cation of 

commercial grade items (CGI5j f.-r use by ABB-CE in building, supplying 

spare parts for, or in servicing sa'ety-related equipment, including 

corrective actions implemented in I esponse to a previously identified 

nonconformance regarding commercial grade dedication 

Review of ABB-CE actions in response to the identification of deviations 

in reactor head closure studs supplied in 1990 to Northeast Utilities' 

(NU'S) Millstone Nuclear Station, Unit 2 (Millstone-2) 

Evaluation of ABB-CE's program and procedures and their implementation 

for reporting of 

Part 21, particu 

Vi ol ati ons: 

defects and 

arly regard 

noncompl 

ng the M 

ance adopted pursuant to 10 CFR 

llstone reactor head stud issue 

None cited. 

1.4 Nonconformance: 

(99900538/95-02-01) Contrary to the requirements of Criterion V of 10 CFR 

Part 50, Appendix By ABB-CE did not (with regard to reactor head studs 

supplied to Millstone-E, in 1990) and, ds a matter of stated practice or 

policy, routinely does not document negative findings in evaluation of 

deviations or failures to comply in the manner prescribed by the procedures 

when the deviations or failures to comply are obviously not safety significant 

(Le., could not create or are not related to a substantial safety hazard) or 

appear not to be safety significant upon initial review. 

of this report). 

(See Paragraph 3.3 

2 
2.0 STATUS OF PREVIOUS INSPECTION FINDINGS 

Nonconformance 99900538/93-01 (Closed). Part 50 of 10 CFR codifies Section 

I11 and Section IX of the American Society of Mechanical Engineers (ASME) 

Boiler and Pressure Vessel Code (ASME Code). 

dedication of commerci a1 grade i tems for use in nuclear safety-re1 ated 

applications addressed procuring material for applications requiring 

compliance with the ASME Code. 

preference for procuring this material from suppl iers with qual i ty assurance 

programs meeting 10 CFR Part 50, Appendix B and holding ASME material supplier 

certificates of authorization, rather than from commercial suppliers and 

dedicating the material. Subsequent to that inspection, the NRC established 

its position that dedication of commercial grade items for use as basic 

components, if performed properly (i .e., all critical characteristics are 

identified and verified), was an acceptable equivalent alternative to 

procurement of basic components, that were designed and manufactured under a 

quality assurance program meeting the requirements of Appendix B to 10 CFR 

Part 50. In September 1995, a new version of 10 CFR Part 21 was published 

reflecting that position. Nevertheless, in the interim, ABB-CE had revised 

its procedures to reflect preferential procurement from Appendix B- and ASMEqualified suppliers. The inspector reviewed ABB-CE's dedication procedures 

ABB-CE's QA procedures for 

However, me procedures did not indicate a 

against the latest applicable. require 

Part 50, Appendix B, and although not 

the NRC staff positions regarding ded 

Generic Letters 89-02 and 91-05. The 

(See Section 3.1 of this report). 

3.0 INSPECTION DETAILS 

3.1 Conmercial Grade Dedication: 

ents of 10 iiR Part ?! and 10 CFR 

regulatory requirements, per se, aga 

cation that were promulgated in NRC 

procedures were found to be acceptab 

nst 

e 

Section 21.3 of 10 CFR Part 21 now defines commercial grade items (CGIs) as 

items [presumably intended for use as basic components, i .e., in safetyrelated plant applications] that were not designed and manufactured under a 

10 CFR Part 50, Appendix B, QA program. It then defines dedication as the 

process, involving verification of critical characteristics, through which 

CGIs are verified to be [or rendered] suitable for use in safety-related 

applications. Finally it defines critical characteristics as identifiable and 

measurable attributes which when verified, provide reasonable assurance that a 

CGI will perform its safety function. In accordance with 10 CFR Part 50, 

Appendix A and Appendix B, structures, systems, and components are required to 

perform their safety functions and not fail in a manner adverse to safety 

under all design basis conditions (i.e., extremes of non-accident service 

conditions, including ranges of operating parameters, harsh environments from 

postulated design basis accidents, and seismic conditions). Dedication, being 

an activity affecting quality, is to be controlled under the applicable 

requirements of 10 CFR Part 50, Appendix B. It may involve several types of 

QA controls, but it most directly provides an alternative means for meeting 

Criterion 111, "Design Control ,'I and Criterion VII, "Control of Purchased 

Material, Equipment and Services." 

3 

42 
During this inspection, the inspector reviewed the latest eff,ctive revision 

of "Mechanical Engineering Department Technical Operating Procedure for 

Commercial Grade Dedication," MISC-MECH-TOP-007, Revision 01, issued 

October 11, 1993. The procedure was reviewed against the above regulatory 

requirements and for consistency with NRC staft positions on dedication as 

promulgated in NRC Generic Letters 89-02 ana 91-05. 

procedure acceptable with the following comments: 

The inspector found the 

3.1.1 

Engineering Department dedication procedures for not requiring the purchase of 

ASME Code material from qualified suppliers when available, rather than 

performing commercial grade dedication or material upgrades. The inspector 

noted that although the NRC's current position on commercial grade dedication 

is that it is considered, if properly performed, an acceptable equivalent 

alternative to procuring items intended for safety-related applications as 

basic components from qualified, 10 CFR Part 50, Appendix B, suppliers, 

Paragraph 4.2.2 of the procedure had nevertheless been revised to require that 

such items be procured as "safety-related items'' unless circumstances (e.g., 

schedule, availability, cast, etc.) prevent it. 

The previous nonconformance was cited against ABB-CE's Mechanical 

3.1.2 Section 4.1 on design review was, in general, excellent except that it 

did not explicitly address analysis of failure modes and effects adverse to 

safety. 

3.1.3 Section 4.2 on CGI applicabilitylsuitability was very strong; although 

it had not yet been revised to reflect the expanded definition of CGIs found 

in the revision of 10 CFR Part 21 published in September 1995. 

3.1.4 Section 4.3 on identification/selection of critical characteristics 

was excellent, reflecting the NRC definition of critical characteristics first 

promulgated in Generic Letter 91-05 and now codified in 10 CFR Part 21. In 

addition, it recognized the relevance of manufacturing process and materials 

in addition to form, fit, and function, and also recognized the need, where 

applicable, for conditioning and screening (e.g. , for particular performance 

capabilities). 

3.1.5 Section 4.4 on acceptance was noted for a particular strength in that 

it recognized the appropriate selection of one or more of the four acceptance 

methods given in Electric Power Research Institute Report NP-5652, "Guide1 ine 

for the Utilization of Commercial Grade Items in Nuclear Safety Related 

Applications (NCIG-07)," and conditionally endorsed in NRC Generic Letter 

89-02, for acceptance or verification of individual critical characteristics, 

that is, that a combination of acceptance methods may be employed to accept a 

given CGI. Table 3 elaborated on the four acceptance methods and was strengthened by the inclusion of the restrictions on the use of Methods 2 and 4 from 

NRC Generic Letter 89-02. 

3.1.6 Paragraph 4.5.2.4 addressed sampling. Although it referred only to 

purely statistical sampling bases as in MIL-STD-1OSE (as opposed to, for 

example EPRI's sampling guideline), it was strengtnened by the restriction of 

only using sampling on lots or batches from the same manufacturer. 

lot traceability and homogeneity or consistencj among batches or lots were not 

Batch or 

43 
mentioned Ruglicitly, but the procedure did require the consideration of 

complexity, safety function, and vendor history, which may be effective in 

establishing batch/lot homogeneity and consistency among batches/lots. 

Paragraph 4.5.2.5 appropriately addressed sample expansion to 100 percent upon 

finding a defect. 

the language of MIL-STD-105E, may introdrice tinwanted imp1 ications in the 

context of dedication because of its special meaning in 10 CFR Part 21. 

Use of the word defect, although possibly consistent with 

3.1.7 

critical characteristics that cannot be verified by inspection or test of the 

finished product. Requiring such justification is within ABB-CE’s prerogative 

and is consistent with regulatory guidance. Then the paragraph addressed (but 

did not require) QC representative surveillance, presumably referring to 

source verification, EPRI Method 3. The use of alternate acceptance methods 

for critical characteristics that are not readily verifiable by Method 1, 

special tests and inspections, is encouraged and is also consistent with 

regulatory guidance. However, the paragraph then stated that such a critical 

characteristic may have to be accepted “solely on the manufacturer’s 

certification ...” This allowance is inconsistent with regulatory guidance in 

that it allows the use of a manufacturer’s certification alone as the basis 

for acceptance of one or more critical characteristics. The NRC has long held 

that vendor certifications, unless validated by some means such as QA audit or 

commercial grade survey, are not adequate as the sole basis for acceptance. 

The inspector recognized that ABB-CE may have compensatory provisions to that 

effect in its procedures for supplier selection and qualification, but as a 

stand-alone provision in this procedure with no qualifying constraints stated, 

the statement in question was inconsistent with regulatory guidance, 

acceptance requirements in the procedure, and may be inconsistent with other 

procedures as well. 

Paragraph 4.7.2 appeared to require justification for not verifying 

3.2 Reactor Head Studs: 

The inspectors reviewed ABB-CE’s disposition of various discrepancies found in 

60 reactor head studs (RHSs) supplied by ABB-CE to Northeast Utilities’ (NU’S) 

Millstone Nuclear Power Station (Millstone) in September of 1990. The 

replacement studs were modified with special ring grooves to accommodate a new 

quick-release hydraulic tensioner. The initial discrepancies were identified 

during receipt inspection at Millstone, Unit 2, and documented in 

Nonconformance Report (NCR) 290-566, dated September 17, 1990. The NRC 

inspectors reviewed this and related subsequent NCRs, their 

dispositions/resolutions, and related drawings and documents at Millstone on 

December 18, 1995. 

On the basis of the original receipt inspection dimensional discrepancies, 

engineers from Millstone, ABB-CE, and the RHS manufacturer, PMC Industries 

(under contract to ABB-CE), performed additional inspections of 100 percent of 

the RHSs and other ancillary equipment. These reinspections resulted in 

documenting additional nonconformances in NCRs 290-575, 290-589, 290-235, and 

290-239. The NRC resident inspectors documented their review of Millstone’s 

resolutions to these NCRs in Section 7.3 of NRC Inspection Report 

50-245192-25, 50-336192-27, 50-423192-24. 

5 

44 
During this December 1995 NRC inspection, the NRC inspectors determined 

through review of the NCRs and their disposition documents and interviews with 

engineers at Millstone and at ABB-CE headquarters that the nature of the 

discrepancies identified by the five NCRs would not affect the safety function 

of the RHSs. The RHSs’ safety function is to maintain the integrity of the 

reactor coolant pressure boundary by fastening the reactor vessel head to the 

reactor v5ssel. This function is performed by the top external closure nut 

threads and bottom external threads. The top external threads accommodate the 

reactor vessel closure nuts which fasten down the reactor vessel head flange 

and the bottom external threads are screwed into the reactor vessel flange. 

The reinspections performed by ABB-CE and PMC on September 24 and 25, 1990, on 

100 percent of the RHSs found that there were no RHS external thread 

dimensional deviations which might lead to thread galling, about which 

concerns had been rai sed. 

The identified dimensional discrepancies in the external threads of the RHSs 

had also given rise to a concern about erosion of the parts as a result. 

licensee concluded (and the inspectors agreed) that there were no conditions 

(due to the identified nonconformances or otherwise) under which any of the 

threads on the studs, nuts, or vessel flange would be expected to undergo 

erosion of any kind. 

The 

The NU Systems Quality Service Department witnessed the 100-percent 

reinspection mentioned above. ABB-CE documented the results of the inspection 

in a September 28, 1990, letter to Millstone (NSPT-90-713). The inspection 

confirmed that all bottom external threads were acceptable using go and no-go 

thread gage rings. All top external threads also passed inspection using the 

modified, slow-lead go-ring except for nine pieces (serial numbers (S/Ns) 11, 

12, 20, 22, 35, 41, 51, 62, 64) which did not fully accept the go-ring gage. 

For those nine RHSs which would not accept the go-ring, ABB-CE recommended 

that Millstone personnel check the RHS upper external threads with their 

closure nuts prior to their use to ensure that manufacturing defects were not 

present. 

that, potentially, the top external threads may be oversized. Millstone 

personnel wrote work order, AWO M2-90-11495, to run reactor vessel closure 

nuts over the threads to ensure proper fit of the nuts. 

able to satisfactorily install the reactor vessel closure nuts over the upper 

external threads of all nine affected studs. The licensee postulated that the 

reason some studs would not accommodate the go-gage fully was due to the 

build-up of manganese phosphate applied to the studs after machining. 

Manganese phosphate is applied as an anti-galling and corrosion inhibiting 

coati ng . 

The inability of these nine RHSs to accommodate the go-ring meant 

The mechanics were 

The ABB-CE and PMC engineers also performed inspection of all top and bottom 

RHS internal threads. The internal threads are machined into each end of a 

1.5-inch hole through each stud to accommodate the measuring rods. In the 

September 28, 1990, letter from ABB-CE to Millstone, ABB-CE stated that most 

studs required minor removal of residual excess phosphate coating materi a1 

from their internal threads. After this, all internal threads accepted the 

specified go-gauge at both ends. However, four studs, (S/Ns 6, 26, 51 and 59) 

were found to have slightly oversized internal threads (i .e. inside diameter 

6 
was slightly too large). ABB-CE recommended that these four studs could be 

used as is based on the following: The top internal threads are used for the 

lifting eyes used during installation, removal and other handling of the 

studs. The top internal threads also hold the top closure screw (to keep out 

foreign material and prevent corrosion) during plant operation. The bottom 

internal threads provide for attachment of the bottom closure screw during 

operation and accept the mounting piece for the measurement rods during 

tensioning. None of the internal threads perform any plant safety function 

(except personnel safety duri ng hand1 i ng . 

Millstone station dispositfoned the oversized internal threads on the bottom 

of studs, S/Ns 6 and 26, as use-as-is because in addition to these threads 

having no safety function, the bottom closing screw is staked in two places to 

ensure that it does not come loose. 

I, 

For the two studs that had oversized top internal threads, Millstone station 

performed a load test at twice the RHS weight for 10 minutes under work order 

AWO M2-90-11711. Load testing of studs S/Ns 51 and 59 (S/N 59 renumbered as 

S/N 47 due to its position) was reported as completed satisfactorily on 

October 10, 1990. 

After completing inspections of all RHSs at Millstone-2, ABB-CE concluded 

that, with the exception of the four studs which were found with oversized 

internal threads, all 60 of the RHSs supplied to Millstone were functionally 

acceptable for their intended service. For those studs which were found with 

oversized internal threads, ABB-CE issued Corrective Action Report No. QP-90- 

024 to PMC and requested PMC to determine the cause and actions taken to 

prevent recurrence. 

During a telephone conversation with PMC’s QA manager subsequent to the 

inspection, the inspectors learned that PMC was not able to determine why the 

oversized internal threads for the four studs were not detected by the PMC 

inspectors. 

dimensions of various parts of the RHSs and had performed other inspections, 

PMC inspectors did not detect the oversized internal threads on 4 of the 60 

RHSs. PMC postulated that internal thread irregularities (e.g., burrs) may 

have prevented acceptance of the No-Go plug, thus masking the oversize 

condition. Then such irregularities would have been removed by special plugs temporarily installed in each end of the studs used for turning them on a 

magnetic particle inspection machine, but it was not PMC’s practice at the 

time to recheck the internal thread dimensions after this test. PMC’s QA 

manager stated that PMC has advised all its inspectors of the problem and that 

gaging the internal threads after magnetic particle inspection would become a 

standard practice. 

According to the QA manager, although PMC had taken as-built 

ABB-CE stated in its response letter to Millstone (CE 1etter:EXT-90-296) that 

it planned the following actions to prevent recurrence of similar problems: 

1. ABB/CENP [Nuclear Projects] would request customer (uti1 ity) 

participation in final source inspection activities at PMC Industries. 

7 

46 
2. The above source inspection is to be performed after completion of 

phosphate coating operation and wi 11 i ncl ude checking of sel ected 

critical and noncritical attributes. 

3. ABB/CENP procurement documents shall require submittal of dimensional 

inspection data for each reactor vesse! stud. 

The NRC inspectors queried two other utilities who had purchased reactor head 

studs from ABB-CE around the 1990 time period to determine whether there were 

discrepancies found at other facilities. 

problem with reactor head studs. 

The two utilities reported no 

The inspectors determined that the non-dimensional discrepancies identified in 

NCR 290-566 were also satisfactorily resolved by Millstone in consultation 

with ABB-CE and PMC. Two of these, Items 1 and 2 were only apparent 

discrepancies. NCR 290-566, Item 1, stated that the NU purchase order (PO 

877539) for the RHSs required the studs to be made in accordance with NU 

Service Company (NUSCO) Drawing 25203-29139, Sheet 8. However, Item 1 stated 

that the certification (from PMC) certified to ABB-CE Drawing E-18767-650-001 

instead. Closer review of the drawings revealed that the originally specified 

NUSCO drawing, 25203-21939, had been revised to reflect the new design of the 

upper part of the RHS. In this reviaion, Revision 9, Sheet 8 of NUSCO Drawing 

25203-29139 no longer showed the details of the RHS, but now referenced a 

revised Sheet 6 of the same drawing. 

redesignation of ABB-CE Drawing E-18767-650-001 that ABB-CE had developed and 

NU had approved for fabricating the new design RHSs. 

Item 2 stated that the PMC certification mentioned only "phosphate" coating; 

whereas, manganese phosphate had been specified by the drawing. 

certification revealed that it certified that the studs had been given a 

"phosphate coating in accordance with ABB-CE M&P Specification 4.4.4.1(A) .I' 

Section 2.1 of this specification identified the material to be used as 

manganese phosphate. Therefore, the licensee had concluded (and the NRC 

inspectors agreed) that the certification was satisfactory in that it 

certified to use of the correct coating material by reference to the ABB-CE 

spec i f i cat i on. 

The new Sheet 6 was simply the NUSCO 

Review of the 

In addition, Item 5 of NRC 290-566 identified what appeared to the receipt 

inspector to be arc strikes on the end of stud S/N 40. The NRC inspectors 

were concerned that although the arc strikes were located on a non-critical, 

non-stressed surface, arc strikes are typically associated with welding which 

should not have been performed as part of fabrication on any part of the studs 

which are ASME Code, Section 111, Class 1 components. 

surface discontinuities should have been documented in the Code Data Report, 

yet none reportedly were. According to a September 10, 1990, letter from PMC, 

magnetic particle inspection of the studs was performed using a coil for the 

longitudinal field shot and direct current application (4200 amperes through 

the stud, end to end) for the circular field shot. 

claimed that no relevant or non-relevant indications were detected on the 

surfaces being examined during testing, the licensee and ABB-CE postulated and 

PMC confirmed (also later in a telephone conversation with the NRC inspector) 

the likelihood of the arc strikes occurring when the magnetic particle 

Any weld repairs of 

Although the PMC letter 

8 
equipment current probes were removed from (or possibly connected to) the stud 

ends without their being first deenergized. 

satisfactory explanation for why the arc strikes were missed other than their 

having most probably occurred after the testing was completed and the studs 

then not being inspected further by PMC. 

inspector in the telephone conversation subsequent to the NRC inspection that 

PMC technicians were aware of this problem as well as others related to 

inadequate QC at PMC and steps had been taken to correct deficiencies 

identified by ABB-CE and prevent recurrence, as documented in ABB-CE’s audit 

reports and re1 ated correspondence. 

PMC was not able to offer a 

The PMC QA manager told the NRC 

While reviewing the various documents for this inspection, the inspectors 

identified some errors in the original NRC inspection report on this issue 

(50-245/92-25, 50-336/92-27, 50-423/92-24). First, the RHSs were purchased 

for Unit 2, not Unit 3, as the inspection report stated. Also, the ABB-CE 

subcontractor who machined the RHSs (from material provided by ABB-CE) was PMC 

Industries and not PCI Energy Services. The reports stated that ABB-CE had 

placed PCI on the ABB-CE approved vendor list based solely on PCI’s holding an 

ASME certificate and that ABB-CE was not verifying that PCI had a quality 

assurance (QA) program tc! control characteristics not covered by the ASME 

Code. 

because PCI was never involved. 

However, this past ABB-CE practice was not relevant to the RHS issue 

During this Cecember 1995 inspection, ABB-CE stated that PMC iildcistries did 

not have an ASME certificate to manufacture the RHSs, but that PMC had always 

manufactured RHSs under ABB-CE’s oversight and using ABB-CE’s ASME 

certification. ABB-CE stated that except for a few studs made by ABB-CE 

itself at its Chattanooga, Tennessee, shop, PMC had fabricated all the RHSs 

supplied by ABB-CE for various sites including some other than ABB-CE-designed 

plants. ABB-CE stated that it had performed a comprehensive QA implementation 

audit of PMC on October 23-25, 1989 to requalify PMC as a supplier to ABB-CE 

in preparation for the Millstone job. 

RHSs at Mil 1 stone, ABB-CE has performed several more annual QA imp1 ementat ion 

audits of PMC, the last one being completed in May 1995. 

Also, based on the problems with the 

The inspectors had no further concerns reaardina the aualitv of reactor head 

studs provided by ABB-CE to Millstone in i990, ;or 

ABB-CE for other sites. 

3.3 10 CFR Part 21 Proaram: 

3.3.1 Procedural Adequacy 

During this inspection at ABB-CE, the inspector rev 

ther stids provided by 

ewed the vendor’s 

procedures adopted pursuant to 10 CFR Part 21. The current revision of 

“Administrative Procedure for Reporting of Defects and Noncompliance,” dated 

September 1994, was, in general, consistent with the provisions of Part 21 and 

had been updated to reflect the changes in the version of the regulation that 

became effective on October 29, 1991, with one exception. Paragraph E.2 which 

described the ways in which information potentially reportable under Part 21 

would be discovered contained a weakness. 

the language of this section could be interpreted by some employees as 

The inspector was concerned that 

9 

48 
requiring that they report only those problems that they believe could create 

a substantial safety hazard or lead to exceeding a technical specification 

safety limit, yet, this section did not address reporting of deviations and 

failures to comply (FTCs) as defined in the procedure and in Part 21. This 

language could have the effect of calling upon any employee who may become 

aware of a nonconforming condition to make judgements about the safety 

significance of that nonconforming condition, i.e., identifying defects and 

FTCs associated with substantial safety hazards (SSHs), that only certain 

employees, managers, and the Nuclear Safety Committee would normally be 

qualified (and authorized by the procedure) to make. 

the procedure as written, potentially safety-significant deviations or FTCs 

might be prematurely dismissed and not be formally evaluated tc identify 

defects or FTCs associated with SSHs as required by §21.21(a). 

deficiency in procedures adopted pursuant to 10 CFR Part 21 constituted a 

minor violation of 10 CFR 21.21(a) as described in NUREG-1600, "General 

Statement of Policy and Procedures for NRC Enforcement Actions," Enforcement 

Policy (previously contained in Appendix C to 10 CFR Part 2 of Title 10 of the 

Code of Federal Requlations), and as such, will not be cited. 

The concern is that with 

This 

3.3.2 Procedural Compl i ance 

Even though the ABB-CE Part 21 procedure defined deviation using the same 

words as the regulation, it was ABB-CE's position, clearly stated by ABB-CE 

staff in discussions with the NRC inspectors, that ABB-CE interpreted the 

regulation (and its procedure) to mean that only those deviations that appear 

to have potential safety significance (as determined by an undocumented 

preliminary evaluation) would undergo formal evaluation pursuant to §21.21(a) 

and Sections E.3, E.4, and E.5 of the procedure to identify defects and FTCs 

associated with SSHs. Conversely, according to ABB-CE staff, conditions that 

could be deemed deviations or failures to comply, as defined in 421.3 and in 

the procedure, would not be evaluated pursuant to §21.21(a) if it was apparent 

to the prel imi nary evaluator (said to typical ly be a department manager, 

supervisor or cognizant engineer), that the condition, al though technically a 

deviation or FTC, did not have any potential safety significance. Although 

Part 21 does not specify who must perform the evaluation, this preliminary 

evaluation is not provided for in the procedures and, contrary to Paragraph 

E.4.b of the ABB-CE procedure and 521.51, negative findings are not typically 

documented and retained under these circumstances using the forms provided for 

this purpose in the procedure. 

The inspector reviewed ABB-CE records of a1 1 521.21 (a) eval uat i ons completed 

in the last five years and found that ABB-CE has documented evaluations of 

several potentially safety significant deviations and FTCs, some of which 

resulted in the identification of defects or FTCs related to SSHs, and some of 

which did not. In all cases in the file, the appropriate reports were made. 

However, during this inspection, the inspectors reviewed a case of apparently 

safety-insignificant nonconforming conditions, which nevertheless constituted 

departures from technical procurement specifications in basic components 

identified after shipment (i.e. deviations, as defined in Part 21 and the 

procedure), but for which ABB-CE did not document evaluation in accordance 

with the Part 21 procedure. 

sition of these issues through other means such as its supplier Deviation from 

Instead, ABB-CE documented evaluation and dispo10 

49 
Contract Request (DCR) procedures or other QA procedures that would otherwise 

normally be used to correct or disposition only those nonconforming conditions 

identified before the basic components involved were supplied/ offered for use 

at NRC-1 icensed facilities. Such nonconforming conditions identified before 

shipment technically would not be deviations, per se, as defined by Part 21. 

Being not yet delivered, the presumption in Part 21 is that the problems can 

be corrected before delivery of the affected basic component(s), and hence, 

Part 21 is not applicable. However, aftw shipment, the lincorrected problems 

become deviations (or FTCs), thus requiring evaluation under §21.21(a). 

for example, in the case of the 60 reactor head closure studs supplied by 

ABB-CE to Millstone-2 in 1990, ABB-CE used its normal QA process (in this case 

the DCR procedure) to disposition certain nonconforming conditions (as 

discussed in Section 3.2 above) in several of the studs. Some obviously 

nonsafety-signi f i cant deviations from technical procurement specifications in 

terms of certain out-of-tolerance, but non-critical dimensions on several of 

the studs, were discovered by the stud manufacturing facility (PMC). These 

were dispositioned in a technically appropriate manner In accordance with ABBCE's DCR procedures as dwumented in correspondence rev'ewed by the 

inspectors. When additional nonconformances apparently missed by PMC (the 

studs were direct-shipped to Millstone from PMC and not inspected by ABB-CE) 

were discovered by Millstone's receipt inspectors, ABB-CE, NU and PMC jointly 

conducted a 100-percent reinspection of all the studs supplied and identified 

a few more nonconformances in September 1990. ABB-CE documented the disposition of all these similar, but post-delivery conditions using the same DCR 

procedure. 

Although the NRC inspectors agreed with the technical conclusions and ultimate 

dispositions reached by ABB-CE, ABB-CE's use of its QA procedures instead of 

the Part 21 procedure, while consistent with ABB-CE's practice and stated 

policy regarding Part 21 evaluations, was not, in the inspectors' judgement, 

strictly in accordance with the requirements of ABB-CE Part 21 procedures. 

However, the only deviations identified by the inspectors that were not 

evaluated and documented strictly in accordance with the Part 21 procedure, 

were nevertheless documented and retained (in accordance with other, QA 

procedures), and affected 1 icensees or purchasers had been informed. In these 

instances, although ABB-CE was not intentionally following its Part 21 

procedures, its actions ultimately met Part 21 requirements, and therefore, it 

was constructively in compliance with 10 CFR Part 21. 

Nevertheless, ABB-CE's failure to document the evaluation of deviations, 

albeit with negative findings, in accordance with procedures governing 

activities affecting quality (i.e., the Part 21 procedure), coupled with ABBCE's stated policy or interpretation that evaluation of obviously non-safetysignificant deviations need not be documented, constituted a nonconformance 

with respect to Criterion V, "Instructions, Procedures, and Drawings," of 

Appendix B to 10 CFR Part 50, and accordingly is cited as Nonconformance 

99900538/95-02-01, 

Note that the failure in this instance to follow the ABB-CE Part 21 procedure 

is not considered a violation of 10 CFR Part 21 because even though ABB-CE's 

11 

50 
stated policy would be considered contrary to Part 21, the ?drt 21 procedure 

in this regard is not. 

in question, the deviations had been evaluated and the records were still on 

file, although not in the form or file prescrihed by ABB-CE’s Part 21 

procedure. As an incidental consideration, it was recognized that any 

question of lack of Part 21 file records of evaluations of the deviations 

cited was moot because at the time of this inspection, the Part 21-mandated 

retention period for any records of the evaluation of the deviations (had 

there been any) in the example cited had already expired. Therefore, at the 

time of this inspection, ABB-CE would no longer have been required to have 

records of the evaluation of these particular deviations in its Part 21 files 

even if they had been documented in accordance with the Part 21 procedure. 

Further, there was evidence that, in the only instance 

Nevertheless, the inspectors cautioned ABB-CE that adherence to its orally 

stated practice and not to the letter of its written Part 21 procedure could 

be viewed by the NRC as violation of the regulation. 

NRC Office of General Counsel, the inspectors confirmed the position that 

Part 21 was written for general use by licensed facilities and suppliers of 

basic components. A supplier, regardless of how knowledgeable, does not have 

the authority to wave the requirement tc, ;valuate an actual deviation or 

failure to comply (and document and retain the results of that evaluation), 

regardless of how trivial it may first appear. 

In consultation with the 

4.0 PERSONS CONTACTED: 

4.1 ABB-CE: 

Gary S. Bloomquist, Manager, Nuclear Quality 

James A. Noyes, Manager, Nuclear Spare Parts 

Theodore S. Bernard, Program Manager, Nuclear Spare Parts 

Ian C. Rickard, Project Director, Engineering Operations 

David Sibega, Supervisor, Plant Components 

Mark W. Stewart, Supervisor - QA, Nuclear Spare Parts 

Charles Molnar, Nuclear Licensing 

Virgil Paggen, Nuclear Licensing 

4.2 PHC Industries: 

Harry Dods, QA Coordinator 

4.3 Northeast Utilities/Millstone-2: 

Mario Robles, Licensing 

Tom A. Moore, System Engineer 

Michael J. Haeflich, Procurement 

John N. Coleman, Procurement Inspect i on 

Mark A. Surprenant, Supervisor of Procurement Engineering 

4.4 m: 

Russel Arrighi, Resident Inspector, Millstone 

Serita Sanders 

Edward Baker 

12 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 2055~0001 

February 6, 1996 

Mr. Lance Simms 

General Manager 

Brand Fire Protection Services, Inc. 

721 Roguet Club Drive 

Addison, IL 60101 

SUBJECT: NRC INSPECTION NO. 99901020/95-01 

Dear Mr. Simms: 

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

(NRC) inspection of Brand Fire Protection Services (Brand) facilities in 

Addison, Illinois by this office on November 6-9, 1995. 

The inspection was conducted to evaluate Brand's quality assurance program as 

implemented in the process for providing fire barriers penetration seals to 

the nuclear industry. 

design change controls, product testing and installation, procurement of 

materials, and document controls. In addition, the inspectors reviewed 

Brand's program for implementing Part 21, "Reporting of Defects and 

Noncompliance," of Title 10 of the Code of Federal Requlations (10 CFR 

Part 21). 

Specific areas of review included, product design and 

During the inspection the inspectors examined procedures and representative 

records, observed relevant activities such as receipt inspections and product 

dedication, and conducted discussions and interviews with personnel. The 

inspectors' observations were discussed with the staff at the conclusion of 

the inspection, on November 9, 1995. 

Based upon its observations and evaluation, the team concluded that the 

silicone-based nuclear power plant fire barrier applications designed, tested, 

and installed by Brand are commercial grade items, as defined in 10 CFR 21.3. 

Although the silicone products supplied for nuclear plant fire barrier 

applications are not subject to the NRC's Part 21 regulation, it appears that 

Brand established and implemented a system which meets the intent of certain 

10 CFR Part 21 requirements, such as 10 CFR 21.21(b). Areas examined during 

the inspection and our conclusions are discussed in the enclosed inspection 

report. 

52 
R. Woodward -2- 

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

this letter and the enclosed inspection report will be placed in the NRC 

Public Document Room. If there are any questions concerning this inspection 

we will be pleased to discuss them with you. No response to this letter is 

requ i red. 

Sincerely , 

ORIGIW SIGNED BY 

Michael R. Johnson, Acting Chief 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Docket No. 99901020 

Enclosure: Inspection Report No. 99901020/95-201 

53 
ORGANIZATION: 

REPORT NO: 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF INSPECTION AND SUPPORT PROGRAMS 

INSPECTION CONDUCTED: 

LEAD INSPECTOR: 

OTHER INSPECTORS: 

REV1 EWED : 

APPROVED : 

Brand Fire Protection Services 

721 Racquet Club Drive 

Addision, IL 60101 

99901020/95-01 

Mr. Lance Simms, General Manager 

Designed, tested and installed rated fire barrier 

penetration seal assemblies in nuclear power 

electrical generating stations. The assemblies 

combined commercial grade silicone products and 

various proprietary additives developed to meet 

specific fire barriers penetration seal fire test 

criteria. 

November 6-9 ,,,.-I995 

( 

L -1 

~~ter-dss 

Speci a1 Inspection Section 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Armajit Singh, NRR 

Christofer Bajwa, NRR 

T.L. Tinkel, Brookhaven National Laboratory 

n 

h 

Venior" Inspection Secti n 

Di vi si on of Inspection and Support Programs 

Off i ce of Nucl ear Reactor Regul at i on 

Special Inspection Bran k! h 

r\ 

\\ A 

Special Inspection 

Division of Support Programs 

Office of Nuclear Reactor Regulation 

Enclosure 

54 
-

1.0 INSPECTION SUMMARY 

Based on the scope of this inspection, the team determined that the Brand Fire 

Protection Services (Brand) does not manufacture and supply any "basic 

components" as defined in Part 21, "Reporting of Defects and Noncompliance," 

of Title 10 of the Code of Federal Resulations (10 CFR Part 21). While some 

customers may contractually impose the requirements of 10 CFR Part 21 on Brand 

for their fire barrier penetration seal assemblies supplied for use in nuclear 

power plants, the inspectors verified that the products did not meet the 

definition of basic components as defined in 10 CFR 21.3. 

inspectors found that Brand has established and implemented a program which 

meets the intent of certain sections of 10 CFR Part 21, such as 10 CFR 

21.21(b). Brand personnel stated to the team that they will continue to 

implement the existing program for meeting the intent of 10 CFR Part 21. 

However, the 

2.0 STATUS OF PREVIOUS INSPECTION FINDINGS 

This was the first NRC staff inspection of this vendor; thus, there were no 

previous findings. 

3.0 INSPECTION FINDINGS AND OBSERVATIONS 

3.1 Entrance Meeting 

The entrance meeting was held on Monday, November 6, 1995. At this meeting, 

Mr. Singh, the acting NRC inspection team leader, discussed the scope of the 

inspection with the Brand staff. 

3.2 Background and Description of Facilities 

Following the Browns Ferry nuclear power plant fire, a new line of fire 

barrier penetration seals, using si1 icone based foam and elastomer materials, 

were developed to seal openings in nuclear power plant fire barriers. 

openings were required for various penetrations such as piping, cables, and 

heating, ventilation, and air conditioning (HVAC) ducts. 

The 

BISCO Construction, which later became Brand, was a major supplier of fire 

barrier penetration seal services from the late 1970s until the demand for the 

work diminished in the late 1980s. 

Brand General Manager, the company has provided fire barrier penetration 

sealing services to approximately 55 nuclear power plants in the United 

States. 

According to information provided by the 

The fire barrier penetration seal technology originally developed by BISCO and 

currently owned by Brand is based primarily on using silicone materials 

manufactured by Dow Corning. 

3.3 Penetration Seal Configurations and Sealing Materials 

Brand (previously BISCO) has tested and qualified different fire barrier 

penetration seal design configurations since the mid 1970s. A review of the 

2 
Brand Fire Test Report index indicates that over 100 tests were performed with 

the earliest tests completed in mid 1975 and the most recent performed in 

early 1992. 

The penetration seal design configurations vary depending on many factors 

including the type of fire barrier construction, the type of penetration, and 

the size of the opening being sealed. 

penetration seal conf igurat i ons i ncl ude : 

Examples of the various kinds of 

0 Penetrations for electrical cables 

0 Penetrations for electrical cable pipe conduits 

0 Penetrations for electrical cable trays 

Penetrations for piping and ducts 0 

Silicone products used by Brand in their fire barrier penetration design are 

designated as the SF-20 and SE foam and as SF-60, SF-lOOL, and SF-150L 

elastomer. 

3.3.1 Material Receipt Inspection 

A material receipt inspection was observed by tk NRC inspector. The Brand 

Quality Assurance (QA) manager performed the receipt inspection for packages 

of SF-60 elastomer that were received at the Brand warehouse. The inspector 

observed the QA inspection of the incoming packages. The shipment of SF-60 

was from a company that packaged the SF-60 ingredients for Brand into plastic 

tubes (Bisc-Kits) used for mixing and applying the SF-60 to form a seal. The 

QA inspector examined the boxes in which the SF-60 Bisc-Kits were shipped for 

any obvious signs of damage and then completed the receiving inspection 

checklist (RI-1). 

inspection was performed in accordance with the applicable procedures. 

The material was accepted by the QA inspector. The receipt 

3.3.2 Commercial Grade Dedication (CGD) 

The inspector observed the QA manager complete the commercial grade dedication 

(CGD) procedure (QCP-204) for the SF-60 elastomer, which certifies that the 

product meets the manufacturer’s specifications. 

taking samples of the material and measuring the density. 

allowed to cure for 24 hours and then checked for a complete cure by 

evaluating the color and structure. 

verification log (form QCT-2) for the dedication procedure. 

This procedure requires 

The samples are 

The QA manager completed the system 

The QA manager also completed a CGD for another material, SF-150NHY which is a 

high density radiation shield/penetration seal material. Both materials met 

all the applicable criteria for dedication and were certified for safetyrelated use. 

adequate. Results were properly recorded. 

The samples were taken according to procedure and controls were 

3.3.3 Materi a1 Storage/Shel f Life 

Storage of materials was also inspected. 

recommendations for shelf life of materials. 

receipt inspection process with an expiration date. 

Brand follows manufacturer’s 

Materials are tagged during the 

Brand will normally place 

3 

56 
a helf life on the Dr ducts th Y shiD of on year from the date that Brand 

ships them to the customer. 

issues were identified. 

Controls’in this area were adequate, and no 

3.3.4 Procurement Document Control 

A sample of Brand customer procurement documents were reviewed. The inspector 

reviewed some recent customer purchase orders (POs) and discussed with Brand 

personnel how nuclear customer orders were handled. Some of Brand’s customers 

contractually impose the provisions of 10 CFR Part 21, while others include 

those provisions on individual POs. Some customers do not invoke 

10 CFR Part 21 at all. Brand’s policy is to treat all orders from nuclear 

customers as invoking 10 CFR Part 21 notification requirements, whether it is 

stated explicitly or not, unless the material is to be used in a non-safety 

related application. However, the Brand QA manual states that if 10 CFR 

Part 21 is not explicitly invoked on a customer purchase order, then Brand 

does not have to comply with 10 CFR Part 21 reporting requirements. 

inspectors noted that Brand management’s intent to continue to notify a 1 

nuclear customers in accordance with 10 CFR Part 21 should be included n the 

applicable QA procedure. 

3.4 

Th 

Quality Assurance and Design Control 

A new quality assurance program was implemented on October 9, 1995. 

of the QA Manual, Revision A, addresses design control and includes the 

following requirements: 

Section C 

a. The adequacy of all designs must have been verified by testing or 

engi neeri ng anal ys i s . 

b. All designs shall be reviewed by a qualified individual other than 

the originator. 

c. The engineering manager or designate has the responsibility for 

the design and/or detailing of work necessary to meet the contract 

speci f i cations . 

d. The reviewing process shall include a qualified individual other 

than the individual who prepared the work. 

consist of spot checking calculations and reviewing the overall 

Brand design and/or detail work. 

The proces’ shall 

3.4.1 Methodology for Qualifying Penetration Seal Configurations 

Brand qualifies the design of a penetration seal configuration using one of 

two methods. 

a. The first method consists of subjecting the penetration seal 

design configuration to a fire endurance test. 

succzssful testing are used to qualify the particular parameters 

of the design. This method was used extensively during the 

devel opment of the fire barriers penetration seal s . 

The results of 

Brand 

4 

57 
maintains a library of over 100 fire penetration test results on 

file. 

b. The second method consists of performing an engineering analysis 

for the penetration seal design configuration requiring 

qualification in the field and establishing a correlation between 

the field penetration and tested penetration configurations on 

file. 

The engineering analysis method 

permits Brand to use a combinat 

multiple prior tests to qualify 

performing additional testing. 

Engineering analysis for qual if 

the following: 

for qualifying penetration seal configurations 

on of engineering judgment and results from 

new or modified design configurations without 

cation consists of but it is not limited to 

a. Analyzing reports of pertinent past tests performed by Brand. 

b. Identifying qualitative and quantitative information for selected 

design parameters that represent the candidate design 

configuration and are covered by prior testing. 

c. Documenting a range of test parameters from the test reports on a 

Testing Matrix Sheet. 

d. Selecting the best results for the particular design and test 

parameters and documenting this information on the summary line of 

the Testing Matrix Sheet. 

e. Using the summary line information on the Testing Matrix Sheet as 

the basis for qualifying the design parameters for the candidate 

penetration seal design configuration. 

The inspectors verified and evaluated the various design parameters identified 

and used by Brand for qualification using engineering analysis. 

technical rationale explaining how to properly extrapolate test results was 

not fully documented for the maximum opening area and maximum free area 

parameters. 

However, the 

All parameters are limited by tested configurations; however, there is no 

limitation on the length and width parameters that produce an acceptable range 

of areas. 

assessment. 

appropriate limitations as needed. 

This appeared to be a weakness in the maximum free area parameter 

Brand management agreed to review the issue and establish 

The inspector noted that another parameter, the total cross sectional area of 

the penetrants (number of penetrants x cross sectional area of penetrants), is 

not identified as a design parameter for qualification using engineering 

analysis. Brand management agreed with the inspector that the cross sectional 

area of the penetrants may in some instances affect the temperature of the 

penetrants on the unexposed side and thus the temperature of the seal material 

5 

58 
on the unexposed side. 

when this parameter does or does not have to be considered. 

Therefore, technical basis would be provided to show 

The inspector assessed the quality of completed engineering analyses by 

reviewing the following resultant Fire Barrier Seal Design Test Reports: 

Test report 748-49 dated July 9, 1981. 

a basis for qualifying three fire barrier penetration seal design 

configurations (Details 118, 129, and 149) for a power plant. The test 

report indicated that the limiting end point temperatures for the 

unexposed surface defined by ANI were met but the limiting end point 

temperatures defined by ASTM E-119 and IEEE 634-1978 were reached at 2 

hours and 45 minutes into the fire test and slightly exceeded the 

limiting temperature at 3 hours for both the 8-inch and 9-inch diameter 

penetration. Further, the 9-inch penetration failed the hose stream 

test. This matter was discussed with the Brand Engineering Manager who 

stated that the information in the test report for the 9-inch 

penetration was not used for the qualification basis. 

Engineering Manager also provided a document (Specification C1027, Rev. 

6) that indicated the subject plant was committed to ANI and IEEE 634 

but not to ASTM E-119. The inspector verified that IEEE-634 does not 

contain the same temperature limit (250" F plus ambient) as ASTM E-119 

for the unexposed surface; thus, the test results satisfied IEEE-634. 

This report was recently used as 

The Brand 

Test report 748-78 

for qual i fying one 

(Detail 109) for a 

dated May 23, 1982. 

fire barrier penetration seal design configuration 

power plant. 

that the test passed the hose stream test specified by IEEE-634, but 

failed the hose stream test specified in ASTM E-119. This matter was 

discussed with the Brand Engineering Manager who provided a document 

(Specification C1027, Rev. 6) that indicated that power plant was 

committed to ANI and IEEE 634, but not to ASTM E-119. 

This report was used as a basis 

A review of the test report indicated 

In conclusion, the review of actual design configurations did not identify any 

design issues. However, the implementation of the engineering analyses to 

qualify tested fire barrier penetration seals on file, for installation into 

fire barrier penetrations of various configuration, partially relies on 

informal engineering judgment. This could be improved by formally documenting 

and controlling such aspects of the engineering analysis methods in a Brand 

engineering standard or procedure. 

3.5 Exit Meeting 

The exit meeting was conducted on November 9, 1995 by the NRC team leader 

prior to the team's departure from the Brand facility. 

6 
Mr. Gary L. Stouffer, President 

Modumend, Inc . 2640-A Lavery Court 

Newbury Park, CA 91320-1589 

February 12, 1996 

SUBJECT: NRC INSPECTION NO. 99901291/95-01 

Dear Mr. Stouffer: 

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

(NRC) inspection of Modumend, Inc. at Newbury Park, California, conducted by 

Messrs. R.C. Wilson and B.H. Rogers of this office on December 5 through 7, 

1995. The purpose of the inspection was to review activities conducted under 

your 10 CFR Part 50, Appendix B, quality assurance program and 10 CFR Part 21 

reporting program. 

and records, interviews with personnel, and observations by the inspectors. 

The inspection consisted of an examination of procedures 

The NRC inspectors found that the implementation of your quality assurance 

program failed to meet certain NRC requirements. Specifically, independent 

veri f i cati on of f i nal acceptance test i ng of repaired nuclear safety-re1 ated 

power suppl ies was not performed. 

testing, and inspection activities. The specific findings and reference to 

the pertinent requirements are identified in the enclosures to this letter. 

Only one person performed the repair, 

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

statement in accordance with the instructions specified in the enclosed Notice 

of Nonconformance. 

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

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

Room. 

Sincerely, 

ORIGINAL SIGNED BY GREGORY C. CWaINA FOR: 

Michael R. Johnson, Acting Chief 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Docket No. 99901291 

Enclosures: 1. Notice of Nonconformance 

2. Inspection Report 99901291/95-01 

60 
NOTICE OF NONCONFORMANCE 

Modumend, Incorporated 

Newbury Park, Cal i forni a 

Docket No. : 99901291 

Based on the results of an inspection conducted on December 5 through 7, 1995, 

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

with NRC requirements as described below. 

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

Drawings," requires, in part, that activities affecting quality shall be 

accomplished in accordance with written procedures. 

Criterion X of Appendix B to 10 CFR Part 50, "Inspection", requires, in part, 

that activities affecting quality shall be verified by individuals other than 

those who performed the activities. 

Section 5.7, "Final Inspection and Test," of the Modumend Quality Assurance 

Manual, Revision A, dated June 27, 1994, states that final inspection and test 

will be performed by a qualified individual other than the person performing 

the repair activity. 

Contrary to the above, on nuclear safety-related purchase orders including 

Rochester Gas and Electric Corporation order number NQ-14588-C-JW dated 

April 18, 1995, covering repair of a Westinghouse power supply, ac/dc model 

RS15N21, Serial Number AC9005, for the Ginna nuclear power plant, the Modumend 

president stated that he, alone, performed all of the repair and testing work. 

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

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

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

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

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

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

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

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

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

actions and preventive measures were or will be completed. 

Dated at Rockville, Maryland 

this 12th day of February, 1996 

Enclosure 1 

61 
T 

REPORT NO.: 

ORGANIZATION: 

ORGANIZATIONAL 

CONTACT : 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF INSPECTION AND SUPPORT PROGRAMS 

99901291/95-01 

Modumend, Inc. 

2640-A Lavery Court 

Newbury Park, California 91320-1589 

Stan Zabaglo, Vice President 

805-499-6464 

NUCLEAR INDUSTRY 

ACT1 VITY : 

INSPECTION DATES: 

INSPECTORS: 

OTHER INSPECTORS: 

REVIEWED BY: 

APPROVED BY: 

Repair of various instrumentation 

power suppl i es 

December 5 through 7, 1995 

7 Z/iZhG 

Richard C. Wilson, Senior Ensineer Date 

Vendor Inspection ~Section 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

- 

Billy H. Rogers, Reactor Engineer 

//5$PT 

Greqor C Cwal ina, Section Chief Date 

VendorvInYpection Section 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

. Johnson, Act i nq Chi ef , 

2fM4 

Date 

Speci'afl/Inspect i on ~ BranchDivision of Inspection and Support Programs 

Office of Nucl ear Reactor Regul at i on 

y 

Enclosure 2 
1 sco 

Modumend , 

E OF INSPECTION: 

Inc. specializes in repairing and testing power supplies, often from 

computer and instrumentation systems where original design power supplies can 

no longer be procured as new components. Refurbishment activities, such as, 

preventive replacement of aging electrolytic capacitors are also performed. 

Modumend’s business, which is rapidly growing, includes military, government, 

and commercial customers, as well as, commercial nuclear power plants. The 

number of employees has grown from seven to nine in the last few years. 

Modumend has also done business as Uptime Depot Repair, Inc. 

The NRC inspectors reviewed the implementation of selected portions of 

Modumend’s qua? ity assurance (QA) program, and reviewed Modumend’s 10 CFR 

Part 21 program. The inspection bases were 10 CFR Part 50, Appendix By and 

10 CFR Part 21. 

1.1 Non-Cited Violation 

Contrary to 10 CFR 21.21, which states that corporations subject to the 

regulations must adopt appropriate procedures to include specific reporting 

requirements, Modumend did not have a procedure in place, which included the 

requirements of 10 CFR 21.21 (non-cited violation; see Section 3.6 of this 

inspection report). 

1.2 Nonconformance 99901291/95-01-01 

Contrary to Criteria V a 

imposed on Modumend by 1 

independent veri f i cation 

power supplies (see Sect 

2 STATUS OF PREVIOUS 

d X of Appendix B to 10 CFR Part 50, which were 

censee purchase orders, Modumend did not perform 

of final acceptance testing of nucl ear safety-re1 ated 

on 3.3 of this inspection report). 

INSPECTION FINDINGS 

There were no previous NRC inspections of Modumend. 

3 INSPECTION FINDINGS AND OTHER COMMENTS 

3.1 Entrance and Exit Meetinqs 

In the entrance meeting on December 5, 1995, the NRC inspectors discussed the 

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

interfaces with Modumend management. In the exit meeting on December 7, 1995, 

the inspectors discussed their findings and concerns with Modumend management. 

3.2 Oualitv Assurance Prosram and Orsanization 

The inspectors selectively reviewed the Modumend Qual ity Control Manual (QAM) , Revision A, dated June 27, 1994, and its implementation. The QAM stated that 

Modurnend’s quality assurance (QA) program was intended to meet the 

requirements of 10 CFR Part 50 Appendix B as it applied to the repair of power 

2 

63 
supplies. 

and customer purchase order (PO) requirements to all work performed. 

Modumend management stated that they applied the QAM requirements 

The Modurnend QAM included an organization chart which identified the 

management functional responsibilities. Due to the small size of Modumend, 

the two company owners/officers performed several functions. The president 

also served as the chief engineer and the quality manager, and the vice 

president also served as the general manager. The inspectors discussed the 

division of responsibilities with the Modumend officers, reviewed 

documentation, and observed examples of work activities, and concluded that 

the organizational structure and its implementation were satisfactory, with 

the exception of the lack of independent verification of fina acceptance 

testing discussed in Section 3.3 of this report. 

Document control was specified by Modurnend's QAM Section 3.9, "Documents," 

which stated that the document control program applied to the QAM, procurement 

documents, operating procedures, inspection and test procedur s, and equipment 

cal i bration procedures. Control led documents were reviewed and approved prior 

to issue, and distributed to employees with associated responsibilities; 

obsolete copies were removed from circulation or marked as inactive. The 

inspectors reviewed several documents and their revisions including the QAM 

and "Equipment Calibration Procedure," Revision A, dated June 27, 1994, and 

determined that they were appropriately marked with revision, date, author 

signature, and reviewer signature. 

Modumend contracted with an outside calibration service to calibrate the 

equipment used to measure power supply performance during final acceptance 

testing. Modumend dedicated the calibration services for safety-related use 

based on annual audits of the supplier by the president/chief engineer. The 

inspectors reviewed the audit reports and purchase orders, and determined that 

the dedication of commercial grade calibration services was adequate. 

Modumend QAM Section 5.0, "Procedure," established the requirements for 

identifying and control1 ing received power supplies. 

that a work order and unit tag be prepared for each received power supply. An 

equipment tag was attached to each item, or kept with it if attachment was not 

feasible. The tag identified the Modumend work order number, equipment 

status, due date, customer, customer purchase order number, location, 

quantity, part number, and serial number. The work order included all 

information listed on the equipment tag, and also listed the tasks to be 

performed such as assembly, repair, and test, the times and dates the tasks 

were accomplished, the technician performing the tasks, the parts replaced, 

and comments. 

received that did not have serial numbers (e.g, 1 of 6). The inspectors 

reviewed various work orders, and verified that the power supplies were tagged 

as required and stored in the designated locations. 

Section 5.1.5 required 

Modumend assigned an in-house identifier to all power supplies 

Modumend performed no design activities. Work typically consisted of testing 

the as-received power suppl ies; rep1 acing defective amponents and/or aged 

electrolytic capacitors with components equivalent in function and tolerance 

to the originally installed capacitors; burning in the repaired power supply; 

and performing final functional performance tests to verify that the 

3 
manufacturer's performance specifications were met. 

purchased as commercial grade items. Modumend verified the quality of the 

replacement components as part of the burn in and testing of the repaired 

power supply. The type of repair and identification of replacement parts were 

documented on the Modumend Trouble and Failure Report supplied to the 

customer. 

re1 ated to design--sel ecting rep1 acement components--was adequately 

control 1 ed. 

Replacement parts were 

The inspectors concluded that the Modumend activity most close'ly 

Modumend maintained an inspection system to determine the performance of the 

received power supplies and the quality of the work performed during various 

points in the repair or refurbishment process. Incoming inspection by the 

technician performing the repair verified the as-received condition of the 

equipment and the existence of any customer-identified conditions. 

inspections and tests were conducted informally by the technician as required 

to verify the work activities. Any non-conforming replacement parts 

identified during the in-process inspection and tests were replaced by the 

technician. A pre-test was performed on each power supply, prior to 

reassembly, to verify operability (formal in-process inspection and pre-tests 

were also performed if contractually required). To assist the technicians in 

the repairs, inspections, and tests, Modumend maintained a large 1 ibrary 

(several f i 1 e cabinets) of technical information re1 ated to the power suppl ies 

which they repaired. The library included catalog information, technical 

manual s and bull et i ns , and schemat i cs . 

In-process 

The inspectors observed the Modumend process for packing refurbished power 

supplies after completion of the final inspection and test. 

recently installed a new system using quick setting foam formed in the carton 

to provide protection for the shipped power supplies. 

additional packaging material was also used. No concerns were noted. 

Modumend had 

Double boxing with 

3.3 Final Inspection and Testing 

Modumend procedures covered final inspection and testing performed to verify 

operability of the power supply, under load, to the manufacturer's 

specifications and to any customer requirements. 

inspection and test were documented on a Power Supply Data Sheet which 

i ncl uded nominal vol tages , currents, 1 oad regul at i on, 1 i ne regul at i on, ri ppl e , and noise. 

part and serial numbers, technician, date, and instruments used. 

The results of the final 

In addition, the Power Supply Data Sheet listed the power supply 

Modumend's president told the inspectors that he personally performed all 

repair and testing of nuclear safety-re1 ated power suppl i es. The inspectors 

noted that this practice was not in conformance with Section 5.7, "Final 

Inspection and Test," of the Modumend QAM, which specified that final 

inspection and test must be performed by a qualified individual other than the 

person performing the repair activity. It further conflicts with Criterion V 

of Appendix By "Instructions, Procedures, and Drawings," which requires that 

activities affecting quality shall be accomplished in accordance with written 

procedures, and Criterion X, "Inspection", which requires that activities 

affecting quality shall be verified by individuals other than those who 

4 

65 
performed the activities. 

verification of final acceptance testing constitutes Nonconformance 

No. 99901291/95-01-01. 

Modurnend’s failure to perform independent 

Modurnend’s president, who also served as quality manager and chief engineer, 

had over 30 years experience in electrical engineering. He explained to the 

inspectors that he performed all of the nuclear safety-related work because 

his experience and expertise exceeded the capabilities of the technicians. He 

felt that the most important work should be performed by him personally. 

There was no evident intention to conceal this nonconformance. The inspectors 

noted that during a mandatory hold point on safety-related repair PO 

No. 565449 from Wol f Creek Nucl ear Operat i ng Corporation , a 1 i censee auditor 

observed testing performed by the president and discussed it with him. 

The inspectors discussed this concern with Modurnend’s management. 

inspectors indicated that observation of the final acceptance testing by a 

qualified Modumend or licensee individual, or independent testing by a second 

individual, would appear to resolve the concern. 

The 

The inspectors observed a demonstration of final acceptance testing of a power 

supply for a non-nuclear customer, using calibrated test equipment, to confirm 

that this operation did adequately verify that the power supply met the 

manufacturer’s performance specifications. The test procedure used was ATP110 

dated June 15, 1995. The same procedure was used on June 23, 1995, for 

testing a Westinghouse Model 1686E16H10, ac/dc Model RS15N21, Serial 

No. AC9005, power supply under Rochester Gas & Electric Corporation PO 

No. NQ-14588-C-JW for the Ginna nuclear power plant. 

supply was a 12 amp Lambda supply, rather than the 21 amp unit; the testing 

was otherwise the same. The demonstration took about 15 minutes including 

brief questions and answers. 

power supplies, testing could require as long as two hours. The inspectors 

had no concerns with the demonstration of final acceptance testing. 

The demonstration power 

The president commented that for mu1 tiple-output 

3.4 Review of Safetv-Related Purchase Orders 

Modumend maintained a separate file cabinet for all POs from NRC licensees. 

Modumend provided the inspectors with a list of the 18 licensees included. 

The inspectors scanned each file to divide the included POs into three groups: 

(1) those that invoked requirements such as Appendix B to 10 CFR Part 50 or 

10 CFR Part 21 that appeared to make them nuclear safety-related; (2) those 

that were clearly not safety-related, such as those for fossil plants; and 

(3) commercial grade POs for nuclear plants covering work that may or may not 

have been intended for subsequent dedication for safety-related use. 

first group is addressed in this section of the inspection report. The second 

group was not reviewed beyond the determination that the work had no nuclear 

safety connection. Most of the POs in the third group, including all that 

gave indication of possible subsequent dedication activity, were reviewed as 

reported in Section 3.5 of this inspection report. 

The 

The following licensees had listed Modumend as an approved Appendix 8 supplier 

and placed safety-re1 ated POs with Modumend: 

5 
Alabama Power Company Farley Nuclear Plant, based on an October 1992 

audit by Southern Nucl ear Operating Company. 

Rochester Gas & Electric Corporation (RG&E) Ginna Station, based on a 

November 1993 audit of Modumend by RG&E. 

Virginia Power Surry Power Station, based on review of the RG&E audit 

report per June 1994 Virginia Power letter to Modumend. 

In addition, the Wolf Creek Nuclear Operating Corporation placed two safetyrelated POs for the Wolf Creek Generating Station with mandatory hold points 

to permit evaluation of Modumend’s activities. 

Farlev: 

Company, performed a commercial grade survey of Modumend on November 30, 1992. 

The inspector’s review of the survey report showed that no deficiencies were 

identified and no conditions were placed on procurement of services from 

Modumend. The report also stated that a commercial grade item checklist 

prepared by a licensee group was used, but the completed checklist was not 

available at Modumend. The Modumend files did not contain any basis for 

subsequent 1995 A1 abama Power Company safety-re1 ated POs for the Farley 

Nuclear Plant (FNP), nor were they required to. 

The Southern Nuclear Operating Company, as agent for Alabama Power 

Alabama Power PO No. QP950244 dated February 16, 1995, covered repair of a 

high voltage power supply, FNP stock number 38258, serial number A211011. 

Although not reviewed in detail by the inspector, the PO invoked 

10 CFR Part 21 and Appendix B to 10 CFR Part 50, and stated that the items 

were safety-related. 

dated September 27, 1995, which was also safety-related and invoked Part 21 

and Appendix B. It covered repair of two power supplies, Westinghouse part 

number 2374A07601, FNP stock number 236, serial numbers 171 and 236, and 

substitution of new capacitors supplied by the licensee. The PO required 

Modumend to supply a repair report listing any parts or material repaired or 

replaced, and a certificate of conformance to the requirements of the PO. 

submittals for review or approval, and no inspection hold points, were 

specified. 

27, 1994 (still current during the NRC inspection) had been reviewed and 

accepted by the purchaser, and required submittal of all subsequent revisions. 

The inspector did review Alabama Power PO No. QP950810 

No 

The PO stated that Revision A of Modumend’s QA Program dated June 

As was typical of all POs reviewed by the inspectors, Modumend provided the 

following documentation for each power supply: a Trouble and Failure Report, 

detailing part changes; a Certificate of Conformance, including a description 

of work performed; and a Test Data Sheet, showing acceptable and actual values 

for all properties tested and identifying the test equipment used (with the 

signature and QA stamp of Modumend’s president). The file did not contain any 

evidence of licensee surveillances. 

Ginna: 

performed on November 2-4 1993, and an internal audit of Modumend on 

June 27-28, 1994. The NRC inspector reviewed the reports of these audits and 

re1 ated correspondence. The RG&E audi t report ident i f i ed one finding , that 

Modumend had not conducted internal audits. Modumend subsequently contracted 

RG&E listed Modumend as a safety-grade supplier based on an RG&E audit 

6 
with the RG&E auditor to lead their internal audit, with the Modumend vice 

president as a team member. The internal audit produced no findings, and its 

performance closed the only finding from the RG&E audit. 

Neither the RG&E audit nor the internal audit identified the Nonconformance 

identified in Section 3.3 of this inspection report. With regard to that 

concern, the RG&E audit report stated on page 4 that if the Modumend president "... performs the actual testing, he delegates the quality function of data 

review to one of the two technicians." A filled out industry group audit 

checklist that supported the audit report stated on page 31 that "If the 

president performs the test, a technician reviews the data for completeness 

and accuracy." The internal audit report, on page 7, stated that "After a 

unit is repaired, either the other technician, or Quality Manager perform 

[sic] the final test on the power supply." 

The inspectors observed that page 31 of the checklist also stated that 

Modumend had not yet accepted any safety-related POs. Nevertheless, no 

provision was identified for surveillance of safety-related work, and the 

Modumend files for subsequent safety-re1 ated POs from RG&E showed no evidence 

of licensee surveillance. 

RG&E PO No. NQ-14588-C-JW dated April 18, 1995, covered repair of a 

Westinghouse power supply, Model 1686E16H10, ac/dc Model RS15N21, Serial 

No. AC9005, for the Ginna Station. The PO invoked 10 CFR Part 21 and 

Appendix B to 10 CFR Part 50. 

audit of Modumend were also specified, relating to identification of 

replacement parts, use of a documented test procedure, and QA approval of test 

data sheets containing acceptance criteria. Additional requirements covered 

an appropriate certificate of conformance, submittal of documentation, and 

packaging. The PO did not, however, require submittal of the test procedure 

to RG&E. The documentation submitted by Modumend appeared to satisfy the PO 

requirements. 

Four points identified during the 1993 RG&E 

Surrv: 

Depot Repair, Inc.) dated June 20, 1994, stated that they were approved to 

repair power supplies for the utility's nuclear power plants based on review 

of the RG&E audit. 

cooperation with members of a licensee group. A Virginia Power letter dated 

September 29, 1994, documented review and approval of four Modumend test 

procedures. 

Virginia Power's letter to Modumend (known at that time as Uptime 

The letter also expressed appreciation for Modumend's 

Virginia Power PO No. BNT 504467 dated July 7, 1995, was a blanket order for 

the Surry Power Station patterned after PO No. BNT 473149 dated August 3, 

1994. The inspector reviewed the files for the blanket order, which stated 

that the materials and services are nuclear safety related, and invoked 

10 CFR Part 21 and Appendix B to 10 CFR Part 50. 

and detailed information to permit evaluation of all substitute parts, and 

defined unit prices for repair of several named power supply types. 

Modumend Failure Analysis Report, Test Data Sheet, al,d Certificate of 

Conformance were required. The blanket order provided for oral authorization 

of work under the blanket order, with no written confirmation. The inspector 

reviewed documents dated October 28-30, 1995, for repair of an ac/dc power 

The PO required notification 

The 

7 

68 
supply, part number RT 151, serial number AD 0201, tracked by Modumend Return 

Authorization No. 2272-02. This power supply type is not one of the examples 

listed in the blanket order, but the September 1994 letter listed the test 

procedure (RT151-118) as reviewed and approved. 

Wolf Creek: 

points intended to provide the opportunity for QA surveillance of Modurnend’s 

safety-related activities. The inspector concentrated on the PO under which 

the first work was performed: PO No. 565449 dated August 22, 1995, covering 

repair of one compensating voltage power supply, Westinghouse part number 

6051D35601, serial number 0192. The PO invoked 10 CFR Part 21, Modumend’s QA 

Manual, and Wolf Creek Quality Program Requirements No. NPRDS-MC80, Revision 

0; Appendix B to 10 CFR Part 50 was not specifically called out, but the PO 

was labeled safety-related. 

of Conformance, and specified that the test procedure should be submitted and 

approved prior to starting work. 

Wolf Creek placed two safety-related POs with mandatory hold 

The PO required a Repair Report and a Certificate 

A Wolf Creek representative performed a surveillance on October 3-4, 1995, for 

the initial testing, repair, and fiwl acceptance testing of the power supply 

on PO 565449. 

specifications. 

advised the Wolf Creek representative to contact Wes inghouse regarding the 

applicability of the performance specifications they had provided to Wolf 

Creek; he stated that the specifications covered a t ghtly regulated supply 

and were inappropriate for the supply being repaired 

Modumend repeated the acceptance testing using newly received specifications, 

and the unit passed. 

The repaired supply did not meet the manufacturer’s 

The Modumend president told the NRC inspector that he had 

In late October, 

The Wolf Creek surveillance report identified a noncompliance concerning 

Modumend’s lack of an Appendix B program to dedicate the commercial grade 

parts used for repairing power supplies, and suggested that Modumend consider 

providing their repair services as commercial grade rather than safety-grade. 

In view of Modumend’s stated intention to resolve the Nonconformance 

identified in Section 3.3 of this inspection report, the NRC inspectors 

advised Modumend that they believed that the parts need not be individually 

dedicated. 

inspectors considered that independently performed or verified final 

acceptance testing would be sufficient to assure the quality of the repaired 

power suppl ies. 

Within the framework of Modumend’s QA program and work scope, the 

Conclusion: Based on review of the Alabama Power, RG&E, and Virginia Power 

files for safety-re1 ated work at Modumend, the NRC inspectors concluded that 

these licensees did not observe performance of any safety-related work at 

Modumend. Thus they fa’iled to identify that no independent verification of 

activities affecting safety was performed, as described in the Nonconformance 

in Section 3.3 of this inspection report. Wolf Creek appeared to be 

proceeding cautiously, and was close to establishing Modumend as an approved 

supplier. Except for the NRC-identified Nonconformance, the inspectors found 

no other evidence that Modumend failed to meet the requirements of POs from 

these licensees. 

8 

69 
3.5 Review of Commercial Grade Purchase Orders 

The inspectors selected and reviewed several commerci a1 grade purchase orders 

from NRC licensees for the repair or refurbishment of power supplies. 

commercial grade POs stated that the power supplies were to be dedicated for 

safety-related application by the licensee, while others did not state an 

intended use; the latter POs generally contained fewer requirements for 

documentation, inspection hold points, and the like. None of these POs 

invoked 10 CFR Part 21 or Appendix B to 10 CFR Part 50. 

Some 

Calvert C1 iffs: Baltimore Gas and Electric Company (BG&E) commercial grade PO 

No. L01522NP, dated April 23, 1993, covered the repair of three Lambda LMEE-48 

power supplies and specified the completion of a BG&E Vendor Report form for 

each power supply. Modumend completed the repairs on April 30, 1993, and 

identified on the Vendor Reports the repairs that had been accomplished-- 

replacement of a drive transistor and the associated power resistor and all 

electrolytic capacitors. This was an example of a commercial grade PO where 

the inspectors found no evidence of special requirements intended to support 

any subsequent dedication efforts. 

Crvstal River: Florida Power Corporation (FPC) Crystal River Unit 3, 

commercial grade PO No. R873830K, dated March 7, 1995, covered the repair of 

one Lambda LXS-E-15R power supply. 

work in accordance with the June 27, 1994, revision of the QAM, prepare a test 

procedure and submit it to FPC prior to performance of the test, and to 

provide a supplier’s deviation report to document the basis for equivalency of 

replacement parts. Modumend completed the repair on May 26, 1995, and 

provided FPC a Trouble and Failure Report/Certificate of Compliance which 

documented that Modumend had replaced all electrolytic capacitors (1 isting the 

original as-found parts and the final installed parts), a Test Data sheet 

which documented the Final Test performance of the power supply, and a 

Supplier Deviation Request which adequately justified all parts replacements. 

The PO also required use of a customer-approved test procedure, and listed 

several performance characteristics to be tested. The inspectors concluded 

that the information required and provided could support a dedication effort 

by the licensee or a third party, if sufficient additional activities were 

performed. 

The PO required Modumend to perform the 

HoDe Creek: Public Service Electric & Gas (PSE&G) commercial grade PO 

No. Pl-429109, dated June 10, 1993, covered the repair of five Lambda 

LGS-6-24 OV-R-8018 power supplies. 

PSE&G with the repair records. 

inspection, final documentation review prior to shipment, and electrical test, 

for commercial grade verifications in accordance with EPRI NP-5652 method 3 

(indicating intended dedication by PSE&G). In addition, the PO indicated that 

PSE&G would verify critical characteristics including part number, dimensions, 

weight, output voltage, ripple, and noise. Modumend completed the repairs on 

June 24, 1993, and provided PSE&G a Trouble and Failure Report/Certificate of 

Compl i ance which documented that Modumend had rep1 xed defective capacitors. 

Again the inspectors concluded that subsequent dedication could be performed 

with sufficient supporting activities . 

The PO required that Modumend provide 

It defined mandatory hold points at final 

9 

70 
Comanche Peak: 

survey of Modumend on August 2-3, 1993, for the Comanche Peak Steam Electric 

Station. 

re1 ated to it . Four def i ci enci es were i dent i f i ed and resol ved . TUE concl uded 

that Modumend's QA program adequately controlled the critical characteristics 

identified by TUE for purposes of dedication by TUE for safety-related 

applications at CPSES, subject to certain PO requirements summarized below. 

Texas Utilities Electric (TUE) performed a commercial grade 

The NRC inspectors reviewed the survey report and correspondence 

TUE commercial grade PO No. S 0144232 6S4, dated December 8, 1994, covered the 

repair of one Westinghouse UPM-44KW power supply. The PO specified the 

required testing parameters including input and output voltages, ripple, and 

overvoltage protection setting, and required submittal for approval of a test 

procedure specifying acceptance 1 imi ts. The PO a1 so required that Modumend 

perform the work under the most recent revision of the Modumend QAM and 

provide any subsequent revisions to TUE, provided requirements on the 

selection of replacement components, and specified the Modumend Final Test and 

Inspection procedure to be used. Modumend completed the repair on February 

21, 1995, and provided TUE a Trouble and Failure Report/Certificate of 

Compliance covering replacement of all electrolytic capacitors and testing of 

the power supply in accordance with the specified test procedure. 

addition, Modumend provided TUE a list of replaced parts and the applicable 

test data sheet. Again the inspectors concluded that subsequent dedication 

could be performed with sufficient supporting activities. 

In 

Conclusion: 

of the reviewed commercial grade POs, including requirements intended to 

support dedication activities by the licensee or a third party, with one 

exception. The Nonconformance described in Section 3.3 of this inspection 

report would prevent taking credit for Modumend's activities in subsequent 

dedication activities unless the licensee performed direct surveillance of 

appropriate activities. 

The inspectors concluded that Modumend had met the requirements 

3.6 10 CFR Part 21 Procrram 

10 CFR Part 21 requires entities subject to the regulation to post copies of 

Section 206 of the Energy Reorganization Act, 10 CFR Part 21, and procedures 

adopted pursuant to 10 CFR Part 21 (or specify where the procedures can be 

found). The inspectors determined that, instead of a procedure, Modumend had 

posted an internal memorandum, "Nuclear Regulatory Commission 10 CFR Part 21 

Requirements," dated November 3, 1993. The memorandum discussed Modumend 

employee responsibilities to identify potential unsafe conditions to 

management; it further stated that Modumend management was responsible for 

notifying the NRC or the customer, and that a Modumend employee could contact 

the NRC if the employee believed that the requirements of 10 CFR Part 21 were 

not being implemented. Modumend personnel stated that the internal memorandum 

was the Modumend document used to address and disseminate the information and 

requirements of 10 CFR Part 21. However, the inspectors noted that the 

memorandum did not satisfy 10 CFR 21.21, which requires that corporations 

subject to the regulation adopt appropriate proCcdures to evaluate and report 

deviations and failures to comply within prescribed time limits. 

10 

71 
Discussion with the Modumend president, who also served as quality manager, 

confirmed that Modumend did not have a procedure for the items required to be 

proceduralized by 10 CFR Part 21. 

10 CFR Part 21 implementing procedure was in place, the president and vice 

president were familiar with the regulations and their responsibilities. The 

inspectors also discussed potential deviations related to Modurnend's work 

scope and determined that no deviation requiring an evaluation had ever been 

identified by Modumend. Thus, Modumend had not failed to follow any of the 

10 CFR 21.21 requirements required to be proceduralized. 

Further discussion showed that, although no 

During the course of the inspection, Modumend prepared and issued Operating 

Procedure OPlOO , "Operat i ng Procedure for Not i f i cat i on and Eva1 uat i on 

10 CFR Part 21," dated December 6, 1995. OPlOO contained the items required 

to be covered by procedure in 10 CFR 21.21 and, in addition, included 

definitions and specific actions for Modumend employees to follow regarding 

the requirements of 10 CFR Part 21. The failure to proceduralize the 

requirements specified in 10 CFR 21.21 is a violation of minor significance 

and is being treated as a non-cited violation. 

4 PERSONNEL CONTACTED 

t * G.L. Stouffer, President 

t * S. Zabaglo, Vice President 

S. Weizer, Sales/Customer Service 

+ Attended the entrance meeting on December 5, 1995 * Attended the exit meeting on December 7, 1995 

11 

72 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 2OSHlOO1 

.- 

' ++**+ March 1, 1996 

Dr. Robert C. Mecredy 

Vice President - Nuclear Operations 

Rochester Gas & Electric Corporation 

89 East Avenue 

Rochester, NY 14649 

Dear Dr. Mecredy: 

SUBJECT: INSPECTION OF ROCHESTER GAS AND ELECTRIC'S MONITORING OF ITS 

VENDORS' 

NUCLEAR POWER PLANT (50-244/96-201) 

QUALITY CONTROL FOR SAFETY-RELATED PRODUCTS FOR THE GINNA 

We are forwarding the report of the h:!!-Ject inspection performed January 

22-26, 1996, involving activities authorized by Operating License Nos. DPR-18 

for the Ginna Nuclear Power Plant. 

staff from the Special Inspection Branch of the Office of Nuclear Reactor 

Regulation conducted the inspection. An exit meeting was held on January 26, 

1996, during which we discussed our findings with Mr. Richard Watts and other 

members of your staff. 

The Nuclear Regulatory Commission (NRC) 

The inspection was performed to assess the adequacy and effectiveness of your 

vendor oversight program for monitoring the quality control by vendors 

supplying safety-related products to Ginna, consistent with the importance, 

complexity, and quantity of the purchased products or services. 

inspectors focused on the adequacy of your qualification and monitoring of 

nine selected vendors supplying safety-related products to Ginna. 

The two 

We assessed the attributes and implementation of your vendor oversight program 

against: 1) the criteria of 10 CFR Part 50, Appendix B, Criterion VII, 

"Control of Purchased Material, Equipment, and Services," and 2) Regulatory 

Guide 1.144, "Auditing of Quality Assurance Programs for Nuclear Power 

Plants." 

inspection report. 

In general, an effective vendor oversight program was implemented. Program 

attributes included appropriate QA organization, monitoring of vendors, 

maintenance of a qualified suppliers list, and qualification and training of 

QA staff. 

adequate and appropriately documented. 

One concern was identified regarding your assessment of the vendor Modumend 

Incorporated (Modumend). In November 1993, your procurement QA team leader 

found that Modumend had not done an internal audit required by Appendix B, 

Criterion XVIII. During that time, QA management encouraged their staff to 

generate revenue for RG&E. 

approval, the QA team leader did the internal audit for the vendor for a fee, 

independently of RG&E. 

Details of the inspection are discussed in the enclosed copy of our 

The basis for qualification of selected vendors appeared to be 

In June 1994, with his immediate manager's 

As a result, RG&E's QA team leader performed 

73 
R. Mecredy -2- 

Modurnend’s quality function, instead of assessing Modumend’s performance of 

the function. It is our understanding that, in the future, current management 

will .avoid any perception of a conflict of interest with a vendor, will not 

suggest to their engineers to perform such an audit, and have instructed QA 

engineers not to perform work for a company that QA was responsible to assess, 

either as independent consultants or as RG&E employees. 

In accordance with Section 2.790 of Title 10 of the Code of Federal 

Re~u7ations (10 CFR 2.790), a copy of this letter and its enclosures will be 

placed in the NRC Public Document Room. 

No response is required to this letter. 

this inspection, we will be pleased to discuss them with you. 

If you have any questions concerning 

Michael R. Johnsou Acting Chief 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Docket No.: 50-244 

Enclosure: Inspection Report 50-244/96-201 

74 
REPORT NO.: 

ORGANIZATION: 

ORGANIZATIONAL 

CONTACT : 

INSPECTION DATES: 

LEAD INSPECTOR: 

OTHER INSPECTORS: 

REVIEWED BY: 

APPROVED BY: 

U. S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF TECHNICAL SUPPORT 

50-244/96-201 

Rochester Gas & Electric Corporation (RG&E) 

89 East Avenue 

Rochester, NY 14649 

Dr. Robert C. Mecredy 

Vice President - Nuclear Operations 

January 22 through 26, 1996 

qz+b 

Ani1 S. Gautam Date 

Vendor Inspection Section 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation (NRR) 

Joseph 3. Petrosino, NRR 

AT 

Greqor U Cwal ina, Section Chief 

VendorY Ildspect i on Section 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nucl ear Reactor Regul at i on 

Date 

h,A 

Michael R. Johnson/’\pcting Chief 

Yanch - 

Speci a1 Inspecti on 

Division of Inspect on and Support Programs 

Office of Nucl ear Reactor Regul ati on 

Enclosure 

75 
1 

1.0 INTRODUCTION 

A vendor oversight program is a licensee's process for monitoring the vendors' 

quality control, consistent with the importance, complexity, and quantity of 

the purchased products or services. Evaluation of the process falls, in part, 

under 10 CFR Part 50, Appendix B, Criterion VII, "Control of Purchased 

Material , Equipment, and Services," which requires 1 icensees to establish 

specific measures to assure that purchased material equipment and services 

conform to the procurement documents; and under Regulatory Guide 1.144, 

"Auditing of Quality Assurance Programs for Nuclear Power Plants," which sets 

forth NRC requirements for auditing quality assurance programs. 

A special inspection to assess the adequacy and effectiveness of RG&E's vendor 

oversight program for monitoring the vendors' quality control for safetyrelated equipment was conducted by staff from the Special Inspection Branch of 

the Office of Nuclear Reactor Regulation, from January 22 through 26, 1996. 

Details of the licensee's vendor oversight program are provided in Section 

2.0. The personnel attending the entrance and exit meetings are listed in 

Appendix A. 

2.0 MONITORING OF VENDORS 

In general, RG&E had implemented an effective vendor oversight program to 

monitor vendor's quality control for safety-related products. The inspection 

focused on assessing RG&E's monitoring of nine vendors supplying safetyrelated products to Ginna. 

I 

i 

Vendor oversight program attributes examined included QA organization for 

monitoring vendors; qualification, audits and surveillance (hereafter referred 

to as audits) of Appendix B vendors; evaluation of audit findings and feedback 

of corrective actions to appropriate staff; maintenance of a qualified 

suppliers 1 ist; and qualification and training of appropriate staff. QA 

documents and procedures were reviewed in the following areas: control of 

purchased material, equipment and services; quality review of procurement 

documents; evaluation and control of vendor performance; reporting and 

followup of audits results; tracking of abnormal conditions during 

procurement; and qualification and certification of personnel performing 

audits. 

2.1 Vendor Oversight Program 

In general, an effective vendor oversight program was implemented. The 

inspectors assessed the attributes and implementation of the vendor oversight 

program run by a QA manager and 10 procurement QA representatives. The 

manager reported to the department manager of Nuclear Assessment, who reported 

to the Vice President of Nuclear Operations. The QA manager had the authority 

to conduct appropriate vendor audits. 

QA evaluated vendors in accordance with RG&E procedures, industry informati on 

and audits pertinent to the vendor. Vendors' use of outside organizations was 

typically identified in the vendor's proposal or bid and assessed by RG&E as 

1 
part of the vendors' control of procurement activities. RG&E's quality 

control (QC) procedures provided guidance on detecting fraudulent material 

during receipt inspection and testing. 

RUE Nuclear Assurance issued purchase orders (POs) to vendors from the Ginna 

site. The Quality Procurement Coordinator performed the quality function for 

safety re1 ated POs. Procurement documents included, as applicable, the scope 

of work or material description, technical and quality requirements, QA 

program requirements, right of access to vendor and subtier vendor facilities, 

and requirements for documenting, reporting, and dispositioning nonconforming 

conditions and 10 CFR Part 21 deviations. The licensee stated that QA 

engineers typically did not review POs but received and reviewed copies of 

those POs which imposed source audits on the vendor (e.g., if a restriction 

was placed on a vendor). 

QA staff ensured that qualification and annual, triennial , and other interim 

audits, were performed by RG&E or outside organizations to monitor vendors 

supplying Appendix B products and services. QA audits were performed in 

accordance with written procedures and checklists. Audits included 

monitoring, witnessing and observil,; xtivities, such as inspections, 

examinations, and performance tests. 

The licensee stated that procedure A-1404, Section 3.5.1 addressed 

distribution of industry events reports on an information-only basis by the 

Operating Experience department (Nuclear Operations) to affected departments. 

NUREG-0040, "Licensee Contractor and Vendor Inspection Status Report, I' was 

routed through the procurement QA group. The inspectors questioned why NRC 

Information Notice 88-81, "Failure of Amp Window Indent Kynar Splices and 

Thomas and Betts Nylon Wire Caps during Environmental Qualification Testing," 

was not reviewed by QA procurement engineers when qualifying Amp (Appendix B 

vendor). The licensee stated that the notice was reviewed by its electrical 

engineering staff in 1988 but not forwarded to QA because no changes to the 

design basis for 10 CFR 50.49 environmentally qualified electrical equipment 

were applicable. 

bulletins pertinent to QA evaluations of vendors were currently received by 

procurement QA engineers and agreed to review this area to determine if 

enhancements were needed. 

The 1 icensee determined that NRC information notices and 

QA tracked defects in materials identified at the vendor facility through the 

vendor's corrective action process. 

deficiencies, restrictions were imposed on the vendor or stop-work orders 

issued. Restrictions on a vendor were noted by QA on the qualified suppliers 

list and updated every month or as changes occurred. Nonconformances for 

which the vendor had recommended a disposition of "use-as-is" or "repair" at 

the vendor's facility were normally reviewed and dispositioned by the RG&E 

procuring organization. 

For significant programmatic 

The inspectors had no concerns in this area. 

2.2 Evaluation of Appendix B Vendors 

The inspectors assessed RG&Ey s qual i f i cation and moni tori ng of ni ne vendors 

supplying safety-re1 ated products to Ginna. The vendors included Modumend 

(refurbished power supplies) , Amp Products Inc. (terminals and spl ices) , 

2 
Fisher Service Company (valve parts) Amerace Electronic Components (Agastat 

re1 ays) , Re1 i ance (electric motors), Rosemount Nuclear Instruments (pressure 

and temperature devices), Babcock & Wil cox Canada (steam generators), 

Automatic Switch Company (solenoid operated valves) and Limitorque 

Corporation (valve operators). 

significance of their products and the history of product deficiencies or 

failures. Histories were obtained from electrical and instrumentation 

corrective maintenance work orders, reports of nuclear assessment activities 

regarding component failures, vendor restrictions noted in the QSL, licensee 

action reports of abnormal conditions and events, nuclear plant reliability 

data system equipment failure reports, 10 CFR Part 21 reports, licensee event 

reports, and NRC inspection reports and information notices. 

The vendors were selected based on the safety 

The NRC inspectors reviewed documentation on the qualification and evaluation 

of the nine sel ected vendors, i ncl udi ng RG&E’ s performance-based annual and 

triennial quality audits of vendors; RG&E’s followup of audit findings and 

corrective actions; and its feedback of appropriate findings and corrective 

actions to management and procurernen; ;taff. Documents indicated that 

qualification, annual audits, triennial audits were performed on time; audit 

findings were evaluated; and feedback prc-:ided to appropriate staff, including 

Nuclear Assurance (site), QC, purchasing, procurement, and engineering 

document control. RG&E’s qualification and performance evaluation of Appendix 

B vendors considered the vendor’s design and manufacturing capabilities; the 

QA program and its capability to provide the required items or services; 

audits of the vendor’s QA program status and performance history; and 

maintenance of vendor data on the qualified suppliers list (QSL). 

Initial qualification and annual and triennial audits for qualifying and 

monitoring the nine selected Appendix B vendors were primarily conducted by 

outside organizations, including the Nuclear Procurement Issues Committee 

(NUPIC) and utilities. The NUPIC sent questionnaires to pertinent licensees 

to determine areas the licensees preferred inspected, and communicated its 

audit results to the 1 icensees. The annual vendor evaluation required vendors 

to identify any changes to products, services, facilities, personnel or 

quality involvement which could impact qualifications. 

to identify and disposition all nonconforming conditions in accordance with 

the vendor’s quality requirements. Nonconformances to the technical and 

quality requirements of the POs dispositioned as “use as is” or “repair” were 

required to be submitted to the licensee. 

problems; however, audit evaluation checklists did not require procurement QA 

engineers to actually question a vendor whether any deficiency or 

nonconformance had been identified pertinent to products in process at vendor 

facilities. 

enhancements were needed, 

POs required vendors 

QA expected vendors to identify any 

The licensee agreed to review this area to determine if 

Use of outside organizations for auditing vendors was based on 10 CFR Part 50, 

Appendix B, and Regulatory Guide 1.144, which allow use of outside 

organizations to reduce the number of external audits as an alternative method 

for qualifying and monitoring vendors as long as all pertinent information is 

adequately evaluated. For example, the initial and ongoing qualification and 

performance eval uation of Amp Products, Inc. (Amp), were completely based on 

audits performed by NUPIC and other organizations, and evaluated by RG&E’s QA. 

3 

78 
Amp supplied electrical terminals and splices for safety-re1 ated appl ications 

at the Ginna station and was qualified by QA in 1985 to 10 CFR 50 Appendix B 

requirements. The last QA annual review was performed on August 10, 1993, and 

the last performance audit was performed by NUPIC on June 9, 1995. 

annual review had noted that Amp had made a revision to its QA manual but not 

provided it to RG&E. 

review. Pertinent audit results were provided to appropriate staff. The 

licensee provided evidence that its electrical engineering staff had evaluated 

Amp splice failures and that no restrict’ons were placed on the vendor because 

RG&E decided to install Raychem splices on the Amp nylon splices installed in 

harsh environments (RG&E EEQ-1 Form Package 39). All purchase orders to Amp 

were completed and closed. 

QA’s 

QA requested the vendor to provide the manual for 

The basis for qualification of the selected vendors, including those performed 

by outside organizations, appeared to be adequate and appropriately 

documented. However, one concern was identified regarding RG&E’s assessment 

of the vendor Modumend. Modumend was audited on November 2, 1993, by RG&E 

procurement QA to determine if the vendor could be placed on the RG&E QSL. 

The procurement QA team leader identified one finding, that the vendor had not 

conducted internal audits (required by Appendix B Criterion XVIII). During 

this time (1993-1994), the Quality Performance Department Manager (currently 

reassigned) and QA Manager (retired) encouraged their staff to generate 

revenue for RG&E. 

approached Modumend and suggested RG&E be hired as an independent contractor 

to perform the internal audit that RG&E had earlier found not to have been 

done. The vendor agreed but RG&E later decided not to perform the audit. 

However, with his immediate manager’s approval, in June 1994, the QA team 

leader did the internal audit for the vendor for a fee, independently of RG&E. 

The QA team leader (with the Modumend vice president as a team member) closed 

the RG&E’s earlier finding and identified no new findings. As a result, 

RG&E’s QA team leader performed Modumend’s quality function, instead of 

assessing Modumend’s performance of the function. 

With the QA manager’s approval, the QA team leader 

The inspectors noted that RG&E’s pol icy statement, “Principles of Business 

Conduct,” dated September 1992, issued by the Chairman, President and Chief 

Executive Officer of RG&E, stated that without the specific written approval 

of the Chairman of the Board, no employee should serve as a director, officer, 

consultant or employee of any business organization which was a competitor or 

supplier of RG&E where there was the risk or appearance that preferential 

treatment may be given or received. 

revenues from and serve as a consultant of Modumend appears to be in conflict 

with this policy. 

Encouraging QA engineers to raise 

The licensee stated that current management would avoid any perception of a 

conflict of interest with a vendor and would not suggest to their engineers to 

perform such an audit. During the NRC inspection, the licensee instructed QA 

engineers not to perform work for a company that QA was responsible to assess, 

either as independent consultants or as RG&E employees. The inspectors had no 

further concerns in this area. 

4 

79 
2.3 Qualified Suppliers List 

QA maintained the QSL on an electronic database and was responsible for making 

changes to existing vendor information. 

information, details of vendor capability, dates of vendor qualification, and 

any vendor restrictions (e.g., RG&E quality checks to be performed during 

manufacture and/or prior to shipment). 

pertinent information for the nine selected vendors was maintained on the QSL. 

The QSL included vendor contact 

The inspectors determined that 

QA used a password to restrict access to the QSL database and protect it from 

inappropriate changes or deletions (e.g., removal of vendor restrictions). 

The accounts payable staff could only access the data base to enter 

information about new vendors. 

The inspectors had no concerns in this area. 

Other RG&E groups had "read only" access. 

2.4 Training and Self-Assessment 

Qualification records indicated that individuals assigned to perform 

procurement activities were properly trained and qualified. Qualification 

records documented the knowledge and training of procurement QA engineers, 

including their performance of quality activities (e.g., assisting in audits) 

before their qualification as auditors or lead auditors. 

satisfactory completion of an oral examination on the QA program and 

procedures, NRC regulations, regulatory guides, NUPIC and Electric Power 

Research Institute (EPRI) guidelines, and industry standards; attendance of 

appropriate training classes; and completion of required reading on qualifying 

and auditing vendors. The procurement QA engineers were certified as audit 

team leaders to American National Standard Institute (ANSI) N45.2.23, and as 

quality system Lead Assessors to the American Registration Assessment Board of 

the International Organization for Standardization (ISO) 9000. The inspectors 

had no concerns in this area. 

Records indicated 

The last independent assessment of RG&E's QA program, including its vendor 

oversight program, was performed on December 22, 1994. An October 16, 1995, 

internal assessment was performed by a technical specialist from EPRI and 

other RG&E team members. The internal assessment team concluded that RG&E's 

overall procurement engineering program was technically sound and 

significantly better than at most nuclear facilities reviewed. However, the 

assessment noted road blocks in two-way communication between procurement 

engineering and QA procurement. The assessment recommended refresher training 

for staff and managers. 

sel f-assessment of the procurement process were planned with a completion date 

of July 1996, and that QA would assess if any remedial training was necessary 

to correct the lack of communication identified by the independent audit. The 

inspectors had no concerns in this area. 

The licensee stated that actions in response to a 

3.0 EXIT INTERVIEW 

.Idle inspectors conducted an exit meeting on January 26, 1996 at the RG&E 

corporate office in Rochester, New York, to discuss the major areas reviewed 

during the inspection. 

identified with an asterisk in Appendix B of this report. 

not identify any documents or processes as proprietary. 

Licensee representatives who attended this meeting are 

The licensee did 

5 

80 
APPENDIX A 

Persons Contacted at Entrance and Exit Meetitigs 

* 

* 

* 

* 

* 

* 

* 

* 

* 

* 

* 

* 

* 

* 

Rochester Gas & Electric 

Richard J. Watts 

Michael P. Lilley 

Mark Shaw 

George Wrobel 

Raymond M. Bozarth 

Paul Hutner 

Gregory R. Amsden 

A1 Pitts 

Jon Bergstrom 

Michelle L. Preik 

Terry Ki rkpatri ck 

Manager Nuclear Assessment 

Manager Quality Assurance 

Manager Material s and Procurement 

Manager NS&L 

SA Team Leader 

QA Engineer 

QA Engineer 

QA Engineer 

QA Engineer 

QA Engineer 

Quality Procurement Coordinator 

Nucl ear Regul atorv Commi ss i oil 

Ani1 S. Gautam Team Leader, NRR 

Joseph J. Petrosino Reactor Inspector, NRR 

Peter D. Drysdale Senior Resident Inspector, Ginna 

Denotes those attending the exit interview on January 26, 1996, at the 

conclusion of the inspection 
Ident i f i er 

Selected Generic Correspondence on the Adequacy of 

Vendor Audits and the Quality of Vendor Products 

Information Notice 96-02 

Information Notice 96-06 

Informat ion Notice 96-07 

Title 

Inoperabi 1 i ty of Power Operated Re1 i ef Val ves 

Masked by Downstream Indication During Testing 

Design and Testing Deficiencies of Tornado 

Damper at Nuclear Power P1 ants 

Slow Five Percent Scram Insertion Times Caused 

by Viton Diaphragms in Scram Solenoid Pilot 

Valves 
~~ 

JRC FORM 335 2-891 

JRCM 1102, 

,201, 3202 

U.S. NUCLEAR REGULATORY COMMISSION 

BIBLIOGRAPHIC DATA SHEET 

(See instructions on the reverse] 

!.TITLE AND SUBTITLE 

Licensee Contractor and Vendor Inspection 

Status Report 

Quarterly Report 

January-March 1996 

I. AUTHOR(S) 

I. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Officeor Region, U.S. Nuclear Regulatory Commission, and mai1ingaddress;if conrracror, provide 

name and mailing a9dmss.l 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

U.S. Nuclear Regulatory Commission 

Washington, D. C. 20555-0001 

I. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, type "Saneasabove'>if contractor,provide NRC Division, Office or Region, U.S. Nuclear Regulafory Commission, 

and mailing addresrl 

Same as above. 

0. SUPPLEMENTARY NOTES 

, REPORT NUMBER 

(Auloned by NRC. Add Vol.. Supp., Rev., and Addendum Numbers, if any.) 

NU REG-0040 

Vol. 20, No. 1 

. DATE REPORT PUBLISHED 

MONTH YEAR 

Ma v 1996 . FIN OR GRANT NUMBER 

b. TYPE OF REPORT 

Quarterly 

. PERIOD COVERED (InclusiveDaresl 

January - March 1996 

1. ABSTRACT (200 words or less/ 

This periodical covers the results of inspections performed by the NRC's Special 

Inspection Branch, Vendor Inspection Section, that have been distributed to the 

inspected organizations during the period from January through March 1996. 

2. KEY WORDS/DESCRIPTORS (List words orphrases rhat willassist researchers in locating the report.) 13. AVAILABILITY STATEMENT 

Vendor Inspection 

1 Unlimited 

14. SECURITY CLASS1 FlCATlON 

I (This Pagel 

Uncl assi f i ed 

(This Report) I 

I 16. PRICE 

IRC FORM 335 (2-89)