NUREG-0040 

Vol. 19, No. 2 

Licensee Contractor 

and Vendor Inspection 

Status Report 

Quarterly Report 

April - June 1995 

U.S. Nuclear Regulatory Commission 

Office of Nuclear Reactor Regulation 

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document. 
NUREG-0040 

Vol. 19, No. 2 

Licensee Contractor 

and Vendor Inspection 

Status Report 

Quarterly Report 

April - June 1995 

Manuscript Completed: August 1995 

Date Published: August 1995 

Division of Inspection and Support Programs 

Ofice of Nuclear Reactor Regulation 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

ABSTRACT 

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

Special Inspection Branch, Vendor Inspection Section, that have been 

distributed to the inspected organizations during the period from April 1995 

through June 1995. 

DISCLAIMER 

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

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

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

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

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

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

United States Government or any agency thereof. 

iii 

CONTENTS 

PAGE 

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

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

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

Automatic Switch Company (99900369/95-01) ............... 2 

Florham Park, New Jersey 

ABB Combustion Engineering 

Nucl ear Operations 

Windsor, Connecticut 

Dow Corning U.S.A. 

Midland, Michigan 

Dresser Industries, Inc. 

Alexandria, Louisiana 

Energy & Process Corporation 

Tucker, Georgia 

Rosemount Nuclear Instruments 

Eden Prarie, Minnesota 

Uni strut Corporation 

Wayne, Michigan 

Whittaker Electronic Systems 

Simi Valley, California 

(99900538/95-01) ............... 15 

(99901284/95-01) ............... 31 

(99900054/95-01) ............... 41 

(99900866/95-01) ............... 49 

(99900271/95-01) ............... 69 

(99900362/90-01) ............... 81 

(99901164/95-01) ............... 115 

Select Generic Correspondence on the Adequacy of Vendor .................. 125 

Audits and the Quality of Vendor Products 

Correspondence Re1 ated To Vendor Issues .................................. 126 

V 

INTRODUCTION 

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

licensing and inspection program is that licensees are responsible for the 

proper construction and safe and efficient operation of their nuclear power 

plants. 

for the inspection of commercial nuclear facilities to provide for multiple 

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

part i ci pates i n a qual i ty veri f i cat i on process in compl i ance with requi rements 

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 establ i shed 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 

i s responsi bl e for ensuring that suppl i ers , contractors and vendors a1 so have 

sui table and appropriate qual i ty programs that meet NRC requirements, guides, 

codes, and standards. 

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

and i nspects nucl ear steam system suppl i ers (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 i censes in vendor-re1 ated areas. 

inspections are done to ensure that the root causes of reported vendor-re1 ated 

problems are determined and appropriate corrective actions are developed. The 

inspections also review vendors to verify conformance with applicable NRC and 

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

coordination between licensees and vendors. 

These 

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

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

and review of equipment problems found during operation and their corrective 

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

the inspected organization is required to take appropriate corrective action 

and to institute preventive measures to precl ude recurrence. 

implications are found, NRC ensures that affected licensees are informed 

through vendor reporting or by NRC generic correspondence such as information 

notices 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 1 ists the nuclear facilities inspected. 

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 

This information will also 

viii 
INSPECTION REPORTS 

1 
***** 

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-0001 

May 15, 1995 

Mr. Randy P. Smith, President 

Automatic Switch Company 

50-56 Hanover Road 

Florham Park, NJ 07932 

SUBJECT: NRC INSPECTION NO. 99900369/95-01 

Dear Mr. Smith: 

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

(NRC) inspection of Automatic Switch Company (ASCO), Florham Park, New Jersey, 

conducted on March 13 and 14, 1995. The NRC inspection team, led by Joseph 

Petrosino and comprised of other inspectors named in the report, conducted a 

1 imited inspection to evaluate the establishment and implementation of 

selected portions of your quality assurance program, including the period when 

ASCO was actively manufacturing NP8323 solenoid-operated valves (SOVs) that 

were supplied for use in nuclear power plant reactor main steam isolation 

valves (MSIVs) for boiling-water reactors. This letter also discusses 

activities that were performed after the completion of this inspection and 

that were discussed among NRC, ASCO and NRC licensees. 

developed after this inspection regarding the conclusions and statements made 

in the enclosed report may be submitted to the NRC. 

conducted to provide a basis for NRC staff confidence that ASCO's NP8323 SOVs 

supplied to the nuclear industry would perform their safety-related function. 

Any new information 

The inspection was 

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

representative records, (b) held discussions, (c) 1 istened to presentations 

and (d) observed an ASCO technician assemble one NP8323 SOV w thout using any 

lubricant and witnessed its related testing. On the basis of this inspection, 

the NRC team identified one nonconformance relative to Append x B to Part 50 

of Title 10 of the Code of Federal Requlations (Appendix B of 10 CFR Part 50). 

The team identified that ASCO failed to ensure that its September 29, 1989, 

engineering department decision to cease using si1 icone based fluids and 

greases in its NP8323 SOVs was translated into an ASCO manufacturing 

department procedure or instruction. As a result, ASCO continued to fabricate 

NP8323 SOVs using silicone fluids and greases. The enclosed inspection report 

contains a detailed discussion of the areas examined and the NRC team 

observations. 

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

statement in accordance with the instructions specified in the enclosed Notice 

of Nonconformance. 

show good cause for us to do so. 

We will consider extending the response time if you can 

2 
R. Smith -2- 

The response requested by this letter is not subject to the clearance 

procedures of the Office of Management and Budget as required by the Paper 

Reduction Act of 1980, Pubic Law No. 96-511. In accordance with 10 CFR 2.790 

of the Commission’s regulations, a copy of this letter and the enclosed 

inspection report will be placed in the NRC’s public document room. 

have any questions regarding this matter, please contact Mr. Joseph Petrosino 

at (301) 415-2979. 

If you 

Sincerely, 

Robert M. Gallo, Chief 

Special Inspection Branch 

Division of Technical Support 

Office of Nucl ear Reactor Regul at i on 

Docket No. 99900369 

Enclosures: 1. Notice of Nonconformance 

2. Inspection Report 99900369/95-01 

cc w/encls.: Mr. F. R. B. Matthews, Vice President 

Product Development and Technology 

Valve Department 

Automatic Switch Company 

50-56 Hanover Road 

Florham Park, NJ 07932 

Mr. David B. Tompsen, Quality Assurance Director 

Automatic Switch Company 

50-56 Hanover Road 

Florham Park, NJ 07932 

Mr. William C. Kirchhoff, NED Superintendent 

LaSall e Site Support Engineering 

LaSal 1 e County Nucl ear Stat i on 

RR#1 Post Office Box 220 

2601 North 21st Road 

Marseilles, IL 61341 

3 
NOTICE OF NONCONFORMANCE 

Automatic Switch Company 

Florham Park, New Jersey 

Docket No.: 99900369 

Report No.: 95-01 

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

conducted at the Automatic Switch Company (ASCO) facility on March 13 and 14, 

1995, it appeared that certain of your activities were not conducted in 

accordance with NRC requirements. 

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

requires, in part, that measures shall be established to assure that the 

design basis for those components to which this appendix applies are 

correctly translated into specifications, drawings, procedures, and 

instructions. 

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

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

prescribed by appropriate instructions, procedures, or drawings and be 

accomplished according to those instructions, procedures, or drawings. 

Contrary to these requirements, as of March 14, 1995, ASCO failed to 

ensure that its September 29, 1989, engineering department decision to 

cease using silicone-based fluids and greases in its NP8323 solenoidoperated valves was translated into a procedure or instruction for 

ASCO's manufacturing department personnel implementation. Subsequent to 

that decision, solenoid-operated valves supplied for nuclear use were 

manufactured using silicone-based greases. (95-01-01) 

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

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

0001, with a copy to the Chief, Special Inspection Branch, Division of 

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

date of the letter transmitting this Notice of Nonconformance. 

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

include (1) a description of steps that have been or will be taken to correct 

this item, (2) a description of steps that have been or will be taken to 

prevent recurrence, and (3) the dates your corrective actions and preventative 

measures were or will be completed. 

This reply 

Dated at Rockvi 11 e, Mary1 and 

this 15th day of May 1995. Enclosure 1 

4 
REPORT NO.: 

ORGANIZATION: 

ORGAN I ZAT I ONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY: 

INSPECTION DATES: 

LEAD INSPECTOR: 

OTHER INSPECTORS: 

REVIEWED BY: 

APPROVED BY: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF TECHNICAL SUPPORT 

99900369/95-01 

Automatic Switch Company 

50-56 Hanover Road 

Florham Park, New Jersey 07932 

Hr. David B. Tompsen, QA Director 

(201) 966-2731 

Manufacturer of solenoid-operated valves 

and pressure and temperature switches. 

March 13 and 14, 1995 

,;I ./ ./ 

Joaeph J. Petrosino, Team Leader 

Vendor Inspection Section (VIS) 

Special Inspection Branch (TSIB) 

Division of Technical Support (DOTS) 

Kamalakar R. Naidu, V1S:TSIB 

Edward R. Schweibinz, Region 111 

Robert M. Gallo, Chief, TS1B:DOTS 

Date 

Enclosure 2 

5 
1 SUHHARY OF FINDINGS 

During this inspection, the NRC inspection team evaluated the Automatic Switch 

Company (ASCO) quality programs and their implementation for the fabrication 

of solenoid-operated valves (SOVs). The inspection was conducted to provide a 

basis for confidence that NP8323 SOVs supplied to the U.S. nuclear industry 

would perform their safety function. 

The inspection basis consisted of the following: 

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

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

Requlations (Appendix B of 10 CFR 50) 

a Part 21, "Reporting of Defects and Noncompliance," o 

One nonconformance was identified and is discussed in Sect 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

During this inspection effort, the NRC inspection team did 

findings identified during previous NRC staff inspections. 

10 CFR 

on 3.7 herein. 

not review any 

3 INSPECTION FINDINGS AND OTHER COMMENTS 

3.1 Bac kqround 

The Ralph A. Hiller Company (RAH), Pittsburgh, Pennsylvania, designed and 

manufactured speciality valve actuators for various nuclear industry 

applications. For nuclear power plant applications, RAH's fluid power 

actuators were generally categorized as being either pneumatic or hydraulic, 

with a number of design variations. MSIV applications typically use a 

pneumatic-hydraul ic combination actuator. RAH pneumatic-hydraul ic combination 

actuators are used to operate MSIVs at Commonwealth Edison Company's (ComEd's) 

LaSalle County Station (LaSalle). A typical RAH actuator consists of two 

subassemblies. 

hydraulic assembly (PHA). 

NP8323 SOVs for the pilot valve to control the opening and closing of the 

MSIVs. The ASCO NP8323 SOV is a dual coil, three-way solenoid valve used by 

many licensees in their main steam isolation valve (MSIV) applications. 

1989, ASCO issued an ASCO field notification to all applicable customers 

advising that its NP8323 SOV series valves would be discontinued and suggested 

that all NP8323 SOVs be removed from MSIV piloting applications. 

recommended to licensees that they use two ASCO NP8320 SOVs instead of one 

NP8323 SOV in its MSIV applications. 

The pneumatic control assembly (PCA) and the pneumatic 

The PCA subassembly was designed to use ASCO Model 

In 

ASCO also 

On February 18, 1995, while LaSalle, Unit 2, was entering into a refueling 

outage, an operator in the control room was unable to close an MSIV when he 

turned the selector switch to the "close" position. 

selector switch to the "test" position, the MSIV slow-closed. 

the switch to the "close" position, the MSIV reopened. 

a second outboard MSIV in Unit 2 would not close when the operator in the 

control room moved the control switch to the "close" position. This MSIV 

When he moved the 

When he moved 

On February 19, 1995, 

2 
would also not close when the operator moved the switch to the "test" 

position. 

which exhibited the problems. 

ComEd stated that ASCO NP8323 SOVs were installed on the MSIVs 

3.2 Entrance and Exit Meetinqs 

During the entrance meeting in Florham Park, New Jersey, on March 13, 1995, 

the NRC team met with members of ASCO management and staff, discussed the 

scope of the inspection, and established contact persons for the NRC team 

within ASCO management and staff. During the exit meeting on March 14, 1995, 

the NRC team summarized the inspection findings for ASCO management and staff. 

Inspection participants and personnel contacted during the performance of the 

inspection-related activities are listed in Section 4 of this report. 

3.3 PrinciDle of ODeration of the MSIV Actuator 

The RAH actuators that operated the MSIVs at LaSalle are used at other nuclear 

power plants. Typically, the 1 icensee staff operators use pushbutton stations 

on the main control panel to operate the actuators to fast-close, slow-close, 

or open the MSIVs from the control room. The reactor protection system logic 

can automatically initiate a signal to the actuator to close the MSIV when the 

1 ogic senses adverse operating conditions. The RAH actuators typically are 

energized with a 120 volt alternating current (ac) electrical power source and 

90 psig air pressure. ASCO's NP8323 series, dual coil, three-way SOVs are 

installed in the PCA and function as pilot valves to direct air in the 

appropriate direction to the PHA to either close or open the MSIV. When 

energized, which is the normal operating state, the SOV directs air into the 

PHA, creating a force against the springs to keep the MSIV open; when the SOV 

is de-energized, it directs the air away from the PHA, which allows the spring 

pressure to close the MSIV. Sticking internals inside the SOV and foreign 

particles lodged in the SOV can prevent change of state, so that the direction 

of the air to the PHA is not reversed and the MSIV remains open. 

Over the last several years, the NRC staff has issued numerous generic 

communications discussing several solenoid valve concerns, such as foreign 

material found lodged inside SOVs and degraded operation of various SOV 

internals that experienced degraded operation due to a variety of problems, 

including sticky residues on certain internals. Examples of generic 

communications include: 

NUREG-1275, Vol ume 6, "Operat i ng Experi ence Feedback Report-Sol enoi d 

Operated Val ve Problems, I' February 1991. 

Generic Letter 91-15, "Operating Experience Feedback Report, SolenoidOperated Val ve Problems at U. s. Reactors, '' September 1991. 

IN 89-66, "Qual i f i cat i on Li fe of Sol enoi d Val ves, I' September 1989. 

IN 87-28, "Air System Problems at U.S. Light Water Reactors," June 1987. 

Supplement 1 to IN 87-28, December 1987; 

IN 86-57, "Operating Probl ems With Sol enoid Operated Val ves at Nuclear 

Power P1 ants, I' July 1986; 

3 

7 
e IN 85-17, "Possible Sticking of ASCO Solenoid Valves," March 1985 

Supplement 1 to IN 85-17, October 1985. 

3.4 Actions and Conclusions Followinq the MSIV SOV Failures 

3.4.1 Commonwealth Edison 

Following the failures of the MSIVs to close on February 18, 1995', as 

discussed in Section 3.1 above, LaSalle personnel continuously energized 

solenoids A and B of the ASCO NP8323 dual solenoid SOV for 1 hour and measured 

the current in both the coils and also monitored the pressure of the air 

supply. 

they observed that the SOV did not change its internal valve air flow 

position, indicating that the SOV had failed. Consequently, the LaSalle 

personnel removed the following SOVs for analysis: the two ASCO NP8323 SOVs 

which had failed, the remaining two outboard SOVs which did not fail, one 

inboard SOV, and one ASCO single-action test solenoid which was mounted 

outboard. The SOVs were disassembled utilizing ASCO's Installation and 

Maintenance Instruction (I&M) sheets. Upon disassembling the SOVs, the 

LaSalle personnel did not find any anomalies on the side of the SOV where 

solenoid A was located. However, on the solenoid B side of several of the 

SOVs, LaSalle personnel found some residue on the disc holder (milk stool), 

backup washer, and stem. LaSalle personnel also observed a brownish 

gelatinous film around the mating surfaces of the core and plug nut of one of 

the failed SOVs, which appeared to be holding the two parts together. 

When the operators de-energized solenoids A and B simultaneously, 

After the LaSalle personnel's initial analysis of the SOVs, LaSalle sent the 

SOV components to ComEd's System Materials Analysis Department (SMAD) for 

further tests. All of SMAD's observations and conclusions were documented in 

SMAD Report No. M-00954-95, "MSIV Solenoid Failures, LaSalle Station, 

(A95-140)," issued March 1, 1995. As the report indicated, the SOV components 

came to SMAD disassembled, and SMAD was informed by the LaSalle staff that two 

of the SOVs failed to change position during operation and that the other four 

solenoids were removed from service and disassembled for comparison. The two 

failed SOVs were identified as being on the Unit 2 outboard MSIVs, 2B21-FO28A 

and 2B21-FO28D. The LaSalle personnel also submitted reference samples to 

SMAD of Dow Corning's 550 silicone fluid and a sample of Nyogel 775A silicone 

grease. 

SMAD photographed all of the ASCO NP8323 SOV core and plug nut sets prior to 

starting any analyses of the components. 

sets showed residue on both core and plug nut sets, another outboard set 

showed a residue, and the fourth outboard set and an inboard set showed no 

residue. However, the inboard and outboard sets which had no residue showed 

evidence of core face wear. The report stated that portions of the residues 

taken from the core and plug nut sets on the failed SOVs were examined by 

microscopic infrared spectroscopy and found to contain methyl phenyl 

polysiloxane and that the organic portion of these residue deposits was 

removed and the residue examined by a scanning electron microscope equipped 

with an X-ray analyzer. 

The report indicated that the failed 

This analysis with the X-ray analyzer showed that the 

As discussed in Section 5.2.4.4 of NUREG-1275, Vol. 6, a similar core 1 

and plug nut adhesion problem occurred at LaSalle in 1987. 

4 

8 
residue contained major amounts of iron; minor amounts of silicon, chromium 

and copper; and a trace of zinc. SMAD also analyzed the NP8323 cores and 

plugs with an alloy analyzer and found that the NP8323 components were 80.1% 

iron, 16.9% chromium, 1% manganese, and 0.3% molybdenum. 

The SMAD report also speculated that due to vibrations in the plant, the 

mating surfaces of the core and plug nut rubbed together (fretted) and created 

microscopic wear part i cl es. 

could then accumulate in a binding material such as Nyogel 775A silicone 

grease. SMAD's report stated that the elemental analysis of the core and plug 

nut deposits indicated a large amount of wear particles and that the two 

failed core and plug nut sets had a total depositing of approximately 

1 milligram. SMAD also speculated that as more and more particles accumulated 

in the grease, the viscosity of.the grease increased making it tacky and the 

formation of this tacky substance between the core and plug nut may have 

caused them to stick when the solenoid was de-energized. 

The report indicated that the wear part i cl es 

The SMAD report stated that the reference lubricants were analyzed by infrared 

spectroscopy and that both reference samples contained methyl phenyl 

polysiloxanes. The report stated that the two spectra were very similar but 

with subtle differences, and that peak-by-peak comparison sheets were 

attached. 

polysiloxane and similar to Nyogel 775A except for the core and plug nut set 

from an outboard NP8323 SOV, which more closely resembled 

poly(dimethylsi1oxane). 

contain a combination of methyl and phenyl functionality , whereas the 

poly(dimethy1 siloxane) contained only the methyl functional ity. The SMAD 

report indicated that its results indicated that another si1 icone-based 

lubricant may have been present. 

The report stated that most of the deposits were methyl phenyl 

SMAD's report stated that Nyogej 775A and Dow 550 

3.4.2 Automatic Switch Company 

ASCO did not receive either of the failed NP8323 SOVs from the LaSalle 

station, nor did it receive any of the SOV components that were analyzed. 

However, LaSalle did provide ASCO with a copy of SMAD's report, which 

contained data sheets from its infrared spectra analysis and associated 

photographs, and also provided ASCO with four unused NP8323 SOVs which had 

been stored in the ComEd warehouse since 1989. ASCO disassembled the four 

NP8323 SOVs and observed a small quantity of Nyogel 775A in the vicinity of 

where the ASCO assembler applied a dab of Nyogel to hold the back-up washer 

onto the milk-stool disc on each SOV. ComEd personnel from LaSalle witnessed 

some of ASCO's disassembly and inspection activities. 

The ASCO report indicated that no other traces of Nyogel 775A were identified 

on the valve internals. However, the team noted that the ComEd staff who 

witnessed the disassembly of the SOVs observed a residue on one of the plug 

nuts. The ComEd trip report stated that ''a small amount of white colored 

residue was observed on the surface that interfaces with the core assembly. 

When picked at with a dental pick, the residue smeared." 

copy of the SMAD report, ASCO staff obtained the services of an independent 

When ASCO received a 

'Functionality refers to the atom or group of atoms that define the 

structure of a particular family of organic compounds and determines their 

properties. 

5 

9 
laboratory to review the SMAD report infrared spectra analysis information and 

to make an independent determination of the SMAD test data. 

its review and analysis in ASCO "Engineering Report 309, Investigation of 

Reported Failure to Shift at Commonwealth's LaSalle Unit [2] Nuclear Facility, 

ASCO SOV NP8323A20V,I1 issued March 27, 1995. In the background section of the 

report ASCO stated that the failed outboard MSIV SOVs were operated in an 

ambient temperature environment of approximately 115°F and the ambient 

temperature in this area could increase to as much as 170°F for several hours. 

The SOVs each saw 90-100 psi and each piloted a "Norgren" valve. 

background section of the report indicated that a correlation existed between 

the SOV failures and the different air supply sources that are used for the 

inboard and outboard MSIVs. That is, the outboard MSIVs, two of which failed, 

used station instrument air supply, whereas the inboard MSIVs, which did not 

have any failures, used a nitrogen system. ASCO also stated that ComEd used a 

"graphoil" paste on the SOV/MSIV interface. 

ASCO documented 

The 

The ASCO report discusses some of the ComEd results, takes exception to some 

aspects, and discusses the details of others. For example, ASCO noted that 

SMAD's photographs of the B side core and plug nut mating surfaces showed a 

brownish residue, whereas the A side components were clean. ASCO indicated 

that a brownish residue was also apparent on one leg of a disc holder assembly 

(DHA) and on the back of a DHA back-up washer. 

from the residues that contaminants were present in the SOVs. 

"Independent Lab Analysis of Residues," of ASCO's report stated: 

ASCO states that it is clear 

Section 3.0, 

the brown residues believed to have caused the two SOV shifting 

failures consist of silicone grease or silicone oil Contaminated 

with an ester-alcohol combination.. . Although there is a strong 

indication of the presence of a silicone grease or silicone oil, it 

is questionable whether it is Nyogel 775A. There are several subtle 

but significant differences between the reference Nyogel 775A 

spectra and the unknown residue of interest. These findings are 

consistent with contaminants that have been found to cause slow 

shifting or failure to shift in ASCO scram solenoid pilot valves. 

In those cases it was concluded that the silicone grease or silicone 

oil became contaminated with volatiles from uncured thread sealants 

or with ester oils possibly used as air compressor lubricants. 

Section 6.0, "Conclusions," of ASCO's report generally indicated that the Dow 

Corning 550 fluid or Nyogel 775A grease alone are not causing the sticking. 

ASCO believes that the major source of the contaminants is from the instrument 

air system piping or the Norgren valve backflow air. 

only the B side of the SOVs have been contaminated, it is possible that the 

cy1 inder connection, the Norgren valve at LaSall e, has introduced the 

contamination. When the ASCO SOV is de-energized, there is a backflow of air 

from the Norgren valve through the B side of ASCO's SOV for exhaust. 

states that any contamination that may be in the Norgren valve or connecting 

piping can therefore deposit in the B side of the NP8323 SOW. 

ASCO states that since 

ASCO 

As a preface to the conclusions in the report, the ASCO personnel recalled to 

NRC staff that in December 1987 LaSalle had experienced a similar problem with 

an ASCO NP8323 SOV used on an outboard MSIV. 

and photographs with the recent report and photographs and stated that its 

review of the two event? "show both to be remarkably similar." 

that in the 1987 event, ComEd concluded that Dow Corning 550 fluid caused the 

ASCO compared the 1987 reports 

ASCO stated 

6 

in 
problem, whereas ASCO concluded that the failure was caused by a brownish 

residue found on the B side of the NP8323 SOV and that this residue contained 

contaminants from external sources. 

circumstance was not mentioned in its 1987 report, an independent laboratory 

and a technical expert identified the brownish residue as a silicone grease or 

oil contaminated with an ester-alcohol combination. 

ASCO stated that, although this 

3.4.3 Current & Future Activities by ComED and ASCO 

Following the NRC inspection at ASCO, ComEd and ASCO continued to review the 

circumstances surrounding the MSIV SOV failures at LaSalle and conducted 

discussions regarding similar ASCO SOV component analyses concerning "brown 

residue" that was identified at other licensee facilities. The activities 

that ComEd and ASCO are performing is an attempt to isolate all of the 

probable causes of the degraded operation of ASCO's NP8323 SOV in certain 

conditions. 

actions, for example. 

ComED and ASCO have kept the NRC staff apprised of the ongoing 

On April 7, 1995, LaSalle staff informed NRC that the last three Unit 2 

inboard MSIV SOVs had been examined. Some of The ComEd activities were 

witnessed by Region I11 NRC staff. 

observed a brownish gelatinous residue on the interfacing junction of the core 

and plug nut of one of the three inboard SOVs. This brownish residue appeared 

to be the same as that found previously on the core and plug nut interface on 

the outboard SOVs. All three SOV stems also exhibited a small amount of a 

opaque residue. 

observed on the core and stem of one of the SOVs. 

staff transported the three SOVs to SMAD for analysis. 

ASCO personnel about the latest finding by ComEd staff. 

The licensee staff stated that it had 

An opaque, white and colored gelatinous substance was also 

Subsequently, the LaSalle 

The NRC staff informed 

During the first three weeks of May 1995, discussions were conducted between 

LaSalle, SMAD, ASCO, and ASCO's independent laboratories. The discussions 

involved the individual analysis results of SMAD and ASCO, and the results of 

previous analyses of other ASCO SOV components which exhibited questionable 

performance at other licensee facilities and in other licensee applications. 

The NRC staff perceives that additional information relevant to the brown 

residue issue will continue to be identified by the research activities of the 

ComEd and ASCO personnel. Consequently, additional information may be 

available to the industry after ComEd and ASCO complete their additional 

reviews of this matter. 

3.5 ASCO Installation and Maintenance Instructions 

While discussing the circumstances of the failures and potential root causes, 

the inspection team queried ASCO about the adequacy of a 90-day cycling 

frequency. ASCO indicated that more frequent SOV cycling is better than less 

frequent cycling. When the inspection team asked ASCO to bound their 

statement, ASCO stated that its I&Ms previously recommended that a 30-day 

maximum cycling be performed to insure proper opening and closing. However, 

due to customer pressures that a 30-day cycling was too conservative and set 

too strict a requirement, ASCO changed the recommendation to allow each 

customer to set its own standards for cycling. The team reviewed several of 

ASCO's current and past I&Ms for its NP8320, NP8323 and NP8321 series SOVs, 

and found that prior to 1981, each I&M stated: "while in service, operate 

valve at least once a month to insure proper opening and c7osing." The team 

11 
noted that the current I&Ms for nuclear NP series SOVs stated: "while in 

service, operate valve periodically to insure proper opening and closing." 

Discussions with regional NRC staff indicated that some licensee MSIV 

applications do not have a cycling policy for its MSIV SOVs. 

indicated that an 18-month period without cycling would not be recommended by 

ASCO. 

The ASCO staff 

3.6 Review of ComEd Purchase Order 

After ASCO notified ComEd and its other customers in 1989 that it would 

discontinue the manufacture of NP8323 dual SOVs, several customers, including 

several nuclear plant facilities, ordered large quantities of NP8323 SOVs. 

For example, ComEd decided to procure 50 NP8323 SOVs so that it could continue 

to use the ASCO's NP8323 SOV i n i t s MSIW application for a few more refueling 

cycles until it decided upon its replacement course of action. As a result, 

ComEd issued purchase order (PO) 330255, dated October 6, 1989, to ASCO for 

the supply of 50 NP8323A solenoid-operated valves with Viton seals and discs, 

normally closed, coil insulation class H, in explosion proof enclosure 

suitable for instrument air and environmentally qualified to Revision 1 of 

ASCO Test Report No. AQR-67368. 

assurance program during the manufacture of the SOVs and to provide a 

certificate of conformance for the SOVs being supplied. 

relevant quality requirements including 10 CFR Part 21. 

The PO required ASCO to implement its quality 

The PO specified all 

The SOVs shipped to LaSalle in 1990 by ASCO were stored for use in future 

refueling outages and the two subject failures were the only problems 

identified with this batch. According to a LaSalle engineer, eight NP8323 

SOVs were typically replaced during each refueling outage and were also 

required to be cycled on a 90-day (2 10%) cycle frequency. However, it was 

noted that the two NP8323 SOVs, which had failed, had their cycling extended 

to 118 days from the normal cycle period. 

3.7 Discontinuation of Silicone Fluids and Greases in NP8323 SOVs 

ASCO informed the inspection team that based on previously experienced 

problems with lubricants in NP8323 type SOVs, it had documented in engineering 

report No. 264, "Discontinuance of Fluid Lubricants from NP8323 Valves," dated 

September 29, 1989, which was part of ASCO's qualification basis, that it had 

planned to discontinue the use of silicone-based fluid lubricants in the 

assembly of NP8323 SOVs. 

have contributed to previous SOV failures, including a 1987 NP8323 failure at 

LaSalle. 

implicate lubricants in the failure of the valve to shift to the de-energized 

state, ASCO is removing all silicone-based fluids and greases from within the 

NP8323 valve bodies, and that ASCO had used the fluids solely for lubrication 

of static seals during assembly and for SOWS operating with alternating 

current to reduce objectionable hum during run-in period. ASCO addressed the 

latter by 100 percent functional and noise testing, which assured chatter-free 

operation in the field. 

clean water, instead of the silicone fluids and greases, to the O-rings at the 

interfaces of the metallic components. 

determined that, because the changes did not affect the operation of the SOV, 

the envi ronmental qual i f i cat i on of the Val ve was not compromi sed. 

ASCO suspected that Dow Corning compound 550 may 

The report stated that although ASCO cannot find evidence to 

The report also recommended a light application of 

ASCO's engineering department 

8 

12 
ASCO engineering expected a1 1 NP8323 SOVs manufactured after September 29, 

1989, to be free of silicone-based fluids and greases; however, the inspection 

team determined that ASCO did not ensure that its decisions to discontinue the 

use of silicone products in NP8323 SOVs were translated into ASCO assembly or 

fabrication procedures or instructions to ensure that the NP8323 SOVs were 

assembled in accordance with the engineering department decision. 

of ASCO's failure to establish new procedures or to modify existing 

procedures, ASCO NP8323 sol enoid-operated valves suppl i ed for nucl ear use 

after September 29, 1989, were supplied and manufactured using silicone-based 

greases. This matter was reviewed and determined to be contrary to Criterion 

I11 and Criterion V of 10 CFR Part 50, Appendix B. Nonconformance 

99900369/95-01-01 was identified in this area. 

As a result 

3.8 ASCO's Actions Pertaininq to Transmitted Information Resardinq NP8323 

In a letter dated August 23, 1989, ASCO informed ComEd's LaSalle station that 

it planned to discontinue offering NP8323 series valves as nuclear qualified 

valves and attached a copy of its field notification. 

October 2, 1989, concerning NRC's Information Notice 89-66 and recent problems 

with NP8323 valves at Grand Gulf Nuclear Station Unit 1, ASCO informed the NRC 

that it was forwarding the following documents to its nuclear customers: a 

letter explaining the reason for the mailing, ASCO's "Field Notification 

Concerning the Qualified Life of ASCO Catalog NP-1 Valves," and a "Revised 

Field Noti f i cat i on of Di scont i nuat i on of the NP8323 Val ve Li ne. I' These 

notifications provided additional information on the failures that occurred 

with NP8323 series valves and helped clarify ASCO's position and intention 

regarding discontinuation of its NP8323 valve line. 

In a letter dated 

On March 16, 1995, ASCO issued a service bulletin to all customers that had 

procured NP8323 SOVs in the 1989-1990 time period. The bulletin reiterated 

that ASCO recommended that its NP8323 SOVs be removed from licensees MSIV 

applications. 

plant planning to continue the use of the NP8323 reconsider this position." 

The service bulletin also stated that "ASCO recommends that any 

3.9 Observation of Assembly of NP 8323 Without Lubricants 

On March 14, 1995, in the presence of representatives from ASCO's quality 

control and LaSalle site engineering department, the NRC inspection team 

observed ASCO personnel assemble one of the four NP8323 SOVs intended for 

LaSalle. The various components for the assembly of the SOV had been 

retrieved from the storage bins and inspected by ASCO's quality control 

personnel. The inspection team observed that clean water instead of 

lubricants was used during the assembly and witnessed the successful 

operational tests performed on the SOV. 

3.10 Conclusions 

The team 

County S 

all play 

higher ambient temperatures; 

products that attract interna 

potential piping and external 

observations of ASCO's assemb 

personnel, the team concluded 

cannot attribute the sticking of the ASCO NP8323 SOVs at the LaSalle 

tation to any one aspect; but concluded that the following factors may 

a part in the SOV failure: less frequent cycling, in conjunction with 

as well as external contaminants; fretting; and 

nstrument air quality; internal oil or grease 

Val ve contaminants . Addi ti onall y , from 

y techniques and discussions with ASCO 

that ASCO's NP8323 assembly and testing methods 

9 

13 
in the past could have caused some excessive or inadvertently applied silicone 

products to be transferred to the core and plug nut sets during some of the 

fabrication or operational testing activities. 

4.0 PERSONS CONTACTED 

Automatic Switch Company 

M. Adase 

A. Byrne 

G. Laubenstein 

J. Shank 

K. Thomas, Jr. 

D. Tompsen 

G. Plechy 

R. Powell 

W. Prudden 

++ 

Commonwealth Edison Companv 

D. Drankham 

R. Tjerlund 44 

Assistant Engineer 

Val ve Service Manager 

Manager, Government/Nuclear Sales 

Director, New Product Design & Development 

Senior Service Engineer 

Director, Quality Assurance 

Senior Quality Control Engineer 

Senior Product Engineering Manager 

Qual i ty Control Supervisor 

LaSall e Site Support Engineering 

LaSall e Site Support Engineering 

++ Contacted by telephone 

10 

14 
.- ***** 

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 205550001 

June 9, 1995 

Dr. Richard S. Siudek, President 

Combustion Engineering, Inc. 

ABB Combustion Engineering 

Nuclear Operations 

P.O. Box 500 

1000 Prospect Hill Road 

Windsor, CT 06095-0500 

SUBJECT: NRC INSPECTION NO. 99900538/95-01 

Dear Dr. Siudek: 

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

Commission (NRC) inspection conducted April 7 through 10, 1995, 

at the Windsor, Connecticut, facility of Asea Brown Boveri (ABB) 

Combustion Engineering Nuclear Operations (CENO). The inspection 

was conducted by Robert Pettis, Jr., of the Special Inspection 

Branch; Jack Strosnider, Barry Elliot and John Tsao of the 

Materials and Chemical Engineering Branch, and Ernest Wilson of 

NRC Region I Office of Investigations. 

The purpose of the inspection was to review records and obtain 

data and information relevant to the assessment of reactor 

pressure vessel integrity. Information reviewed during the 

inspection included the review of the Reactor Vessel Group (RVG) 

data base, including supporting documents and quality assurance 

records, plant specific summary reports, licensee correspondence 

and ABB CENO prepared reports for the Combustion Engineering 

Owners Group. 

The inspection team held discussions with ABB CENO personnel and 

reviewed documentation relevant to activities performed by the 

RVG. On the basis of this inspection, the staff determined that 

ABB CENO’s activities, related to the assembly and evaluation of 

design and fabrication records used to support assessments of 

reactor pressure vessel (RPV) integrity, were in compliance with 

the ABB CENO Quality Assurance Program Manual for Service Related 

Activities (QAM-loo), Fourth Edition, Revision 2, and with the 

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

Federal Requlations. The enclosed inspection report contains a 

detailed discussion of the areas examined during the inspection. 

The inspection also identified that although ABB CENO has data 

that some licensees have not acquired for use in their RPV 

assessments, ABB CENO does not possess all the information that 

licensees should have to perform a thorough RPV evaluation. 

15 
R. Siudek -2- 

Original fabrication records, surveillance information and 

research results are in the possession of various vendors, 

utilities, institutes and laboratories. For some RPVs, 

particularly older vintage, it appears that no single vendor 

possesses all the information relevant to performing an 

assessment of RpV integrity. 

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

a copy of this letter and its enclosures will be placed in the 

NRC Public Document R%m. 

SmouPd you have any questions concerning this inspection, we will 

be pleased to discuss them with you. 

cooperation during this process., 

Thank you for your 

Sincerely, 

Robert M. 'Gallo, Chief 

Special Inspection Branch 

Division of Technical Support 

Office of Nuclear Reactor Regulation 

Docket No.: 99900538 

Enclosure: Inspection Report No. 99900538/95-01 

16 
ORGANIZATION: 

REPORT NO.: 

CORRESPONDENCE 

ADDRESS : 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF TECHNICAL SUPPORT 

ORGANIZATIONAL 

CONTACT : 

INSPECTION DATES: 

TEAM LEADER: 

INSPECTION TEAM: 

REVIEWED BY: 

APPROVED BY: 

Combustion Engineering, Inc. 

Asea Brown Boveri (ABB) 

Combustion Engineering Nuclear Operations 

99900538/95-01 

Dr. Richard S. Siudek, President 

Combustion Engineering, Inc. 

ABB Combustion Engineering Nuclear Operations 

P.O. Box 500 

Windsor, Connecticut 06095-0500 

Ron Fitzgerald, Director 

Nuclear Operations Quality Assurance 

April 7 through 10, 1995 

Robert L. Pettis, Jr.', P.E. Date 

Vendor Inspection Section (VIS) 

Special Inspection Branch (SIB) 

Division of Technical Support (DOTS) 

Office of Nuclear Reactor Regulation (NRR) 

Jack Strosnider, Team Manager, EMCB 

Barry Elliot, EMCB 

John Tsao, EMCB 

Ernest Wilson, Region I, 01 

v Gregorfld Cwalina, Chief 

Date 

SIB/DOTS/N 

Enclosure 

17 
1 

2 

INSPECTION BASES: 

e 

e 

a 

e 

Appendix A, "General Design Criteria for Nuclear Power 

Plants," General Design Criterion 10, Reactor Design, 

to Part 50 of Title 10 of the Code of Federal 

Reaulations (CFR) . 

Appendix B to 10 CFR Part 50, ##Quality Assurance 

Criteria for Nuclear Power Plants and Fuel Reprocessing 

Plants. 

10 CFR Part 21, ttNotification of Failure to Comply or 

Existence of a Defect." 

Asea Brown Boveri (ABB) Combustion Engineering Nuclear 

Operations (CENO), Topical Report, document CENPD-210A, 

Revision 7A, and ABB CE Nuclear Operations Quality 

Assurance Manual for Service Related Activities 

(QAM-loo), Fourth Edition, Revision 2. 

SUMMARY OF INSPECTION FINDINGS: 

No findings were identified during this inspection. 

3 STATUS OF PREVIOUS INSPECTION FINDINGS 

Previous inspection findings were not addressed during this 

inspection. 

4 INSPECTION FINDINGS AND OTHER COMMENTS 

4.1 Introduction 

On February 28 and March 6, 1992 (Revision l), the NRC issued 

Generic Letter (GL) 92-01,ttReactor Vessel Structural Integrity, 

10 CFR 5O.54(f),lt which was part of a program to evaluate reactor 

pressure vessel (RPV) integrity, ensure that licensees are 

complying with the requirements of 10 CFR 50.60 and 10 CFR 50.61, 

and are fulfilling commitments made in response to GL 88-11,"NRC 

Position on Radiation Embrittlement of Reactor Vessel Materials 

and its Impact on Plant Operations." 

All licensees responded to the GL and provided data relevant to 

the assessment of their RPVs. The results of the NRC staff's 

review of the GL responses are documented in letters to each 

licensee and in NUREG-1511, "Reactor Vessel Status Report," dated 

December 1994. 

records and information, determine if ABB CENO has data or 

information relative to the assessment of RPV integrity that had 

not been submitted to the NRC as part of the GL responses and to 

determine the extent to which data were shared between ABB CENO 

and licensees. 

The purpose of the inspection was to review 

-2- 

18 
4.2 Company Backqround 

ABB CENO was formed in October 1993, when ABB Combustion 

Engineering Services and ABB Combustion Engineering Nuclear Fuel 

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

reorganization, Engineering and Development in both of the former 

organizations were combined into a new organization, Engineering 

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

within the ABB CENO organization include: 

Total Quality 

Plant Equipment 

Engineering Operations 

Field Services 

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

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

ABB Atom is a subsidiary of Asea Brown Boveri AB of the ABB Asea 

Brown Boveri Group. 

ABB CENO (Engineering Operations) provides engineering, analysis 

and operations services to the nuclear power utilities. All 

nuclear safety-related activities are performed using controlled 

procedures that meet the requirements of Appendix B to 10 CFR 

Part 50, NQA-1 and ABB CENO Quality Assurance Program Manual 

(QAM-loo), which is supported by implementing procedures 

contained in (QAM-101). In addition, ABB CENO holds an American 

Society of Mechanical Engineers N, NA and NPT certificates and a 

National Board NR certificate. 

4.3 Entrance and Exit Meetinss 

During the entrance meeting in Windsor, Connecticut, on April 7, 

1995, the inspection team met with members of ABB CENO 

(Engineering Operations) management and staff, discussed the 

scope of the inspection and established contacts. The persons 

who participated in the inspection are listed in Appendix A of 

this report. 

the NRC on April 6, 1995, is shown in Appendix B. The inspection 

team conducted an exit meeting with ABB CENO staff and management 

on April 10, 1995. 

Information provided electronically to ABB CENO by 

4.4 Identification of Available Information 

The information sought by the NRC staff in connection with the 

assessment of RPV integrity was in the form of a data base 

maintained by the Reactor Vessel Group (RVG) of ABB CENO. The 

RVG was chartered in 1991 by a group of 14 member utilities 

operating with ABB CENO vessels. 

utilities and are licensees for 23 RPVs. There are 70 operating 

reactors with RPVs that were fabricated by ABB CENO. The data 

base contains most of the data relevant to domestic operating 

The 14 members are all domestic 

-3- 

19 
reactors because it contains information on non-member sister 

vessels. ABB CENO was the original equipment manufacturer for 

approximately one-half of the operating reactor pressure vessels 

in the United States. 

The objective of the RVG was to develop a verifiable data base of 

RPV information such as a description of the processes, 

procedures, chemistries and material properties for the welds, 

plates, forgings and cladding used in the fabrication of the RPVs 

of the member utilities. The assembly and evaluation of the data 

was performed by personnel at the ABB CENO Southeast Nuclear 

Service Center (SNSC), located in Chattanooga, Tennessee. Over 

450 boxes of original vessel fabrication records and drawings 

were reviewed. The documents included material certifications, 

procedures, specifications, fabrication, laboratory log book 

entries, inspection and test records. Due to the complexity and 

technical nature of these records, several former SNSC employees 

were hired to participate in the interpretation and evaluation of 

this information. ABB CENO fabricated RPVs, steam generators and 

pressurizers for domestic utilities at this facility. 

ABB CENO collected the records for the fabrication of the RPVs 

for the 14 member utilities. Additional data was collected from 

the fabrication records of steam generators and pressurizers, if 

the steam generators and pressurizers were fabricated by ABB 

CENO. Data was also collected from non-member RPVs, steam 

generators and pressurizers, if the data was applicable to the 

member plant's RPV. As a result, the RVG data base consists of 

information from about 20 percent of the 450 boxes; however, it 

represents a larger percentage of the data of greatest interest. 

The data base does not contain all surveillance data from ABB 

CENO fabricated RPVs. The last RPV material surveillance capsule 

reviewed by ABB CENO was in 1983. Since the data base does not 

contain all surveillance data and does not contain all the RPV 

data for non-member plants, it is not a complete data base. 

The RVG data base program consisted of two phases. Phase I 

identified the subtasks necessary to prepare a detailed 

evaluation of the records. Phase I1 evaluated the records, which 

included abstracting critical data from each record selected, 

entered the data into the Global Data Matrix (GDM), verified the 

input data and provided identification of sister vessel 

relationships. Vessel-specific records were reviewed and 

prepared for each nuclear utility who participated in the 

project. The Phase I Final Report was issued by ABB CENO in 

August 1992. The Phase I1 Final Report is scheduled for issuance 

in June 1995. At the conclusion of Phase 11, ABB CENO will 

provide each member with the results of the evaluation for each 

RPV owned by that member. 

-4- 

20 
4.5 Computerized RVG Data Base 

4.5.1 Data Abstraction Process 

As stated previously, the information which was input into the 

RVG data base was abstracted from the original fabrication 

records and documents available in Chattanooga, Tennessee. 

Relevant information was extracted from each document and then 

recorded on data entry forms which were assigned unique RVG 

numbers. 

for accuracy by a reviewer. The data entry forms were then 

signed by the inspector, clerk and reviewer. Each original 

document and the corresponding data entry forms identified by the 

RVG numbers, were placed in a folder prior to being input into 

the data base. 

The forms were then typed by a clerk and were verified 

4.5.2 Description of Data Base 

The RVG data base consists of 12,000 data sheets and computer 

files which have been input into the GDM. The software system 

used for storing and manipulating information in the data base is 

Paradox Runtime 4.5, a relational data base program. The data 

base consists of 54 primary, secondary and support data files. 

To review data stored in these files, the operator must link the 

various files that contain the needed information. Hence, the 

operator must have a working knowledge of Paradox including the 

data contained in the data files. 

4.5.3 Quality Assurance of RVG Data Base 

4.5.3.1 Procedures Review 

The NRC inspection team reviewed the QA activities related to the 

preparation of the RVG data base. 

Manual (QAM) for Service Related Activities, QAM-100, Fourth 

Edition, Revision 2, December 21, 1994, is the upper tier quality 

manual which governs the activities of ABB CENO Nuclear 

Operations. The implementing procedures required under QAM-100 

are contained in ABB CENO Nuclear Services QA Procedures Manual, 

QAM-101, Revision 10, November 8, 1993. Quality Assurance 

Procedure (QAP) QAP 1.1, Revision 2, requires preparation of a 

Quality Plan for each project related to the supply of Quality 

Class 1 items or services, as defined in QAP 3.1. The plan 

provides specific information concerning the application of the 

ABB CENO quality program to the specific project. 

ABB CENO Quality Assurance 

The inspection team reviewed ABB CENO Integrated Project Quality 

Plan (IPQP), "Reactor Vessel Group Program for Assembly and 

Evaluation of Design and Fabrication Information,tt (Revision 6, 

dated January 10, 1995), which specified that experienced 

personnel, familiar with the document contents, materials, and 

processes used during vessel manufacture, be selected to 

-5- 

21 
participate in this program. ABB CENO Specification MISC-MECHSP-020, "Specification for Phase I of the RVG Program," Revision 

1, November 4, 1992, was prepared for Phase I of the RVG Program 

Plan and stated objectives and requirements for various subtasks 

of the program. ABB CENO Specification MISC-MECH-SP-021, 

t@Specification for Phase 11, Detailed Records Evaluation,*I 

Revision 0, November 4, 1992, was prepared for evaluation under 

Phase I1 of the project. As stated in Section 3.8 of the IPQP, 

the provisions of 10 CFR Part 21 only apply to Phase I1 of the 

project . 

Other procedures prepared by ABB CENO to accomplish various 

program tasks included RVG-PROC-92-001, "Records Rough Sorting 

and Assembly,*1 Revision 1, November 4, 1992, which was prepared 

to guide the rough sorting and inventorying effort of Subtask 2, 

and RVG-PROC-92-002, "Abstracting of Data, Data Entry and Data 

Matrix Verificationttl Revision 0, November 5, 1992, which 

provided guidance for preparation and verification of data into 

the GDM to ensure it performed consistently. 

contained over 25 data forms intended to simplify the abstraction 

and data entry process for development of the GDM. 

The procedure 

The procedure also included generic and plant-specific records 

inventory forms and document trees which show the 

interrelationship between the documents and assists in 

determining data traceability. Each form contained signature and 

date blocks for the abstractor and the data entry personnel along 

with the independent verifier for both activities. 

4.5.3.2 Verification and Validation of Paradox Computer Code 

Section 3.7, IIQA'd Computer Codes,Il of the IPQP requires 

certification of the Paradox computer code used to execute the 

RVG data base. Reference 2.6.1 of QAP 3.13, "Computer Software," 

Section 4.7.2(a) requires verification and validation of 

nonsafety-related, commercially procured software where the 

results are used for safety-related applications. 

The inspection team reviewed the final report prepared by ABB 

CENO to verify and validate the latest version of the code. The 

report, ABB-CENO MISC-PENG-CALC-025, "Paradox and Paradox Runtime 

Version 4.5,Il February 8, 1995, contained documentation to 

support verification and validation of the code. Input/output 

functions for Versions 4.02 and 4.5 of the DOS version of Paradox 

were previously verified, validated and documented in ABB CENO 

MISC-MECH-CALC-157, Revision 0, January 27, 1994. 

The methodology employed to accomplish the verification and 

validation of Runtime Version 4.5 was to print five existing 

records from the RVG data base and check that the information 

matched that on the input forms. This appeared adequate to the 

inspection team since the program's only function is to perform 

-6- 

22 
input/output related functions, not design calculations. 

Additionally, the same input forms were used to generate new 

records in the database which were printed and compared to the 

output from the existing records. This verified that the Paradox 

output matched that which was entered into the program. The file 

contained a completed copy of 'IChecklist #3," February 7, 1995, 

which is required by the QAM. 

4.5.3.3 Indoctrination and Training of RVG Personnel 

Section 6.0, "Training and Qualification Requirements," of the 

IPQP references the requirement of QAP 2.1, "Indoctrination and 

Training,lI for training of personnel involved in the preparation 

or independent review of safety-related documentation. As stated 

previously, the abstraction, data entry preparation and 

independent verification phase of this program was performed at 

the SNSC facility. 

The NRC inspection team reviewed the list of SNSC personnel, 

qualified as of August 1, 1994, which identified 12 persons and 

their level of qualification. The list is required to be 

maintained in Windsor pursuant to Paragraph 6.2.4, Revision 5, of 

the IPQP. Actual training records of the individuals on the list 

were not reviewed during the inspection. 

4.5.3.4 ABB CENO Performed Internal Audits 

The inspectors reviewed ABB CENO Audit Report No. ME-1093-02, 

November 12, 1993, which documented an internal audit performed 

of Mechanical Engineering, RVG Project. The purpose of the 

audit, conducted on October 4-6, 1993, at both the Chattanooga 

and Windsor facilities, was to verify compliance to the ABB CE 

Nuclear QA Program, as implemented by QAM 100, Revision 1, QA 

Procedures Manual, QAM 101, Revision 9 and RVG IPQP No. MISCMECH-IPQP-030, Revision 2. 

The audit identified several conditions which required corrective 

action. These included the lack of documentation with respect to 

the verification and validation of the computer program Paradox, 

Versions 4.0 and 4.02, and definition of the training and 

experience requirements for personnel performing data extraction 

and independent verification activities. As a result, Corrective 

Action Request (CAR) CAR No. AC-93-134, October 28, 1993, was 

generated to track closure of the items identified. ABB CENO 

Inter-Office Correspondence No. RVG-93-073, December 15, 1993, 

provided a response to the CAR and accepted its findings. 

Corrective actions, which included revisions to the IPQP to 

address the training and experience issues (Revision 4), were 

completed on March 4, 1994. Verification and validation of 

Paradox for its initial use was documented in MISC-CALC-057, 

January 27, 1994, and independently verified on February 3, 1994. 

-7- 

23 
The NRC inspection team reviewed a second internal audit 

performed by ABB CENO on August 16-19, 1994. Audit Report No. 

OE/PENG(RVG)-0894-04 documented an audit of Engineering 

Operations, Performance Engineering and the RVG, and was 

performed at both the SNSC and Windsor facility. The purpose of 

the audit was to verify compliance to the ABB CENO QA Program, as 

defined by the IPQP. 

The audit reviewed the abstraction and transposition, including 

independent review, of archival reactor vessel records, data base 

entries including verification thereof and subsequent analysis 

and compilation of the data into various reports for submittal to 

the participants of the RVG. No findings were identified during 

the audit. 

4.5.3.5 External Audit of ABB CENO 

The inspectors reviewed a joint audit report of ABB CENO which 

was performed by the Nuclear Procurement Issues Committee (NUPIC) 

on October 25-29, 1993. The audit, lead by Union Electric, and 

representatives from Virginia Power, Entergy, Yankee Atomic, 

Consolidated Edison, Consumers Power and Illinois Power, was 

documented in NUPIC audit report No. AG93-20, November 16, 1993. 

The audit covered the supply scopes of ABB CENO's four major 

business units and included spare parts, commercial grade items, 

engineering services, design services and outage services. With 

respect to the activities associated with the RVG program, two 

observations were identified by the NUPIC inspection team which 

had been previously identified by ABB CENO during its internal 

audit discussed in Section 4.5.4.4 above. 

The first observation identified that personnel performing the 

review of RVG data collection were not qualified as independent 

reviewers, as required by the IPQP. Since this requirement was 

viewed by ABB CENO as overly restrictive, corrective actions 

included revising the IPQP to delete this requirement. 

The second observation identified that the input/output 

capabilities of Paradox, the commercial software program used by 

Performance Engineering to access the RVG data base program, had 

not been validated under the requirements of the IPQP. 

Performance Engineering committed to track the item under CAR No. 

AC-93-134, October 28, 1993, which was acceptable to the audit 

team. As stated previously in Section 4.5.4.2 above, Paradox was 

verified and validated in January 1994. 

4.5.4 Current Status of the RVG Data Base 

The RVG computer files are currently in draft status and 

currently under review to assure that they contain all the 

information applicable to the member utilities. In addition to 

-8- 

24 
the electronic data base, ABB CENO will be sending reports and 

data sheets, that contain all of the information applicable to 

its RPVs, to each of its member utilities. 

4.6 Plant Specific Summary Reports 

A report was prepared for each of the reactor units whose owners 

are participants of the RVG. The NRC inspection team requested 

ABB CENO to provide copies of the Maine Yankee, Fort Calhoun and 

Kewaunee RVG reports. 

the RVG, only the Maine Yankee and Ft. Calhoun reports were 

available for review. Each RVG report summarized material data 

for plates, forgings, welds and claddings and contained an 

Introduction, Vessel Overview, Vessel Materials and Sister Vessel 

Analysis section. 

Since Kewaunee was not a participant of 

The Introduction section provided a general background on 

fabrication methods and specifications of plates, forgings, 

welds, and cladding. The Vessel Overview section described the 

RPV materials and assembly procedures used during manufacture and 

included RPV diagrams. The Vessel Materials section discussed in 

detail the fabrication of the plates, forgings, welds and 

claddings and included the properties and chemical compositions 

of RPV materials. The Sister Vessel Analysis section provided 

material data from fabricated RPVs, steam generators and 

pressurizers that have the same heat number as the RPV materials 

listed in the RVG report. This section also identified selfsister materials (materials from the same heat number used at two 

or more locations in the same reactor vessel) and the 

availability of archived materials in sister plants. 

The NRC inspection team identified some discrepancies in the data 

base. 

heats (e.g., W-5214) that the inspectors knew was available. 

For example, there was no chemistry data for certain weld 

4.7 Combustion Enqineerinq Owners Group (CEOG) Reports 

The NRC inspection team reviewed a topical report prepared by ABB 

CENO and the CEOG on reactor vessel integrity. The report, 

"Quantitative of Evaluation of Chemical Composition Variability 

in Reactor Pressure Vessel Welds,tt CE NPSD-944-P, presented test 

results and statistical analysis of the results of several 

archived surveillance specimens. The purpose of the report was 

to determine the range of chemical composition variability that 

could typically be expected for ABB CENO fabricated RPV beltline 

welds. 

composition. 

The review identified that chemistry results of the test 

specimens were within the expected values except for weld heat 

number 1P3571, flux 1092. The weld, heat number 1P3571, showed 

an average value of copper content of 0.363 percent with a 

Twelve archived weld seams were tested for chemical 

-9- 

25 
standard deviation of 0.075 percent. This chemical composition 

showed a higher mean copper content and variability for the 

1P3571 weld than had been previously reported to the NRC. 

However, the data were relatively new having been generated in 

1994. The chemistry data in CE NPSD-944-P had been provided to 

the licensees that contributed to developing the data. The 

report had not been provided to licensees that had not 

participated in the project. 

4.8 Vendor/Licensee CorresDondence 

The inspectors reviewed correspondence between ABB CENO and the 

Kewaunee and Ft. Calhoun plants to verify that the data in the 

RVG data base was consistent with the data transmitted by ABB 

CENO and whether relevant sister plant data existed but was not 

shown in the data base or the RVG report. A review of ABB CENO 

correspondence to Kewaunee on December 4, 1985 and April 15, 

1988, noted that Kewaunee was informed that the chemistry in weld 

heat number 1P3571, for the Kewaunee RPV, may have a higher 

copper content than that previously reported by Kewaunee to the 

NRC. ABB CENO stated in the correspondence that the copper 

content could be as high as 0.31 percent and the nickel content 

as high as 0.78 percent. 

As a result of the recent CEOG study reported in ABB CENO report 

CE NPSD-944-P, ABB CENO informed Kewaunee, on April 6, 1995, that 

a weld, fabricated using weld wire heat 1P3571, had an average 

copper content of 0.363 percent with a standard deviation of 

0.075 percent. The inspectors also confirmed that the copper 

data that led to an average copper value of 0.363 percent was 

included in the RVG data base. 

After reviewing the Ft. Calhoun correspondence file, the 

inspectors determined that ABB CENO informed Ft. Calhoun that a 

weld, fabricated using tandem electrodes with heat numbers of 

12008 and 27204, had a copper content of 0.27 percent. 

response to NRC GL 92-01, Ft. Calhoun reported a copper content 

of 0.19 percent for the same weld heats. The inspectors were not 

able to confirm if the higher copper content was included in the 

data base the licensee used to determine the reported 0.19 

percent value. 

In 

4.9 MRC Information Obtained 

The NRC inspection team requested copies of certain information 

considered to have potential safety significance relative to RVG 

integrity assessments. This information, considered proprietary 

by ABB CENO, included material which will be used by the NRC 

staff to evaluate plant-specific information requested by the 

staff pursuant to GL 92-01, "Reactor Vessel Structural Integrity, 

10 CFR 50.54(f),It Revision 1, March 6, 1992. 

- 10 - 

26 
With the exception of a copy of the proprietary RVG data base, 

which was provided to the inspection team at the exit meeting on 

April 10, 1995, the information requested during the inspection 

was provided to the NRC under separate cover on April 13, 1995. 

It included an affidavit which stated that the material contained 

in certain documents is considered proprietary and requested that 

it be withheld from public disclosure in accordance with the 

provisions of 10 CFR 2.790. 

4.10 Summary 

An important observation of the inspection is that although ABB 

CENO has data that some licensees have not acquired for use in 

their RPV assessments, ABB CENO does not possess all the 

information that licensees should have to perform a thorough RPV 

evaluation. Original fabrication records, surveillance 

information and research results are in the possession of various 

vendors, utilities, institutes and laboratories. In summary, for 

some RPVs, particularly older vintage, it appears that no single 

vendor possesses all the information relevant to performing an 

assessment of RPV integrity. 

, 

- 11 - 

27 
APPENDIX A 

PERSONS CONTACTED 

ABB Combustion Enaineerina. Inc. 

e .* 

.A 

e* 

e 

e* .* .* 

* 

Curtis, P. 

Leombruni , P. 

Ghergurovich, J. 

Fitzgerald, R. 

Lizotte, C. 

Toelle, S. 

Byrne, S. 

Probst, V. 

Bischoff, G. 

Gimbrone, C. 

Brinkman, C. 

Vice President, Engineering Operations 

Manager, Plant Components, Mechanical 

Project Engineer 

Director, CENO Quality Assurance 

Sr. Quality Assurance Engineer 

Manager, Nuclear Licensing 

Consultant, Primary Systems 

Attorney 

ABB CEOG Representative 

Supervisor 

Manager, Washington Nuclear Operations 

U.S. Nuclear Reaulatorv Commission 

Pettis, R. 

e* Strosnider, J. .* Elliot, B. 

Tsao, J. .* Wilson, E. 

Sr. Reactor Engineer, Team Leader, NRR 

Branch Chief, EMCB, NRR 

EMCB 

EMCB 

Investigator, Region I, 01 

0 Attended entrance meeting * Attended exit meeting 

A- 1 

28 
APPENDIX B 

INFORMATION PROVIDED TO ABB COMBUSTION ENGINEERING, INC. 

PRIOR TO THE ONSITE INSPECTION 

Prior to the start of the NRC inspection on April 7, 1995, the 

staff electronically forwarded the following information to ABB 

Combustion Engineering, Inc., on April 6, 1995. The purpose was 

to allow ABB CENO management and staff additional time to provide 

responses to the inforrnation requested. 

INSPECTION SCOPE 

1. General Ouestions 

During the entrance meeting have personnel available to address 

the following questions: 

e Is all the data from the boxes represented in the computerized 

data base? 

What data in the boxes is in the data base and what data in the 

boxes is not in the data base? 

What is the schedule for completing the review of the boxes and 

placing the data in the data base? 

0 Describe the process of how the hard copy information and 

documentation in the boxes was assimilated into the computerized 

data base and how the information in the boxes is referenced. 

0 Which licensees are aware of the data and which licensee's have 

completed a review of the data? 

Do non-member plants know that this type of information available 

for their plants? 

0 

2. General Data Reauest 

Provide the entire computerized RVG data base to the NRC with a users 

manual. Make the data base available during the inspection for review 

by NRC. 

available (including foreign plants) and provide the following data: 

Identify all plants for which the following information is 

(a) Provide to the NRC a list of weld wire heat numbers used to 

fabricate the beltline welds for each commercial nuclear reactor 

vessel fabricated by ABB CENO, 

B- 1 

29 
For each heat of weld wire identified in item (a), provide to the 

NRC all copper and nickel chemistry data and the sources of those 

data (including foreign reactor data), 

For each heat of weld wire identified in item (a), identify 

archival material available at ABB CENO and indicate its 

relationship to either surveillance material or ABB CENO 

fabricated beltline welds, 

Provide to the NRC any data that demonstrates the variability of 

copper within welds fabricated using a single heat of weld wire 

or within a weld, in which the welds were fabricated by ABB CENO 

using a submerged arc process and copper coated primary 

electrodes, 

Provide to the NRC any data that demonstrates the variability of 

nickel within welds fabricated using a single heat of weld wire 

or within a weld, in which the welds were fabricated by ABB CENO 

using a submerged arc process and the primary source of nickel 

from the primary electrodes, 

Provide to the NRC any data that demonstrates the variability of 

nickel within welds fabricated by ABB CENO using a submerged arc 

process and the primary source of nickel from a cold nickel wire. 

Describe the traceability of the data in (a) and (b) above to the 

boxes of stored information/data. 

3. SPecific Data Review 

Have available for review the boxes of information/data and 

computerized database for the following sample of 

plants/reactor vessels. 

0 Kewaunee and its sister plants 

0 Fort Calhoun and its sister plants 

Provide the number of coils of weld wire represented by the copper and 

nickel measurements for the weld data for Kewaunee, Fort Calhoun and 

their sister plants. (Note: Sister plants- are those plants with 

reactor vessels fabricated by ABB CENO with the same heats of weld 

wire as Kewaunee and Fort Calhoun). Be prepared to provide to the NRC 

selected information from the boxes of information/data as identified 

by the NRC inspectors during the inspection. 

Note: Where information/data is requested to be provided to the NRC, 

these data and information should be made available in a form such 

that NRC can arrange to take possession of the subject information and 

transport it to NRC headquarters in Rockville, Maryland. 

B-2 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-OOOl 

' +**** 

May 31, 1995 

Hr. Ronald J. Woodward 

Director Total Qual i ty & Supply 

Dow Corning U.S.A. 

Dow Corning Corporation 

Midland, MI 48686-0994 

SUBJECT: NRC Inspection No. 99901284/95-01 

Dear Mr . Woodward : 

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

(NRC) inspection of Dow Corning Corporation (Dow Corning) facilities in 

Midland, Michigan, and El izabethtown, Kentucky, conducted by Messrs. P. Kol tay 

and J. Petrosino of this office on March 23-24, 1995, and April 3-5, 1995. 

The report also addresses the NRC staff observations on March 21-23, and 

April 18-21, 1995, of activities that Dow Corning performed to investigate its 

concern regarding cure time and flame sel f-extingui shing time parameters of 

its silicone products that are used in nuclear power plant fire barrier 

appl i cat i ons . 

The inspection was conducted to evaluate Dow Corning's qual ity assurance 

program and its implementation in selected areas such as (1) manufacturing 

process control, (2) product design control, (3) procurement document control, 

and (4) a review of its program for implementing Part 21, "Reporting Defects 

and Noncompliance," of Title 10 of the Code of Federal Requlations (10 CFR 

Part 21). 

As part of the inspection the inspectors examined procedures and 

representative records, conducted discussions and interviews with personnel, 

and observed several of Dow Corning silicone product manufacturing and testing 

processes. 

staff at the conclusion of the inspection activities, on April 5, 1995. 

The inspectors observations were discussed with the Dow Corning 

Based upon the NRC staff observations and discussions during this review, it 

was concluded that the Dow Corning silicone products used in nuclear power 

plant fire barrier applications and which Dow Corning manufactures at its 

El izabethtown, Kentucky facility are comercia7 grade items, as defined in 

10 CFR 5 21.3. 

products supplied for nuclear plant fire barrier applications are not subject 

to the NRC's Part 21 regulation, it appears to have adequately established and 

implemented a system which meets the intent of certain 10 CFR Part 21 

requirements, such as 10 CFR tj 21.21(b). Areas examined during the NRC 

activities and our conclusions are discussed in the enclosed inspection 

report. 

The team also concluded that although Dow Corning's silicone 

31 
Mr. Ronald 3. Woodward 2 

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. 

we will be pleased to discuss them with you. No response to this letter is 

requi red. 

If there are any questions concerning this inspection 

Sincerely, 

Robert M. Gallo, Chief 

Special Inspection Branch 

Division of Technical Support 

Office of Nuclear Reactor Regulation 

Docket No. 99901284 

End osure: Inspection Report 99901284/95-01 

cc: Mr. James G. Sharpe 

El izabethtown Site Manager 

Dow Corni ng Corporat i on 

760 Hodgenvi 11 e Road 

Elizabethtown, KY 42701 

Mr. Robert M. Schroeder 

Fire Protection Engineer 

Dow Corning Corporation 

2200 West Salzburg Road 

Mail C043A1 

Auburn, MI 48611 

Mr. Christopher J. Johnson 

Engineering Services 

Underwriters Laboratories Incorporated 

333 Pfingsten Road 

Northbrook, Illinois 60062-2096 

32 
ORGANIZATION: 

REPORT NO. : 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY: 

INSPECTION CONDUCTED: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF TECHNICAL SUPPORT 

LEAD INSPECTOR: 

OTHER INSPECTORS: 

REV I EWED : 

APPROVED : 

Dow Corning U.S.A. 

Dow Corning Corporation 

Midland, Michigan 48686-0994 

99901 284/95-01 

Mr. Robert M. Schroeder, Fire Protection Engineer 

(517)496-8330 

Manufactures commercial grade silicone products used 

in fire barrier penetration seal assemblies for the 

nuclear industry. 

customers who either supply and/or develop and test 

the fire barrier penetration configurations before 

installation at nuclear power plants. 

The products are sold to six 

March 23-24; April 3-5, 1995, and April 18-21, 1995 

-&Sq k. 7 

Peter S. Koltay, Team Leader 

Special Inspection Section 

Special Inspection Branch (TSIB) 

J. J. Petrosino 

Vendor Inspection Section, TSIB 

E. Connell , Plant Systems Branch 

Division of Systems Safety and Analysis 

T.L. Tinkel, Brookhaven National Laboratory 

Speci a1 Inspecti on Branch 

Division of Technical Support 

33 
1 INSPECTION SUMMARY 

Based on the scope of this inspection, the team determined that the Dow 

Corning Corporation (Dow Corning) does not manufacture and supply any "basic 

components" as defined in Part 21, "Reporting Defects and Noncompliance," of 

Title 10 of the Code o f Federal Requlations (10 CFR Part 21). While some 

customers contractually imposed the requirements of 10 CFR Part 21 on Dow 

Corning for silicone products supplied for use in nuclear power plant fire 

barrier penetration seal applications, the inspectors verified that the 

products met the definition of commercial grade items, as defined in 10 CFR 

9 21.3. 

requirements of 10 CFR Part 21. However, the inspectors found that Dow 

Corning has established and implemented a program which meets the intent of 

certain sections of 10 CFR Part 21, such as 10 CFR 5 21.21(b). Dow Corning 

personnel stated to the team that although the inspectors determined that Dow 

Corning is not subject to the requirements of Part 21, it will continue'its 

existing program for meeting the intent of 10 CFR Part 21, but will 

discontinue issuing certifications to its customers indicating that Dow 

Corning products were manufactured in compliance to 10 CFR Part 21. 

Therefore, the team concluded that Dow Corning is not subject to the 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

This was the first NRC staff inspection which was conducted at either of the 

two Dow Corning facilities discussed in this report; thus, there were no 

previous findings. 

3 

3.1 

INSPECTION FINDINGS AND OBSERVATIONS 

Entrance and Exit Meetinqs 

The entrance meeting on March 23, 1995, was held at the Dow Corning Corporate 

headquarters in Midland, Michigan. The meeting was attended by corporate 

personnel responsible for the product line. The exit meeting on April 5, 

1995, was held at the Elizabethtown, Kentucky manufacturing facility w th the 

plant manager and the manufacturing team responsible for the product 1 ne in 

attendance. 

product applications participated in both meetings. 

3.2 Backqround and Description of Facilities 

A staff attorney and the fire protection engineer respons ble for 

Dow Corning manufactures four products that have commercial applications and 

are typically used in fire barrier penetration seals at nuclear power plants, 

(1) Sylgard 170 Silicone Elastomer, (2) 3-6559 Cure Accelerator (typically 

used in conjunction with Sylgard 170), (3) 3-6548 Silicone RTV Foam, and 

(4) 96-081 RTV Adhesive/Sealant (typically used in conjunction with 3-6548 RTV 

foam). These products, available through published product description, are 

sold by Dow Corning to commercial customers including six customers who either 

supply or facilitate the installation of the materials at NRC licensees as 

part of fire barrier penetration seal assemblies associated with 10 CFR 

§ 50.48, "Fire Protection." The six customers who procure the silicone 

products from Dow Corning are (1) Brand Fire Protection, (2) North Brothers 

Company, (3) Promatec, (4) PCI-ICMS, (5) Radian Corporation, and (6) Transco 

Products, Incorporated. 

2 

34 
By letter dated November 17, 1994, Dow Corning informed the NRC and their 

customers of a deviation which it had identified regarding sales 

specifications for flame self-extinguishing time of certain lots of RTV 

Adhesive Sealant 96-081. The letter to the NRC stated that certain customers 

were being advised of a sales specification deviation on flame selfextinguishing time for 96-081 RTV Adhesive/Sealant. Dow Corning further 

stated that this product was supplied to certain customers under 10 CFR 

Part 21. 

on retained samples from three lots of the product showed a flame selfextinguishing time ranging from 35 to 130 seconds. 

tests on these lots at the time they were originally produced showed a flame 

self-extinguishing time that met the sales specification of 10 seconds or 

1 ess. 

In the letter sent to the customers, Dow Corning advised that tests 

According to Dow Corning, 

By letter dated November 28, 1994, Dow Corning informed the NRC and their 

customers of another deviation that it had identified regarding sales 

specifications for cure time of Sylgard 170 silicone elastomer. The letter to 

the NRC stated that certain customers of Sylgard 170 Elastomer and Sylgard 170 

Fast Cure Elastomer were being advised that the cure rate for certain lots 

which they had received from Dow Corning was slower than the sales 

specification. 

to certain customers under 10 CFR Part 21. In the letter sent to the 

customers, Dow Corning advised that tests on retained samples of certain lots 

of material manufactured between May 1994 and November 23, 1994, indicated 

that the silicone products did not meet the sales specification for cure rate. 

According to Dow Corning, tests on these lots at the time the product was 

manufactured indicated the lots did meet the sales specification for cure 

rate. 

to an ingredient used in Part A of these materials. 

stated that tests on retained samples of Daw Corning 3-6548 RTV Silicone Foam 

manufactured within the last year showed that this material met specifications 

and was satisfactory. 

products to their customers. 

Dow Corning further stated that these products were supplied 

Dow Corning stated their belief that the slow cure rate was traceable 

Dow Corning further 

Dow Corning offered to provide replacement silicone 

3.3 Identification of Deviations to Sales Specification 

3.3.1 Initial Product Performance Indicators and Investiqation 

In June-July 1994, Dow Corning received reports from a commercial electronics 

customer indicating that it had observed slower cure times of the silicone 

products in its electronic component applications. 

customer notification, Dow Corning started an investigation to resolve the 

customer report. Dow Corning started its investigation by testing retained 

samples and determined cure times were slower than normal, but the cure times 

As a result of the 

'At the Dow Coming Elizabethtown. Kentucky silicone product manufacturing facility, it was observed 

by the HRC inspectors that Dow Coming typically tests each silicone product lot for its specification 

characteristics for baseline data, and also obtalns and stores samples of each silicone product lot for future 

analysis as required. 

3 

35 
were still within specifications. In early August 1994, Dow Corniig received 

reports from another electronics customer regarding slow snap time on 

Sylgard 170 fast cure. 

applicable retained lot specimens and confirmed that the snap time was slower 

than as stated in the specification. 

Consequently, Dow Corning performed tests on the 

In mid-August 1994, additional testing by Dow Corning identified that the cure 

rate of Sylgard 170A was changing as the product aged, but the cure rate was 

still within specifications. In late August-September 1994, Dow Corning 

identified that a certain ingredient used in the Sylgard 170 Part A component 

appeared to be related to the slower cure time observed in the Sylgard 170 and 

the slow snap time observed in the Sylgard 170 Fast Cure. 

variability in cure time from lot to lot was observed. In September 1994, Dow 

Corning observed that cure rate increased if the mixing time for the 

ingredient was increased. 

process during the mixing of the ingredient into the Part A compound also 

improved the cure rate. This latter observation indicated to Dow Corning that 

possible contaminants were affecting the cure rate. In October 1994, Dow 

Corning received additional reports from customers about the slow cure time of 

Sylgard 170, and subsequent testing on retained samples for six lots of the 

Sylgard 170 now showed that the cure time on two lots were out of 

speci f i cat i on. 

Significant 

Dow Corning also determined that a certain vacuum 

3.3.2 Dow Corninq Investiqation Findinqs and Corrective Actions 

As a result of inquiry letters that Dow Corning had sent out earlier to its 

suppliers, Dow Corning received a November 1994 letter from the supplier of 

the suspect ingredient stating that certain lots of the ingredient contained 

higher than typical levels of sulphur. Sulphur is a known contaminant that 

can affect the cure rate and flame self-extinguishing time of the products. 

Dow Corning also noted that it had not specifically required the supplier to 

monitor or control sulphur content in that ingredient. In November 1994, Dow 

Corning concluded that the slower cure rate of Sylgard 170 was related to the 

sulfur content of the ingredient used in the mixture and advised their 

customers and the NRC of the problem. 

that varying levers of sulphur in that ingredient appeared to correlate with 

the variations in Sylgard 170 cure rate. 

a new manufacturer that was capable of supplying the ingredient with higher 

purity and less potential for sulphur contamination. 

Additionally, Dow Corning confirmed 

As a result, Dow Corning switched to 

As part of its corrective action plan Dow Corning, in conjunction with 

Underwriters Laboratory, Incorporated (UL) , engineering services staff from 

UL's Northbrook, Illinois facility developed a fire endurance test program of 

its silicone products used at nuclear power plants. Dow,Corning contracted UL 

to perform and monitor the fire endurance test program to determine if the 

manufacturing changes, that resulted in the deviations reported to the NRC and 

'Snap time is a certain point in the curing process of Sylgard 170 elastomer and 3-6548 Foam just prior 

to tho rolidification of the silicon. product. 

4 

36 
licensees (self-extinguishing time and cure rate)', had any significant affect 

upon the material's performance as a fire barrier penetration seal. The test 

program was devel oped to encompass the product sales specification parameters for each product. 

period as well as post and pre-suspect time periods for testing. 

testing was set up to test multiple configurations of each lot. For example, 

the tests at UL included four different lots of Sylgard 170 Elastomer that met 

Dow Corning's sales specification and three different lots of the material 

that deviated from it's sales specifications. 

lot of 96-081 RTV Adhesive Sealant that met Dow Corning's sales specification 

and one lot of the material that deviated from it's sales specifications. 

test program was established to ensure that it encompassed the following 

aspects . 

Dow Corning obtained sample lots from the suspect time 

The UL 

The UL staff also tested one 

The 

0 

0 

Non-suspect lots of Sylgard 170 and RTV Adhesive/Sealant 96-081 & 3-6548 

Foam 96-081 produced with the old source of the suspect ingredient 

Suspect lots of Sylgard 170 RTV Adhesive/Sealant 96-081 & 3-6548 Foam 

ingredient 

and 96-081 which were produced with the old source of the suspect 

0 Lots of Sylgard 170 and RTV Adhesive/Sealdnt 36-081 & 3-6548 Foam 

produced with the new source of suspect the ingredient 

3.3.3 Fire Endurance Test at the UL facility in Northbrook. Illinois 

During the week of March 20, 1995, the team observed the fabrication of the 

Dow Corning fire endurance test specimens, and during the week of April 17, 

1995, the inspector witnessed the performance of fire endurance tests for Dow 

Corning's 3-6548 Silicone RTV Foam, 96-081 RTV Adhesive and Sylgard 170 

Elastomer. The inspector noted that the fire endurance test program was 

established and conducted in accordance with the requirements of ASTM E-119, 

"Standard Test Methods for Fire Tests of Building Construction and Materials." 

A total of five tests were witnessed by the inspector which encompassed all 

three types of materials from both the suspect group, manufactured between May 

1994 and November 1994, and newly manufactured materials. Dow Corning 

conducted a total of approximately 25 fire tests under this effort. 

variations or anomalies of significance were observed in the fire barrier 

performance of penetration seals constructed with materials from the suspect 

lots and the newly manufactured materials. 

program on April 29, 1995. 

No 

Dow Corning completed its test 

3.4 Procurement Document Control and Product Data Sheets 

3.4.1 Procurement Document Control 

On March 23-24, 1995, the team reviewed a sample of Dow Corning customer 

procurement records associated with the six suppliers identified in 

Section 3.2. 

discussed the process with the Dow Corning personnel who are actually 

The team reviewed some recent customer purchase orders (POs) and 

5 

37 
responsible for receiving and processing the nuclear customer orders. 

inspectors observed that none of the customer's POs imposed any nuclear power 

plant safety-related design, qualification, manufacturing, test or other 

unique nuclear requirements upon Dow Corning. The team found that two of the 

six customers contractually imposed the provisions of 10 CFR Part 21 on Dow 

Corning, while the other four customers did not impose 10 CFR Part 21 

requirements on Dow Corning. 

the inspectors ordered Dow Corning products by silicone product name and 

formulation number, such as "Sylgard 170 (fast Cure Silicone Elastomer), 

Part By" and did not specify any nuclear design or specification requirements. 

Some POs required Dow Corning to have an "approved quality assurance [QA] 

program" for the manufacture of its silicone products. 

The 

All of the customer POs which were reviewed by 

The typical Dow Corning nuclear customer PO package also contained a Dow 

Corning "certificate of analysis." A review of several certificates of 

analysis by the inspectors, for both commercial customers and nuclear 

customers noted only one difference between the nuclear and commercial 

certificates for like-products. The certificates that were in the PO package 

for the customers who imposed 10 CFR Part 21 contained a note which stated, 

"Complies With 10 CFR 21;" the others did not. The inspectors informed Dow 

Corning staff that its note certifications that stated compliance with Part 

21 were not applicable, since Dow Corning manufactured and supplied commercial 

grade items for nuclear power plant use. The Dow Corning staff stated that 

they would discontinue issuing the certificates of analysis stating compliance with 10 CFR Part 21. 

3.4.2 Product Data Sheets 

The NRC inspectors reviewed several Dow Corning product data sheets, 

application guide specifications, and s'ilicone foam cell structure and color 

comparison charts. The product data sheets reviewed were for, Sylgard 527 

primerless silicone dielectric gel; 3-6548 si1 icone RTV foam; 96-081 RTV 

adhesive/sealant; and Sylgard 170 A & B silicone elastomer. None of the Dow 

Corning specification sheets indicated any specific end-use applications or 

industry specific use, such as chemical or nuclear. 

3.5 10 CFR Part 21 & Associated Proqrams 

The inspectors reviewed the program that Dow Corning had adopted to implement 

the provisions of 10 CFR Part 21, and evaluated the applicability of 10 CFR 

Part 21 to Dow Corning in regard to the silicone products used at nuclear 

power plants. The inspectors found that Dow Corning had developed procedures 

to address the provisions of 10 CFR Part 21, "Dow Corning Corporation Standard 

071, "Reporting of Deviations and Noncompliance as Required by Federal 

Regulation 10 CFR Part 21," and had also established an associated quality 

system which ensured that the subject deviations were reported to its 

customers. 

not verify compliance with the provisions of 10 CFR Part 21. However, the 

inspectors did verify that the established Dow Corning program met the intent 

of certain elements of the Part 21 regulation, such as 10 CFR § 21.21(b), 

which requires that vendors inform its customers of deviations and failures to 

comply with regulations. 

Since Dow Corning was not subject to Part 21, the inspectors did 

6 

38 
The NRC inspectors a1 so reviewed associated qual i ty systems that typically 

would interface and provide input into a 10 CFR Part 21 program. The team 

concluded that Dow Corning’s International Organization for Standardization 

(ISO) 9002 qual i ty system controls regarding customer complaints appeared to 

ensure that the 10 CFR !j 21.21(b) aspect of informing customers of deviations 

to procurement specifications was effectively executed. For example, the 

subject deviations initially were received from customers who used the 

silicone products in their comnercial electronics applications, and the 

problems were documented and logged into the customer complaint system by the 

Dow Corning customer service personnel. Once the problems were documented and 

logged into the Dow Corning quality system, an investigation of the potential 

product problems were initiated. As a result, Dow Corning found that the 

problem was generic to certain lots, was caused by sulfur in a particular 

ingredient, and customers were informed. 

Therefore, the team concluded that the quality system controls which Dow 

Corning had established would appear adequate to meet the intent of certain 

provisions of 10 CFR Part 21, such as 10 CFR § 21.21(b). 

3.6 Manufacturins Process Control and Product Desiqn Control 

The NRC inspectors conducted discussions, reviewed documents and observed 

manufacturing process controls re1 ated to the Dow Corning si1 icone products 

used at nuclear power plants. 

between the commercial and nuclear products or their associated design or 

manufacturing processes. 

No difference was identified by the inspectors 

3.7 InsDection Team Conclusions 

The team concluded that the silicone products which are procured from Dow 

Corning by its six nuclear customers identified in Section 3.2 are commercia7 

grade itens as defined in 10 CFR 5 21.3. Further, the team noted to Dow 

Corning that a commercia7 grade item can not be transformed into a basic 

component, as defined in 10 CFR 5 21.3, merely by imposing the provisions of 

Part 21 in a PO without imposing other unique nuclear requirements for the 

design, manufacturing, testing and supply of that item. 

Based upon the document reviews, observations by the team, and discussions 

with Dow Corning manufacturing and corporate personnel the inspectors a1 so 

concluded that Dow Corning satisfactorily performed its investigative and 

corrective actions of its silicone product concerns regarding cure time 

parameters of its silicone products that are used in nuclear power plant fire 

barrier applications. Additionally, the portions of Dow Corning’s quality 

system controls reviewed by the inspectors were found to be effectively 

executed. 

7 

39 
4.0 PERSONNEL CONTACTED 

Dow Corninq CorDoration CorDorate Office 

8. 

J. 

8. 

R. 

L. 

R. 

M. 

M. 

L. 

H. 

Woodward 

Boone 

Schroeder 

Love1 and 

Krauss, Esq. 

HcKel 1 ar, Esq . Gill 

Hemi nger 

Bergman 

Laden burger 

Total Quality & Supply 

Total Quality & Supply 

Fire Protection 

Building Materials 

Legal 

Legal 

Customer Service 

Customer Service 

Product Steward 

Product Line 

J. Sharpe 

ti. Chapman 

P. Alexander 

W. Clark 

G. Dehli 

F. Nass 

J. Thomas 

T. Vessels 

C. Johnson 

A. Hauser 

S. St. Pierre 

Dow Corninq Elizabethtown Manufacturina Facility 

P1 ant Manager 

Technical Development 

Qual i ty Engineer 

Techno1 ogy/Qual i ty 

Product Team Leader 

ISO-Audits 

QA Data Center 

QA Laboratory 

Underwriters Laboratory. IncorDorated 

Engineering Services 

Engineering Services 

Laboratory Technician 

8 

40 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 205550001 

May 16, 1995 

Mr. John M. Cooper, Executive Vice President 

Dresser Industries, Inc. 

Industrial Valve Operation 

Dresser Valve & Controls Division 

P.O. Box 1430 

Alexandria, Louisiana 71309-1430 

SUBJECT: NRC INSPECTION NO. 99900054/95-01 

Dear Mr. Cooper: 

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

(NRC) inspection of Dresser Industries, Inc., Industrial Valve Operation 

(Dresser), Alexandria, Louisiana, conducted by Robert Pettis, Richard Mclntyre 

and Billy Rogers, of this office, on March 13-16, 1995. The NRC inspection 

team conducted an evaluation of the Dresser quality program and the 

implementation of that program as it relates to the manufacture of safetyrelief valves and spare parts supplied to the nuclear industry. 

The NRC inspection team reviewed documentation, procedures, and representative 

records, conducted interviews and held discussions with members of Dresser's 

staff. On the basis of this inspection, the inspection team determined that 

the implementation of the Dresser quality assurance program complies with the 

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

Resulations. 

the areas examined. No response to this letter or its enclosure is required. 

The enclosed inspection report contains a detailed discussion of 

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

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

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

discuss them with you. Thank you for your cooperation. 

Sincerely, n 

Robert M. Gallo, Chief 

Speci a1 Inspection Branch 

Division of Technical Support 

Office of Nuclear Reactor Regulation 

Docket No.: 99900054 

Enclosure: Report No. 99900054/95-01 

41 
REPORT NO.: 

ORGANIZATION: 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY : 

INSPECTION DATES: 

LEAD INSPECTOR: 

OTHER INSPECTORS: 

REVIEWED BY: 

APPROVED BY: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF TECHNICAL SUPPORT 

99900054/95-01 

Dresser Industries, Inc. 

Industri a1 Valve Operation 

Dresser Valve and Controls Division 

P.O. Box 1430 

A1 exandri a, Loui si ana 71309-1430 

John W. Richardson 

Manager, Quality Assurance 

Manufacturer of various model safety and pressure 

relief valves supplied to the nuclear industry. 

March 13-16, 1995 

todrt L. Pettis, Jr., P.E. 

Vendor Inspection Section (VIS) 

Special Inspection Branch (SIB) 

Division of Technical Support (DOTS) 

Office of Nuclear Reactor Regulation (NRR) 

Richard P. McIntyre, VIS/SIB/DOTS/NRR 

Billy H. Rogers, VIS/SIB/DOTS/NRR 

VIS/SIB/DOTS/NRR 

Date 

5- /I6 I '1 4- 

Robert M. Galdo, Chief 

S IB/DOTS/NRR 

Date 

Enclosure 
1 SUMMARY OF INSPECTION FINDINGS 

During this inspection, the NRC inspection team evaluated the implementation 

of the Dresser Industries, Incorporated, Industrial Valve Operation, (Dresser) 

qual i ty assurance program re1 ated to the manufacture of various model safetyrelief valves supplied to the nuclear industry. The inspection was conducted 

to determine Dresser’s compliance with the requirements of Appendix B to Part 

50 of Title 10 of the Code of Federal Requlations (Appendix B) and the 

provisions of 10 CFR Part 21. The inspection team reviewed technical 

information, procedures and representative records, conducted interviews, held 

discussions and observed manufacturing activities. No violations or 

nonconformances were identified during this inspection. 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

None were reviewed dur 

3 INSPECTION FINDINGS 

3.1 Backqround 

ng this inspection. 

AND OTHER COMMENTS 

Dresser Industries, Inc., has served the power industry since 1880. Today, 

Dresser Industries has more than 100 manufacturing plants and sales, service 

and other facilities in some 50 countries around the world. The nuclear 

industry product line includes various safety and pressure relief valves 

manufactured to the overpressure protection requirements of Sections I and 

VI11 of the American Society of Mechanical Engineers (ASME) Boiler and 

Pressure Vessel Code, and to an ASME Section I11 approved Quality Assurance 

Program which also meets the requirements of Appendix B. 

continues to rely on these types of valves to provide overpressure protection 

suitable for the high pressure and temperature applications required by modern 

power generation equipment. 

accordance with ASME Code rules to establish rated capacities and have been 

certified by the National Board of Boiler and Pressure Vessel Inspectors. 

Dresser Industries’ world headquarters is located in Dallas, Texas. 

The power industry 

All valves are flow tested by Dresser in 

3.2 Entrance and Exit Meetinqs 

During the entrance meeting, held on March 13, 1995, the NRC inspection team 

met with members of Dresser management and staff, discussed the scope of the 

inspection, and established organizational contacts. During the exit meeting, 

held on March 16, 1995, the inspection team summarized its findings with 

Dresser management and staff. The Appendix of this report lists the persons 

contacted during the inspection. 

3.3 Review of 10 CFR Part 21 

The inspection team reviewed Dresser Quality Assurance Procedure (QAOP) 1-9, 

“Evaluating and Reporting of Deviation and/or Noncompliance Affecting Safety 

Related to NRC Regulation 10 CFR Part 21,” Revision 2, March 1, 1995, which 

was well written and appeared to be properly implemented. In addition, the 

inspection team observed Dresser’s 10 CFR Part 21 posting as required by 

10 CFR 21.6. 

of 1974 (Section 206) and a notice which described 10 CFR Part 21 and QAOP 1-9 

and to whom reports could be made. 

Dresser had posted Section 206 of the Energy Reorganization Act 

The inspection team noted a weakness in 

2 

43 
the posting regarding the location where QAOP 1-9 and 10 CFR Part 21 could be 

examined and other minor discrepancies between the notice and QAOP 1-9. 

Dresser indicated that the notice would be revised to be in agreement with the 

current revision of QAOP 1-9 and to indicate where 10 CFR Part 21 and QAOP 1-9 

could be examined. 

The inspection team reviewed Dresser's records associated with 10 CFR Part 21 

evaluations performed from 1990 to 1995. 

performed 13 evaluations and made one report to the NRC. 

appeared to have been performed in a thorough and timely manner with one 

exception. The inspectors noted that 10 CFR Part 21 file 94-02 indicated an 

evaluation which lasted from May 19, 1994, until July 28, 1994, in excess of 

the 60 days allowed by Part 21. Dresser indicated that the basis for the 

listed evaluation completion date was the issue date of Dresser Report SRV237A, "Review of Pub1 ished Adjusting Ring Positions Given In Service Manuals 

for 1900 Safety Relief Valves Nuclear Plants," Revision 0, dated July 30, 

1994, which discussed aspects of the 10 CFR Part 21 evaluation but also 

supported additional work. Based on review of SRV-237A and associated 

supporting documentation and discussions with Dresser, the inspection team 

concluded that the actual evaluation completion date (during which it was 

determined that a defect did not exist) was on or about May 19, 1994, and that 

no violation of Part 21 had occurred. Dresser also indicated that the file 

would be corrected to indicate the actual completion date. 

team discussed the importance of accurately documenting 10 CFR Part 21 

evaluations and reporting activities and concluded that this was a weakness in 

the implementation of Dresser's 10 CFR Part 21 evaluation program. 

During this period, Dresser had 

The evaluations 

The inspection 

3.4 Dedication Activities 

The inspection team reviewed Dresser Procedure EG37, "Quality Classification 

of Nuclear Pressure Relief Valve Parts," Revision 12, August 18, 1994. The 

procedure is used to establish the quality classification for valve parts 

supplied to licensees and provides instruction on the use of commercial grade 

items (CGIs) when installed in valves classified as safety-related. 

Appendix A of the procedure, "Commercial Grade and Non-Safety-Related Items 

Quality Class C and D Parts," provides information for establishing the 

suitability of using CGIs in a safety-related valve. The procedure also 

establishes definitions for acceptance methods, basic component, CGI, critical 

characteristics, dedication, safety-related and safety function. It also 

provides specific examples of how the definitions are applied to Dresser 

valves and the dedication of CGIs used in a safety-related application. 

dedication process includes performance of a technical evaluation and the 

determination of the appropriate acceptance method(s). 

The 

The technical evaluation assured that the specification requirement for the 

CGI were adequately trans1 ated into the procurement document. 

methods were documented in the Nuclear Standard Master Control Document, 

assigned to the applicable utility, and the applicable assembly instruction. 

The acceptance methods included material verification, documentation of 

traceability and dimensional inspection. The inspection team concluded that 

the procedure effectively imp1 emented the Dresser requirement of performing 

100 percent inspection and testing of all parts used in safety-related valves. 

The acceptance 

3 

44 
3.5 Internal Audits and Corrective Actions 

The inspection team reviewed Section 16, "Corrective Action," and Section 18, 

"Audits," of the Dresser "Quality Assurance Manual (QAM) For Section I11 

Nuclear Valves," Revision 23, June 5, 1992; Quality Assurance Operating 

Procedures (QAOP) 1-6, "Internal Audits, I' and QAOP 1-7, "Corrective Action," 

both Revision 17, March 1, 1995. 

assurance (QA) program would be audited at least annually by a panel of at 

least four representatives from various departments. A1 so, monthly audits 

were scheduled in one or more of the sections of the QA Program. These 

monthly audits were conducted using checklists for each section of the QA 

program and cover the following four product areas: 

The requirements stipulated that the quality 

ASME Section 111 Nuclear Valves 

NR - Nuclear Repair/ ASME Section XI 

ASME Section I, IVY VI11 - Non Nuclear 

VR - Non Nuclear Repair 

The inspection team reviewed the reports for the "Annual Review of the Quality 

System Status and Adequacy of QA Programs Valves and Parts" performed for 1990 

through 1993 (the report for 1994 had not yet been completed). included a list of closed and outstanding corrective actions for audit 

findings identified during the monthly audits. The inspection team noted that 

several audit findings had been open for several years. Dresser stated that 

some issues were being reviewed by special task groups and others were left 

open to reverify implementation of corrective actions over a period of time. 

These reports 

The inspectors selected a sample issue to review from the list of corrective 

actions. Corrective Action Request (CAR) QA-92-21, June 6, 1992, documented a 

finding concerning safety-re1 ief valves being assembled without the use of 

torque wrenches for the bonnet bolting as required by Assembly Instruction AS132. 

than the original bonnet, which did not provide adequate clearance for a 

torque wrench without additional actions such as machining (the original 

bonnet had provided adequate clearance). 

was used only on commercial valves and had not been used on any nuclear valves 

and AS-132 was not appl icable to nuclear valves. 

The valves, which were sold as nuclear safety-related, were manufactured and 

assembled separately from the commercial product line using different assembly 

and test instructions and procedures. 

The issue concerned the use of a new style bonnet, shaped differently 

Dresser stated that the new bonnet 

The inspection team reviewed EG-34, Revision 7, March 8, 1995, "Nuclear 

1ns.tallations Pressure Re1 ief Valves," which is the Engineering Instruction 

(EI) that listed all of Dresser's nuclear customers and the valves sold to 

them by model number along with the accompanying Master Control Document 

(MCD). The inspection team reviewed a sample of 16 valves representing 11 

different nuclear utilities. 

identified the instructions and procedures to be used during manufacturing 

that were specific to the nuclear product line. 

The review of the MCD verified that it properly 

4 

45 
The inspection team also reviewed EG-2, Revision 8, March 6, 1995, "Nuclear 

Application Manual ,I' which lists all the instructions and procedures that are 

applicable for the nuclear product line. These included, but were not limited 

to: 

e 

e 

e 

e 

e 

e 

e 

a 

e 

e 

e 

e 

Assembly Instructions 

Cleaning Instructions 

Funct i onal Test i ng Instruct i ons 

Hardsurfacing instructions 

Heat Treatment Instructions 

Hydrostatic Testing Instructions 

Mechanical Testing Procedure 

Nondestructive Testing Procedure 

Purchasing Specification 

Repair Instructions 

Welding Instructions 

Master Control Documents (Nuclear Standard) 

The inspection team did not identify any instances where nuclear valves were 

assembled or tested using a commercial product line instruction such as 

AS-132. 

reference to an E1 or procedure contained in EG-2. 

that valves manufactured on the nuclear product 1 ine were manufactured using 

nuclear specific instructions and procedures, that the applicable instructions 

and procedures required bonnet bolting, and that adequate documentation 

existed to support the required torque. 

All MCDs reviewed by the inspection team contained the appropriate 

The inspectors verified 

3.6 Analysis of Gasketed Joints 

The inspection team reviewed the engineering basis which supports gasket 

integrity and bolt stress calculations for the joint between the valve body 

and the bonnet. To accomplish this, Dresser developed a computer program 

called GENFLG which analyzes the joint in accordance with the requirements of 

ASME Section I11 and customer specifications. The program calculates the 

resulting bolt stress from input loadings obtained from each customer's piping 

stress analysis which is supplied to Dresser. 

The program has been used at Dresser since 1972 and uses valve input data from 

various tables in the American National Standards Institute (ANSI) ANSI B16.5, 

"Pipe Flanges and Flanged Fittings," ANSI 616.34, "Valves-Flanged, Threaded, 

and Welded End," and ASME Section 111. For example, values for various 

gaskets, identified as M and Y values in ANSI B16.5, are used to approximate the compressive strength of the gasket based on its material. 

The program performs a structural analysis of the joint in accordance with 

Appendix I1 of Section VI11 of the ASME Code. External loading in the form of 

forces and moments are input into the program which results in the calculation 

of bolt forces and stresses. Natural frequency calculations are performed 

5 

46 
using the Dresser computer program VIB, which calculates the first mode 

natural frequency of the SRV. 

The inspection team reviewed joint calculations for customer purchase order 

(PO) No. 18-62951-0, Arizona Public Service Company (APSC), Palo Verde Nuclear 

Plant, Units 1,2 and 3, August 30, 1991. The order was for eight, ASME 

Section 111, Class 2, 6-inch, 150-pound ANSI rated SRVs. The PO invoked the 

provisions of 10 CFR Part 21 and APSC Design Specification No. 13-JN-6918, 

Revision 1. 

identification, materials and valve application data were provided in Dresser 

Contract Document 3NC-3112. The file and its supporting documentation 

appeared complete and in accordance with Dresser QA procedures. 

issued APSC a Certificate of Compliance on November 22, 1991. 

The general design features of the SRV, its component 

Dresser 

3.7 Quality ImDrovement Team Activities 

Dresser established a Quality Improvement Team (QIT) to review processes or 

situations where improvements could be made and to make recommendations on 

methods of achieving those improvements. The inspection team reviewed the 

conclusions of one such QIT, the Safety Relief Valve (SRV) team, and discussed 

with its members the origin of the SRV team, its composition, the methods 

used, the conclusions drawn, and implementation of improvements based on the 

conclusions. The SRV team had been assembled by Dresser management in 1991 to 

review the circumstances of a high rejection rate of commercial SRVs during 

final production testing. The rejection rate applied to commercial SRVs being 

tested as a part of confirmation of the production process prior to shipment of the SRVs to customers. If the commercial SRV did not pass the production 

test it was reworked on the production line and retested as required prior to 

shipment. 

The SRV team's conclusions and recommendations, documented in a February 1993 

report, were accepted by management and incorporated into the commerci a1 SRV 

production activities. 

commercial SRVs had been significantly reduced and that a constant review of 

commercial SRV production resulted in the low failure rate. 

Dresser indicated that the rejection rate of 

The inspection team verified that the activities of the SRV review team only 

applied' to the commercial valve product 1 ine and that a separate set of 

production methods and procedures are used for nuclear SRVs. Commercial 

are typically manufactured one hundred at a time and inspected on a samp 

basis, while nuclear SRVs are manufactured one or two at a time and 

individually inspected. The primary distinction between the nuclear and 

commercial manufacturing operations is that additional time is allotted 

performing the same operation for the nuclear line. 

For example, the machining of an interior angle in the valve body would 

typically take 15 minutes for a commercial SRV using numerical control 

machining, contrasted with 60 minutes for a nuclear SRV using manual 

machinins. The insDection team reviewed the results of the SRV team, 

SRVs 

e 

the 

or 

discussea the issue' with SRV team members and Dresser management, reviewed the 

applicable commercial and nuclear procedures, and concluded that the earlier 

SRV rejection rate did not apply to nuclear SRVs and that there were no 

concerns identified with nuclear SRVs. 

6 

47 
APPENDIX 

PERSONS CONTACTED 

The following persons were contacted during the inspection and except as 

noted, attended both the entrance and exit meetings. 

Dresser: 

J. Cooper 

R. Danzy 

T. Bordelon 

J . Richardson 

A. Casler 

J. Watz 

R. Huffman 

D. Sharma 

R. Warrington 

R. Thomas 

R. Maxey 

Executive Vice President and General Manager 

Vice President, Engineering 

Vice President, Sales and Marketing 

Manager, Quality Assurance 

Manager, Quality Processes 

Principal Qual i ty Assurance Engineer 

Senior Product Engineer 

Supervisor, Application Engineering 

Industrial Engineer 

Qual i ty Assurance Engineer 

Production and Inventory Control 

U. S. Nucl ear Regul atory Commi ss i on : 

* 

R. Pettis 

G. Cwalina 

R. McIntyre 

B. Rogers 

Team Leader 

Chief, Vendor Inspection Section 

Senior Reactor Engineer 

Reactor Engineer 

Attended exit meeting only 

- A-1 - 

48 
' +**** 

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-oOo1 

June 14, 1995 

Mr. Mark T. Capallo, President 

Energy & Process Corporation 

2146-B F1 intstone Drive 

Tucker, GA 30084-5000 

SUBJECT: NRC INSPECTION NO. 99900866/95-01 

Dear Mr. Capallo: 

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

of your facility at Tucker, Georgia, conducted by Messrs. U. Potapovs and 

L.L. Campbell of this office on May 8 through 11, 1995, and the discussions of 

their findings with you and members of your staff at the conclusion of the 

inspection. 

program and its implementation in selected areas such as (1) control of 

purchased materi a1 and servi ces, (2) materi a1 traceabi 1 i ty control , (3) upgrading of material purchased from non-qualified sources (commercial 

grade item dedication), and (4) a review of your program for implementing 

Part 21, "Reporting Defects and Noncompl ance," of Title 10 of the Code of 

Federal Resul at ions (10 CFR) . 

The inspection was conducted to evaluate your quality assurance 

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

enclosed inspection report. 

procedures and representative records, discussion, and observations by the 

inspectors. 

This inspec ion consisted of an examination of 

Our review of your activities in these areas indicated that, although your 

qual i ty assurance program and its implementation adequately addressed material 

supplied under the rules of the ASME Code, it failed to meet certain NRC 

requirements applicable to your commercial grade item dedication practices 

utilized when supplying material under the quality requirements of 10 CFR 

Part 50, Appendix B. Specifically, your quality assurance program failed to 

provide adequate instructions or procedures for control1 ing the commercial 

grade item dedication process and also failed to provide a documented basis to 

substantiate that your sampling plans for verifying critical characteristics 

provided reasonable assurance that the dedicated items met the applicable 

procurement document requirements. 

methods of manufacturing tensile test specimens used in the dedication 

process. 

are identified in the enclosures to this letter. 

Deficiencies were a1 so identified in your 

The specific findings and references to the pertinent requirements 

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

statement in accordance with the instructions specified in the enclosed Notice 

of Nonconformance. 

show good cause for us to do so. 

We will consider extending the response time if you can 

49 
Mr. M. Capallo -2- 

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

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

by the Paperwork Reduction Act of 1980, Pub1 ic 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 

n 

y , 

-- 

'Robert M. Gallo, Chief 

Special Inspection Branch 

Division of Technical Support 

Off i ce of Nucl ear Reactor Regul at i on 

Docket No.: 99900866 

Encl osures : 1. Notice of Nonconformance 

2. Inspection Report 99900866/95-01 

50 
Energy & Process Corporation 

Tucker , Georg i a 

"xket No.: 990 56/95-01 

Based on the results of an NRC inspection conducted on May 8 through 11, 1995, 

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

w i t h NRC requ i remen t s . 

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

Title 10 of the Code of Federal Requlations (10 CFR) Part 50 requires, 

in part, that activities affecting quality shall be accomplished in 

accordance with instructions, procedures, or drawings. 

Contrary to the above, Energy & Process Corporation (EP) had not 

developed instructions or procedures that provided requirements for 

controlling commercial grade item (CGI) dedication activities such as 

(a) selecting critical characteristics to be verified, (b) identifying 

proper verification methods for confirming the acceptability of critical 

characteristics, and (c) ensuring heat or lot traceability to 

manufacturer's certification. (Nonconformance 99900866/95-01-01). 

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

of Appendix B to 10 CFR Part 50 requires, in part, that measures shall 

be established to assure that purchased material conforms to procurement 

documents. 

Contrary to the above, EP had not established a documented basis to 

substantiate that its destructive testing and dimensional inspection 

sampling plans for verifying critical characteristics provides 

reasonable assurance that dedicated CGIs supplied met applicable 

procurement documentation requirements (Nonconformance 99900866/ 

95-01-02). 

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

10 CFR Part 50 requires, in part, that activities affecting quality 

shal J be accompl i shed in accordance with instructions , procedures, or 

drawings . 

Section 26.0.2 of EP Procedure QCP 26, "Tensile Testing Procedure" 

states that tensile testing shall be performed in accordance with 

Specification A/SA 370, "Specification for Test Methods and Definitions 

for Mechanical Testing of Steel Products". 

Section 6.3, "Procurement of Test Specimens," of ASTM Specification 

A 370, requires, in part, that when test specimens are manufactured, 

care shall be taken to remove by machining all cold worked areas from 

the section used in evaluating the test. 

Contrary to the above, the method used to manufacture tensile specimen 

from a stainless steel elbow supplied under Virginia Power Co. purchase 

Enclosure 1 

51 
order CNT 479183 produced a test section containing cold worked areas 

which may have contributed to unusually high yield strength recorded for 

this elbow. (Nonconformance 99900866/95-01-03). 

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

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

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

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

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

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

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

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

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

and preventive measures were or will be completed. 

Dated at Rockvi 11 e, Mary1 and 

this 14th day of June, 1995 

52 
REPORT NO.: 

ORGAN I ZAT ION : 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY: 

INSPECTION DATES: 

LEAD INSPECTOR: 

OTHER INSPECTORS: 

REVIEWED BY: 

APPROVED BY: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF TECHNICAL SUPPORT 

99900866/95-01 

Energy 81 Process Corporation 

2146-B F1 intstone Drive 

Tucker, GA 30084 

Charles. E. Thornton, Manager 

Quality Assurance 

Energy & Process Corporation is a supplier of metal 

products to the nuclear industry. 

May 8 through 11, 1995 

U1 k+-? di s Potapovs 

oa;-8-9S 

Vendor Inspection Section (VIS) 

Special Inspection Branch (TSIB) 

Larry L. Campbell 

Quality Assurance and Safety Assessment Section 

Qual i ty Assurance and Maintenance Branch 

Speci a1 Inspection Branch 

Robert M. Gallo, Chief, TSIB 

Enclosure 2 

53 
1 SUMARY OF INSPECTION FINDINGS 

During this inspection, the NRC inspectors evaluated Energy & Process's (EP's) 

implementation 

directly affected the quality and performance capability of their product. 

These activities included the control of purchased material and services, 

material traceability control, and commercial grade item (CGI) dedication. 

The team also reviewed EP's program for implementing Part 21, "Reporting 

Defects and Noncompliance," of Title 10 of the Code of Federal Requlations (10 

CFR) . 

of quality assurance measures for those activities which 

The inspection basis consisted of the following: 

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

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

Federal Requlations (10 CFR Part 50) 

10 CFR Part 21 

Section I11 of the ASME Boiler and Pressure Vessel Code (Code) 

EP's Quality Systems Manual, Revision 0, dated November 24, 1993 

1.1 Violations 

No violations were identified during this inspection. 

1.2 Nonconformances 

1.2.1 Nonconformance 95-01-01 

This nonconformance, described in Sections 3.4.2.1 and he report 

identifies EP's failure to develop appropriate instructions or procedures that 

provide requirements for controlling commercial grade dedication activities 

such as selecting critical characteristics to be verified, identifying 

acceptable verification methods, or assuring material traceability. 

.2.4 of 

1.2.2 Nonconformance 95-01-02 

This nonconformance, described in Section 3.4.2.4 of the report identifies 

EP's failure to establish a documented basis for substantiating that its 

material sampling plans for verifying critical characteristics provide 

reasonable assurance that the dedicated CGIs supplied to their customers meet 

applicable procurement specification requirements. 

1.2.3 Nonconformance 95-01-03 

This nonconformance, described in Section 3.4.2.3.3 of the report identifies 

an instance where the method used to manufacture a tensile test specimen 

failed to meet the applicable specification requirements. 
2 STATUS OF PREVIOUS INSPECTION FINDINGS 

This was the first NRC inspection of Energy & Process Corporation 

3 INSPECTION FINDINGS AND OBSERVATIONS 

3.1 Entrance and Exit Meetinqs 

During the entrance meeting on May 8, 1995, the NRC inspectors discussed the 

inspection scope and developed general information about EP’s products and 

activities. During the exit meeting on May 11, 1995, the NRC inspectors 

discussed their findings and observations with EP’s management. 

3.2 Description of Facilities 

EP has been accredited by the American Society of Mechanical Engineers (ASME) 

as a material supplier of ferrous and nonferrous bars, threaded fasteners, 

castings, forgings, plate, seamless fittings, flanges, NPT stamped tubular 

products, structural shapes, welding material, standard supports, and similar 

items. The scope of their Quality Systems Certificate (QSC) also includes the 

qualification of material manufacturers and suppliers of subcontracted 

services as well as upgrading of stock material. EP has also been accredited 

as a materi a1 manufacturer of ferrous and nonferrous threaded fasteners, 

seamless fittings and flanges. Their QSC’s expire on April 28, 1996. 

According to EP management, approximately 85% of their products are supplied 

for nuclear appl i cat i ons. 

supplied under their ASME quality systems program and certified under their 

QSCs, with the remaining 35% supplied under 10 CFR Part 50, Appendix B quality 

assurance requirements. 

commercial grade items and dedicated. 

Approximately 65% of their nuclear orders are 

Most of the Appendix B material is purchased as 

EP does not perform any material manufacturing operations at their facility. 

EP has the capability to perform limited threading operations and material 

subdividing. 

performing chemical analyses (optical emission spectrometer), as well as 

tensile and hardness testing. 

A material test laboratory is located on site and is capable of 

3.3 10 CFR Part 21 Prosram 

3.2.1 Imp1 ement i ng Procedures. 

The NRC inspectors determined that EP maintained the required 10 CFR Part 21 

postings in a conspicuous place. 

and Noncompliance,” dated January 7, 1994, to implement the provisions of 

10 CFR Part 21. 

requirements of Section 21.21, ”Notification of failure to comply or existence 

of a defect and its evaluation,” of 10 CFR Part 21 and identified several 

weaknesses. Specifically: 

EP prepared QCP 16, “Reporting of Defects 

The NRC inspectors reviewed QCP 16 for compliance with the 

3 

55 
QCP 16, for the most part, only referenced the use of 10 CFR Part 21 and 

did not contain the prescriptive text required by Section 21.21 of 

10 CFR Part 21. 

QCP 16 stated, in part, that "Should the evaluation conclude that the 

condition is reportable under 10 CFR Part 21, the QA Manager shall 

notify all affected customers after tracking the shipments of the 

applicable material." The NRC identified to EP that if it concludes 

that the condition is reportable, in addition to notifying its 

customers, it must also notify the NRC in accordance with the 

requirements of Section 21.21 of 10 CFR Part 21. 

EP informed the NRC inspectors that its 10 CFR Part 21 implementing procedures 

would be would be revised to address the weaknesses identified in this 

inspection report. 

3.2.2 EP Procurement Documents 

The NRC inspectors examined the EP Approved Vendor List for companies that 

were approved Appendix B material suppliers and that provided calibration and 

nondestructive evaluation services. The NRC inspectors reviewed documentation 

for material and services provided by three companies and found that EP had 

imposed the requirements of 10 CFR Part 21 and 10 CFR Part 50, Appendix By and 

that the companies had acknowledged acceptance of these requirements. 

3.3.2 Notification Concerning Deviating material 

On January 25, 1994, PECO Energy Company (PECO) issued a corrective action 

request (CAR) to EP concerning 150 ASME SA 194, Grade 7, heavy hex nuts which 

were supplied by EP in May 1993. 

analyses for 143 of the 150 nuts did not agree with the values shown on the 

certified material test report (CMTR) suggesting a loss of heat traceability. 

The CAR also noted that the manufacturer's and grade markings of the 143 nuts 

appeared different (stamped) than those of the remaining 7 nuts (raised). 

The CAR was issued after PECO's chemical 

The nuts were originally purchased from A&G Engineering Co. 11, Inc. (A&G) by 

Hub Inc. (now EP) in 1986. At the time of the procurement, A&G was an ASME 

QSC holder and the material was procured under their ASME Certificate. 

However, A&G declared bankruptcy in 1993 and is no longer in business. In 

February 1995, the NRC issued Information Notice 95-12, to alert licensees 

that A&G may have supplied nonconforming fasteners. 

In response to PECO's CAR, EP performed chemical analyses on several nuts from 

the same heat of material and determined that the carbon contents of these 

nuts was below the specification requirements. 

the material supplier was no longer in business, the root cause of the 

deviation could not be determined and issued a return authorization for these 

nuts. EP also notified other recipients of nuts from the same heat of 

material that their testing of these nuts had identified carbon contents below 

the minimum allowed for SA 194 Grade 7 material. 

EP advised PECO that, since 

Subsequently, EP issued a 

4 

56 
follow up to the customer notifications suggesting that the low carbon values 

may be attributed to a decarburization zone caused by heat treatment and that 

retest results with the material surface layer removed showed carbon contents 

within specification. 

The inspectors determined that EP’s disposition of the reported deviation was 

consistent with the requirements of 10 CFR Part 21. 

3.4 Oualitv Assurance Proqram ImDlementation 

3.4.1 Material Supplied to ASME Code Requirements 

3.4.1.1 Program Control s 

EP’s program for supplying ASME Code material is described in their Quality 

Systems Manual(QSM), Revision 0, dated November 24, 1993 which is committed to 

meeting the requirements of ASME Section 111, NCA 3800 as well as the 

applicable portions of ANSI N45.2, IS0 9002, 10 CFR 50, Appendix B, and NQA-1 

Basic Requirements. The program controls are described in the manual and in 

referenced implementing procedures. Selected areas of the manual and its 

implementing procedures were examined during this inspection. 

EP’s QSM and its implementing procedures require that purchase orders for ASME 

Code material be placed only with vendors appearing on EP’s Approved Vendor 

List (AVL). Vendors are placed on the AVL based on holding a valid ASME QSC 

or following the completion of a satisfactory survey by EP in accordance with 

the requirements of Subarticle NCA 3800 of the ASME Code, Section 111. 

Provisions for upgrading of stock (Commercial grade) material were reviewed 

and determined to be in compliance with the applicable ASME Code requirements. 

These procedures require the performance of all tests specified in the 

applicable material specification and in the Code on each piece of the stock 

material or, when material traceability is established by audit, the 

performance of the required tests on one sample of each heat of material. 

Chemical analysis is required on each piece of the stock material in either 

case. 

The NRC inspectors reviewed several recent customer purchase orders and 

accompanying data packages and determined that in most instances the material 

appeared to have been supplied in accordance with the applicable QSM 

provisions. Sufficient documentation was generally available to demonstrate 

compliance with customer purchase order and ASME Code requirements. 

document review identified two areas where the bases for acceptability of 

material from EP suppliers did not appear to be fully supported. 

relates to inadequate verification of compliance with specific EP purchase 

order requirements during material receiving inspection as shown in the 

following example: 

The 

One of these 

On February 23, 1995, TVA issued a purchase order to EP for 1/2 inch x 

3/8 inch 3000 lb couplings to be supplied in accordance with ASME 182, 

TP316L. EP ordered the material from Alloy Stainless who is a qualified 

5 

57 
supplier and specified in their purchase order that mill test reports 

and certificate of compliance must be provided with the material. 

Instead of providing mill test reports, Alloy Stainless supplied (and EP 

accepted) this material with "Test Report Certification" on Alloy 

Stain1 ess 1 etterhead. 

While the acceptance of this material does not violate the currently 

applicable ASME Code- or specification requirements, it suggests confusion over 

the definition of terms or inadequate guidance for material acceptance. 

Several other packages contained similar inconsistencies. 

The second area involved the acceptance of manufacturer's letter as apparent 

basis for not performing hydrostatic test on pipe which was purchased as stock 

material as summarized below: 

On December 2, 1994, Florida Power & Light Co. issued a purchase order 

to EP which included 96 feet of 4 inch diameter schedule 40, SA 106, 

Grade B, seamless pipe to the 1989 edition of ASME Code. EP ordered the 

pipe from MarmonlKeystone who, in turn, obtained it from US Steel. 

Upgrading by EP i ncl uded performance of chemical analyses, tensi 1 e, and 

flattening tests. 

Instead of performing a hydrostatic test to verify compliance with the 

material specification, EP accepted a statement on US Steel Co. 

letterhead noting that they hydrotest all seamless standard and line 

pipe to the requirements of the applicable material specification. 

The inspectors noted that basing the acceptance of stock material on 

certification by the supplier in lieu of test may not fully comply with the 

material upgrading requirements of the ASME Code. 

identified to EP as weaknesses in the implementation of their ASME QSM. 

The above examples were 

3.4.2 Non Code material, Supplied to 10 CFR 50, Appendix B Requirements 

(Commercial Grade Dedication) 

3.4.2.1 Program and Procedures 

The documented requirements for EP's dedication process were limited to the 

following text in Section 6.7 of the EP Quality Systems Manual. 

Stock (commercial grade) material may be used for safety-related, 

non-ASME Section I I I purchase orders provided selected tests are 

made to verify compliance with the applicable material 

specification. Such tests are conducted per heat or lot. Results 

of the testing shall be passed on to the customer. 

The NRC inspectors determined that EP had developed procedures that controlled 

activities supporting the dedication process including (1) processing incoming 

orders, (2) receiving inspection, (3) laboratory testing, (4) audits and 

6 

58 
surveys, and (5) qual i ty control i nspector training and certi f i cat i on. 

However, EP informed the inspectors that it had not developed procedures for 

control1 ing commercial grade item (GI) dedication activities such as 

selecting and verifying critical characteristics.The implementation of EP's 

dedication activities was reviewed and is discussed in Sections 3.4.2.3 and 

3.4.2.4 of this report. 

The NRC inspectors determined that incoming customer purchase orders (POs), 

were initially reviewed by the EP's sales department and the need for 

purchasing material testing services was identified and placed in a hold 

status until the material was receipt inspected. 

need for testing to support the dedication of CGIs were being documented on 

Form 105, "QC Inspection Plan," and Form 113, "Material Cutting Ticket." 

Receiving inspection and the 

7, 1994 

report. 

EP's fa 

for the 

aPProPr 

charact 

Additionally, the NRC inspectors determined that EP screens all incoming CGIs 

by reviewing the material certifications provided by the nonqualified supplier 

for compliance to the applicable material specification and by verifying 

markings or tagging provided with the CGIs . that the selection and verification of critical characteristics, at times, 

appeared to be based on performing the tests that could be performed on a CGI 

rather than those tests that should be performed (see section 3.4.2.3.2 of 

this report). The test methods, at times, also appeared to be based on EP's 

judgement rather than accepted test methods (see section 3.4.2.3.3 of this 

report). EP informed the NRC inspectors that it had issued an E-Mail, dated 

January 11, 1995, increasing the sampling size of the chemical analyses 

required for CGI dedication. This E-Mail was issued to employees involved in 

CGI dedication activities and identified that the number of chemistry analyses 

to be performed would be in accordance with the sampling plan presently being 

used for performing dimensional inspection , Attachment 20.1, "Lot Sampling 

Plan," of QCP 20, "Receiving Inspection," dated January 7, 1994 and Attachment 

5.1 "Lot Sampl ing Plan" in QCP 5, "Qual i ty Control Inspection" dated January 

EP's use of sampling is further discussed in Section 3.4.2.4 of this 

The NRC inspectors determined 

lure to establish a documented CGI dedication program or procedures 

selection of critical characteristics and identification of 

ate verification methods to confirm the acceptability of critical 

ristics was identified as a nonconformance with Criterion V of 

10 CFR 50, Appendix B (Nonconformance 99900866/95-01-01). 

3.4.2.2 Dedication Process Strengths and Weaknesses 

The NRC inspectors determined that even though EP's CGI dedication program did 

not have procedures or instructions for the identification and verification of 

critical characteristics, the dedication practices reviewed included a number ' of the essential elements of CGI dedication. EP's practice of attempting to 

obtain the original mill test reports of commercial grade material was 

identified as a strength in their CGI dedication process. 

and staffed material testing laboratory was also identified as a significant 

positive factor. 

EP's well equipped 

7 

59 
However, in addition to the nonconformances 

report, the inspectors also observed severa 

are summarized bel ow. 

0 The technical bases for not verifying 

requirements such as hydrostatic test 

certain material mechanical DroDertie 

identified in Section 1.2 of this 

program/process weaknesses which 

certain material specification 

ng of dedicated CGI pipe and 

I. of CGIs that are too small for 

the manufacture of a test specimen were not required to be documented. 

Laboratory test results were accepted as meeting the material 

specification requirements without questioning what effects these 

results have on other material specification requirements not verified 

during the dedication process. This is of concern when the other 

material requirements, not verified, are considered critical 

characteristics or when there are questionable differences between the 

EP test results and the test results provided by the material supplier. 

The NRC inspectors questioned EP’s practice of including unvalidated 

supplier material certifications (stamped ”QA OK” during the initial 

screening of incoming commercial grade items) in documentation packages 

supplied to customers when, at times EP’s confirmatory material property 

overchecks clearly show that the supplier’s or EP’s certification was 

nitactinnihla Ialcn COD Cartinn 2 A 7 A nf thic rnnnrtl 

3.4.2.3 Review of Dedication Packages 

The NRC inspectors reviewed the following completed CGI dedication packages to 

determine if critical characteristics for the items had been properly 

identified and verified, and if adequate procedural controls were in place. 

3.4.2.3.1 

No. CNT 486707,dated December 12, 1994 with a Revision, dated January 10, 

1995, Item No. 3, was issued to EP for the supply of 9 each, 1 1/2 inches, 

chrome-moly, American Society for Testing and Materials (ASTM) A 182, Grade 

F22, 90 degrees, pipe elbows. EP issued PO No. E256-00823, to a nonqualified 

suppl i er, Pennsyl vani a Machi ne Works (PMW) , and purchased 10 commerci a1 grade 

pipe elbows for this Virginia Power PO. EP verified that markings on the tees 

were correct, reviewed the nonqual ified manufacturer’s material test report 

for conformance with the material specification requirements, performed a 

hardness check, chemistry analyses, and tensile test (including elongation and 

reduction of area), on 1 of the tees, and dimensional inspections on 4 of the 

tees. The results of the overchecks performed by EP indicated that the 

attributes tested were in compliance with ASTM A 182, Grade F22, Class 1, 

specification requirements. The sampling used for inspections and tests 

performed on the dedicated items is discussed in Section 3.4.2.4 of this 

inspect on report. 

3.4.2.3 2 Virginia Power Purchase Order PO No. CNT 494177, dated March 2, 

1995, L ne Item No. 7, was issued to EP for the supply of 20 each, 1 1/2 

inches, chrome-moly, ASTM A 182, Grade F22, 3000 pounds, pipe tees. EP issued 

8 

Virginia Electric and Powet- (Virginia Power) purchase order (PO) 

60 
PO No. €256-01402, dated March 8, 1995, to a nonqualified supplier, PMW and 

purchased 25 commercial grade pipe tees for this Virginia Power PO. EP 

verified that markings on the tees were correct, reviewed the nonqualified 

manufacturer’s material test report for conformance with the material 

specification requirements, performed a hardness check and chemistry analyses 

on 5 of the tees, and dimensional inspections on 4 of the tees. The results 

of the overchecks performed by EP indicated that the attributes tested were in 

compliance with ASTM A 182, Grade F22, Class 3, specification requirements. 

The sampling used for inspections and tests performed on the dedicated tees is 

discussed in Section 3.4.2.4 of this inspection report. 

The NRC inspectors discussed the dedication of the tees with EP and identified 

an inconsistent selection of tests to verify compliance to A 182 material 

specification requirements. Previously, EP supplied Virginia Power 

ASTM A 182, Grade F 22, 90 degrees, elbows (see Section 3.4.2.3.1 of this 

report) and performed the following additional tests to verify compliance to 

the A 182 material specification requirements: (a) tensile testing of the 

material to determine the material’s tensile and yield strength, and (b) 

elongation and reduction OS area tests to determine the material’s ductility. 

Based on discussions with EP, it appears that EP does not always select the 

tests used to dedicate an item based on ensuring that the dedicated item was 

in compliance with the applicable material specification. At times, it 

appears that the selection of tests was based on convenience. According to 

EP, tensile tests were not performed on the tees because tensile test 

specimens could not be obtained from the tees. 

that the inability to obtain test specimens from the tees was not an adequate 

basis for not verifying the material’s strength and ductility. 

noted that the observed inconsistencies in selection and verification of 

critical characteristics appeared to be attributed to EP’s failure to develop 

written procedures or instructions that provided requirements for determining 

the types of tests to be selected and appropriate verification methods to be 

used to ensure that a dedicated CGI was in compliance with applicable material 

specification and customer requirements (see Section 3.4.2.1 of this report). 

EP informed the NRC inspectors that, on occasion, it has contacted customers 

and informed them that certain tests, required by the material specification, 

could not be performed. 

required testing were considered a part of the CGI dedication package and 

needed to be documented. 

The NRC inspectors informed EP 

The inspectors 

The NRC inspectors noted that customer waivers of 

3.4.2.3.3 Virginia Power Purchase Order PO No. CNT 479183, dated September 

21, 1994, Line Item No. 3, was issued to EP for the supply of 4 each, 1 inch, 

schedule lOS, belled ends, ASTM A 403, Type 304L, stainless steel, 90 degrees, 

el bows. EP issued PO No. 98177, dated June 6, 1992, to a nonqual ified 

supplier, Bestweld Inc. (Bestweld), for the manufacture of 19 commercial grade 

pipe elbows for this Virginia Power PO. 

manufacturing these elbows. EP purchased the pipe used to manufacture the 

elbows from Sandvik Steel (Sandvik), a qualified EP supplier. Sandvik, an 

EP supplied Bestweld pipe used for 

9 

61 
ASME Quality Systems Certificate Holder, supplied the pipe in accordance with 

the requirements of Specification ASTM/SA A-312, Type 304L, as ASME Section 

111, Class 2, material. 

EP verified that markings on the elbows were correct, reviewed the 

nonqualified manufacturer's (Bestweld's) material test report for conformance 

with the material specification requirements, and performed a chemical 

analyses and tensile test (tensile strength, yield strength, and elongation) 

on one elbow. The sampling used for inspections and tests performed on the 

dedicated items is discussed in Section 3.4.2.4 of this inspection report. 

A1 though the e? bow tensile test results met the requirements of Specification 

A 403, Type 304L, the NRC inspectors questioned EP's acceptance of these test 

results. 

annealing heat treatment at a temperature of 1900 degrees F minimum followed 

by rapid water quench to below 800 degrees F in less than three minutes. This 

treatment resulted in the pipe having a yield strength of 42,200 psi. 

According to the certification received from Bestweld, the elbows were bright 

annealed at 1925 degrees F (+,- 25 degrees F) and, as permitted by 

Specification A 321, Bestwell certified that the yield strength was 

essentially the same as reported by Sandvik because Bestweld subjected the 

elbows to the same annealing temperature as Sandvik. 

performed its independent tensile test on one of the elbows, a reported yield 

strength of 73,000 psi was obtained. Because the Bestweld material 

certification listed a yield strength for the elbows as 42,200 psi and the EP 

test identified a yield strength of 73,000 psi, the NRC inspectors questioned 

the validity of this test for demonstrating adequacy of Bestweld's 

manufacturing process. 

The pipe used to manufacture the elbows was subjected to a final 

However, when EP 

The NRC inspectors determined that EP appears to have manufactured the tensile 

test specimen from one of the elbows improperly. 

Procedure, QCP 26 requires tensile tests to be performed in accordance with 

specification A/SA 370, "Specification for Test Methods and Definitions for 

Mechanical Testing of Steel Products." The test specimen was obtained by 

cutting a strip from the outer radius of the elbow and then flattening the cut 

strip and machining/cutting the strip to the correct dimensions to form the 

test specimen. It appeared that the requirements of Section 6.3, "Procurement 

of Test Specimens," of A 370 were not followed by EP. 

in part, that when test specimens are manufactured care shall be taken to 

remove by machining all cold worked areas from the section used in evaluating 

the test. In this instance, EP appeared to have cold-worked, by flattening 

the outer radius of the elbow, the test specimen over its entire length. The 

cold working could have resulted in EP reporting a yield strength that may not 

be representative of the actual yield strength of the elbows supplied. 

NRC inspectors concluded that EP failed to comply with A 370 requirements in 

manufacturing the tensile specimen. 

EP's Tensile Testing 

Section 6.3 requires, 

The 

(Nonconformance 99900866/95-01-03). 

EP informed the NRC inspectors that originally it had requested Bestweld to 

subject a piece of the supplied pipe to the same heat treatment that the 

elbows received. However, this piece of pipe was lost and according to EP, 

10 

62 
the only way to verify that Bestweld, a nonqualified supplier, had properly 

manufactured and heat treated the elbows was to perform a tensile test on an 

elbow. The NRC inspectors expressed concern that, because the piece of pipe 

had been lost, and the tensile test could not be performed in accordance with 

Specification A 370 requirements, EP attempted to verify the item’s tensile 

properties using a nonconforming specimen. It was further discussed that EP 

has no procedural guidance for determining the proper methods for verifying 

critical characteristics. 

The NRC inspectors also determined that EP does not compare any of its test 

results to the nonqualified supplier’s material certifications and bases its 

certification on the test results meeting the applicable material 

specification requirements. 

the items received from a nonqualified supplier to be of one heat or lot and 

only performs tensile tests on one item from the nonqualified heat or lot. 

The NRC inspectors expressed concern over EP practice of not comparing its 

test results to those received from the nonqualified supplier. EP’s 

definition and use of nonqualified supplier’s heat and lots and its failure to 

compare the nonqualified supplier’s material certification to its test results 

as part of its dedication process is discussed in Section 3.4.2.4 of this 

inspection report. 

EP informed the NRC inspectors that it considers 

The NRC inspectors also determined that EP had supplied an additional 5 elbows 

for Virginia Power PO CNT 492269 using the same tensile test results supplied 

with Virginia Power PO CNT 479183. 

3.4.2.3.4 Tennessee Valley Authority (TVA) PO P-93N3F-41823E-002, dated 

February 22, 1995, Item 1, was issued to EP for the supply of 3/4 inch 

diameter, 40 feet total, A 276, Type 304, cold finish, round metal bar. EP 

issued PO No. E256-01313, dated February 23, 1995, to a nonqualified supplier, 

Centerless Grinding Service (Centerless), and purchased 48 feet (4 each, 12 

feet lengths) of the 3/4 inch diameter A 276, Type 304, commercial grade round 

metal bar. 

had purchased the bars from Huta Baildon, a Polish bar manufacturer. EP 

verified markings and performed dimensional checks on each bar, reviewed the 

nonqualified distributor’s, supplier’s and manufacturer’s certifications, 

performed chemical analyses on 2 of the bars, and performed tensile tests and 

an intergranular corrosion susceptibility test on one bar. 

were not subjected to any material overchecks. The results of the overchecks 

performed by EP indicated that the attributes tested were in compliance with A 

276, Type 304, requirements. 

dedicated bars is discussed in Section 3.4.2.4 of this inspection report. 

Centerless purchased the bars from Green Bay Supply Company which 

Two of the bars 

The sampling used the tests performed on the 

3.4.2.3.5 Virginia Power PO No. CNT-485585, dated November 29, 1994, Item 3, 

was issued to EP for the purchase of 1 piece of 1 inch thick X 48 inches X 

48 inches, A 516, Grade 70, plate. EP issued PO No. E256-00778, dated 

December 5, 1994, to a nonqualified supplier, Ferro Union, for a piece of 1 

inch thick X 48 inches X 96 inches A 516, Grade 70 piece of plate. 

Union purchased the plate from Highveld. EP performed a visual and 

dimensional inspection on the plate received from Ferro Union and cut a 

Ferro 

11 

63 
48 inches X 48 inches piece for shipment to Virginia Power and performed a 

chemical analyses and tensile tests on another piece cut from the plate 

received from Ferro Union. 

indicated that the attributes tested were in compliance with A 516, Grade 70, 

requirements. 

The results of the overchecks performed by EP 

3.4.2.4 Material Sampling and Traceability 

EP informed the NRC inspectors that CGIs purchased from nonqualified suppliers 

and dedicated as basic components were inspected and tested by EP in 

accordance with the following sampling plans. 

Visual inspection is performed on 100 percent of the CGI items received. 

Dimensional inspection is performed in accordance with Attachment 5.1, 

"Lot Sampling Plan," to EP Procedure No. QCP 5, "Quality Control 

Inspection," dated January 7, 1994 and Attachment 20.1 "Lot Sampling 

Plan" of QCP 20, "Receiving Inspection", dated January 7, 1994. The NRC 

inspectors discussed with EP that this sampling plan appeared to have 

been taken from Table 7-2, "Single-Sampling Plans for Normal 

Inspection, I' of MIL-STD-105D7 "Sampl ing procedures and Tables for 

Inspection by Attributes" with an Acceptable Quality Level (AQL) of 10. 

BATCH SIZE 

2 TO 8 

9 to 15 

16 to 25 

26 to 50 

51 to 90 

91 to 150 

151 to 280 

etc. 

etc. 

3201 to 10,000 

SAMPLE SIZE 

2 

3 

5 

8 

13 

20 

32 

-- 

-- 

200 

ACCEPT 

1 

1 

1 

2 

3 

5 

7 

- 

- 

21 

REJECT 

2 

2 

2 

3 

4 

6 

8 

- 

- 

22 

The NRC inspectors questioned the use of an AQL of 10 because, for 

example, this sampling plan would permit a sample size of 3 for a batch 

of 11 items to be accepted even though one of the items sampled is 

rejected. EP could not provide the NRC inspectors with the basis for 

its selection of this sampling plan or the selection of an AQL of 10. 

EP informed the NRC inspectors that even though QCP 5 permits the 

acceptance of a batch of items having one or more items rejected, its 

practice was to reject the entire batch or to inspect each remaining 

item when one of the items sampled was rejected. 

According to Section 6.7 of the EP Quality Systems Manual, stock 

material may be used for safety-related, non-ASME Section I11 purchase 

orders, provided selected tests are made to verify compliance with the 

applicable material specification. Section 6.7 continues by stating 

12 

64 
that such tests are conducted per heat or lot. 

dedication of material to be supplied under Appendix B, no surveys or 

audits are performed to verify heat traceability. 

inspectors that it performs one tensile test on each heat or lot of CGIs 

to be dedicated. EP further informed the NRC inspectors that it intends 

to follow its lot sampling plan in QCP 5 for the performance of chemical 

analyses. According to EP, its destructive sampling plans for 

dedicating CGIs are used for all dedications regardless of the size or 

type of product, or whether the CGIs were purchased directly from the 

manufacturer or through a distributor. 

The NRC inspectors discussed the use of EP's sampling methodology in 

detail during the inspection and asked EP to provide the bases for its 

destructive sampling plans as providing reasonable assurance that the 

CGIs not tested had the required material chemistry and strength and 

other properties required by the applicable material specification. EP 

informed the NRC inspectors that it had not established a quantitative 

basis, such as by the use of statistics, for its destructive sampling 

pl an. 

However, for the 

EP informed the NRC 

The NRC inspectors noted that Appendix B requires,in part, that a 

quality assurance program be established to provide controls over 

activities affecting quality of components and that compliance with this 

criterion would generally require EP to impose applicable provisions of 

Appendix B on their suppliers and to assure their compliance with these 

provi si ons . representative sampling is considered an activity affecting quality. 

The inspectors also noted that, for certain applications, the NRC staff 

has accepted commercial grade dedication as an alternative method to the 

programmatic quality assurance measures of Appendix B for assuring that 

i tems purchased from commerci a1 sources will meet the applicable 

purchase specification requirements. When the dedication process is 

based on sampling of material to verify critical characteristics, the 

sampling plan is expected to have an adequate technical basis and to 

provide a high level of assurance of the dedicated item's suitability 

for service. 

Maintaining materi a1 heat traceabi 1 i ty to assure 

The inspectors concluded that, without material traceability verification, 

EP's sampling practice for dedication of commercial grade material does not 

provide reasonable assurance that such material will conform to the 

specification requirements. (Nonconformance 9990866/95-01-02) 

The NRC inspectors discussed the overall effectiveness of the dedication 

process including the utilization of supplier performance history in 

development of sampling strategies and the evaluation of product markings as 

indicators of material heat homogeneity. It was also noted that product 

screening by chemical analysis of unqualified material and comparison of such 

analyses with the mill test report can be a valuable tool for Val idating 

materi a1 traceabi 1 i ty . 

13 

65 
The NRC inspectors also discussed with EP the evaluation of EP's tensile, 

chemistry, hardness, and other test results when they vary from those 

identified on the nonqualified supplier's material certification. The NRC 

inspectors determined that there are no requirements in EP's dedication 

program for comparing and evaluating EP's test results to the vendor-suppl ied 

test reports, and for identifying questionable vendor test reports when they 

are being supplied to a customer. 

documentation packages for dedicated material forwarded by EP to its customers 

and determined that the nonqualified supplier's certification used for the 

initial EP screening of material was included in these packages. The NRC 

inspectors noted that EP's practice of including the unvalidated supplier 

certification (stamped "QA OK" during the initial commercial grade material 

screening process) in the documentation packages supplied to customers without 

identifying to its customers what the "QA OK" stamped certification 

represented could mislead the customers to assume that EP had verified the 

authenticity and accuracy of these certifications. 

The NRC inspectors reviewed several 

The NRC inspectors concluded that EP had not developed adequate procedures or 

instructions to assure that items received from a nonqualified supplier were 

from the same manufacturer's heat or lot that was identified on the supplier's 

material certification. (Nonconformance 99900866/95-01-01) 

3.5 Nuclear Utilities Procurement Issues Committee (NUPIC) 

3.5.1 NUPIC Audit at EP 

The NRC inspectors determined that NUPIC had performed an audit of EP, 

formerly HUB Incorporated, Energy & Process Division (HUB/EP) on September 27 

through October 1, 1993. The NUPIC lead audit utility was Public Service 

Electric Gas Company (PSE&G) with auditors from Wolf Creek Nuclear Operating 

Company and Pacific Gas & Electric Company. 

Audit Report for HUB/EP, dated November 1, 1993, states, in part, that the 

audit team concluded that HUB/EP "has demonstrated their Revision 13, dated 2- 

13-93 of their Quality System Manual and implementing procedures are in 

compliance with the applicable provisions of 10 CFR 50 Appendix B, ANSI N45.2, 

10 CFR Part 21, IS0 9002, ASME Section 111, NCA 3800, and EPRI 5652." 

The summary section of the NUPIC 

The NUPIC audit report continued by further addressing commercial grade 

dedication and stated, in part, that HUB/EP had established an acceptable 

traditional commercial grade dedication program and that items procured 

commercial grade from unaudited/unsurveyed suppliers were dedicated for 

safety-related use by inspecting outbound material using an established lot 

sampling plan specified in Procedure QCP-5, Revision 7, and by performing 

tests on one sample from each heat or lot. The NUPIC audit report also 

14 

66 
indicated that specific critical characteristics for various items supplied by HUB/EP were provided to NUPIC auditors by several licensees. 

findings identified in the NUPIC audit report. 

There were no 

3.5.2 NRC Comments on NUPIC's Evaluation of EP's CGI dedication Program 

EP informed the NRC inspectors that the requirements in Section VI, 

"Examination, Tests, and Reports," of its Quality Systems Manual, and QCP 5, 

"Quality Control Inspection," for commercial grade dedication and sampling 

were the same during the conduct of the NRC inspection as when MUPIC performed 

its audit in 1993. Information, such as the following, was not provided in 

the NUPIC audit report and may be beneficial to NUPIC members (NRC licensees) 

in evaluating the use of items, dedicated by EP, in safety 

significant/cri tical pl ant appl ications. 

0 The NUPIC audit report did not provide any details of the sampling plan 

in QCP 5. 

105D, and to have an AQL of 10 which permits accepting the lot of items 

with 1 or more of the sampled items being rejected. As discussed in 

Section 3.4.2.4 of this report, this sampling plan did not appear to 

provide a high level of assurance of the dedicated item's suitability 

for service. 

could not provide the NRC inspectors its rationale for selecting a 

sampling plan having an AQL of 10. 

The sampling plan in QCP 5 appeared to be taken from MIL-StdHUB/EP had no documented basis for this sampling plan and 

0 HUB/EP had not established a documented basis to substantiate its 

assumption that a heat or lot of items supplied by a nonqualified 

manufacturer or distributor is a valid heat or lot and that testing only 

one item from each heat and lot provides reasonable assurance that the 

dedicated CGIs not tested met applicable procurement documentation 

requi rements. 

dedication program met EPRI NP 5652, "Guideline for the Utilization of 

Commercial Grade Items in Nuclear Safety Related Applications 

(NCIG-07)." However, as discussed in this inspection report, HUB/EP had 

not developed procedures that provided requirements for control1 ing CGI 

dedication activities such as (a) selecting critical characteristics to 

be verified, (b) identifying proper verification methods for confirming 

the acceptability of critical characteristics, and (c) ensuring that a 

heat or lot received from a nonqualified supplier was a valid heat or 

lot. 

0 The NUPIC audit report stated, in part, that HUB/EP commercial grade 

4 PERSONS CONTACTED 

The NRC staff participating in the inspection and EP personnel contacted 

during the inspection are listed below. 

15 

67 
Energy b Process Corporation 

t* Capallo, M 

+* Thornton, Jr. , C 

+* Walker, C 

t Kennedy, K 

+ Buffington, R 

Prasad, A 

t* Wolfe, C 

President and General Manager 

Quality Assurance Manager 

Operations Manager 

Sales 

Qual i ty Control Inspector 

Laboratory Manager 

QC Inspector Trainee 

U.S. NUCLEAR REGULATORY COMMISSION 

+* Potapovs, U 

+* Campbell , L 

Team Leader, VIS/TSIB 

Reactor Engineer, TQMB 

* Attended the entrance meeting 

+ Attended the exit meeting 

16 

68 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 2W5SOOO1 

June 20, 1995 

Hr. Mark Van Sloun, Vice President 

and General Manager 

Rosemount Nuclear Instruments , Incorporated 

12001 Techno1 ogy Drive 

Eden Prairie, Minnesota 55344-3695 

SUBJECT: NRC INSPECTION Ni3. 99900271/95-01 

Dear Mr. Van Sloun: 

This letter transmits a corrected copy of the subject report. The enclosed 

corrected copy rep1 aces U. S. Nucl ear Regul atory Commi ssion (NRC) Inspecti on 

Report 99900271/95-01 in its entirety for the NRC inspection activity that was 

conducted at the Rosemount Nuclear Instruments, Incorporated (Rosemount), Eden 

Prairie facility on January 23 through January 26, 1995. 

After reviewing the initial NRC inspection report, the Dow Corning Corporation 

(Dow Corning) informed the NRC, by letter dated May 2, 1995, of a 

misunderstanding that occurred during a telephone di scussi on and provided 

clarification of their comments that were contained in the report. The staff 

has corrected the affected section, on page 8 of the report, and identified 

the revised section with a vertical line in the margin of page 8. The NRC 

staff also reviewed the revised report and determined that the revised section 

of NRC report 99900271/95-01 did not affect any conclusions regarding the 

Rosemount sensor cell oil crystallization issue. 

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

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

No response to this letter or its enclosure is required. Should you have any 

questions, concerning this report, we will be pleased to discuss them with you. 

Sincerely, 

Special Inspection Branch 

Di vi si on of Technical Support 

Off i ce of Nucl ear Reactor Regul at i on 

Docket No. : 99900271 

Enclosure: As stated 

cc: See next page 

69 
Mr. Mark Van Sloun 2 

Mr. Kenneth E. Ewald 

Business Unit Manager 

Rosemount Nuclear Instruments, Incorporated 

12001 Techno1 ogy Drive 

Eden Prairie, Minnesota 55344-3695 

Mr. Arnold R. Huntress 

Associate Process Consultant 

Materi a1 s Development 

Dow Corning Corporation 

Midland, Michigan 48686-0995 

Mr. Paul Blanch 

135 Hyde Road 

West Hartford, Connecticut 06117 

Ernest Hadley, Esquire 

414 Main Street 

Post Office Box 3121 

Wareham, Massachusetts 02571 

70 
REVISED JUNE 20, 1995, TO REFLECT CORRECTIONS ON PAGE 8. 

U. S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF TECHNICAL SUPPORT 

REPORT NO.: 

ORGANIZATION: 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY: 

INSPECTION DATE: 

LEAD INSPECTOR: 

OTHER INSPECTORS: 

REVIEWED BY : 

99900271/95-01 

Rosemount Nuclear Instruments, Incorporated 

12001 Technology Drive 

Eden Prairie, Minnesota 55344-3695 

Mr. Kenneth E. Ewald, Business Unit Manager 

(612) 828-8241 

Designer, manufacturer, and suppl ier of temperature 

detectors and sealed gage, differential, and high-line differential pressure transmitters used extensively in 

nuclear safety-re1 ated applications. 

Shashi kant V. Athavale 

Instrumentation and Controls Branch 

Division of Reactor Control and Human Factors 

K. I; Parczewski 

Materials and Chemical Engineering Branch 

Division of Engineering 

?bhr 

Date 

APPROVED BY: 

Enclosure 

71 

REWISED JUNE 20, 1995, TO REFLECT CORRECTIONS OM PAGE 8. 
1 SUMMARY OF INSPECTION FINDINGS 

During this inspection, the NRC inspection team evaluated the adequacy of the 

process that Rosemount Nuclear Instruments, Incorporated (Rosemount), 

established related to the use and conditioning of the Dow Corning 704 

(DC 704) silicone fluid that is used by Rosemount for its Models 1153 and 1154 

safety-related pressure transmitters. The inspection was conducted in 

conjunction with an NRC Headquarter's staff review to assure that Rosemount's 

applicable safety-related pressure transmitters supplied to the U.S. nuclear 

industry would perform their safety function within their designed operating 

parameters with little probability of a potential for silicone oil 

crystallization below 70" Fahrenheit (F) applications. 

The inspection basis consisted of the following: 

Rosemount Nuclear Qual i ty Manual (NQM), Document Number 09000115, 

Revision C, dated June 1, 1994 

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

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

Federal Requlations (10 CFR Part 50) 

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

No violations or nonconformances were identified by the NRC inspection team 

during this inspection. 

1.1 Inspector Follow-Up Item 99900271/95-01-01 

As discussed below in Section 2.2, the NRC staff will review Chanhassen's 

procedure which it adopted to implement the provisions of 10 CFR Part 21 

during a future NRC inspection. (95-01-01) 

1.2 Inspector Foll ow-Up Item 99900271/95-01-02 

As discussed below in Section 3.10, the team had some concerns regarding the 

quality assurance (QA) program requirements imposed and implemented by the 

Rosemount transmitter sensor cell manufacturing activities at the Chanhassen 

facility which will be reviewed during a future NRC inspection. (95-01-02) 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

During an inspection of Rosemount conducted on February 1 through 4, and 

March 8 through 12, 1993, the NRC inspection team determined that certain of 

Rosemount's activities were in violation of NRC requirements, as specified in 

NRC inspection Report 99900271/93-01 and associated transmittal letters. 

Additionally, the inspection team found several weaknesses in Rosemount's QA 

program in both QA program establishment and implementation as required by NRC 

requirements contractually imposed on Rosemount by purchase orders (POs) from 

nuclear power reactor utilities. 

actions for the following violations and nonconformances were reviewed, 

evaluated and di sposi ti oned as stated bel ow. 

The adequacy of Rosemount's corrective 

2 

72 
2.1 Viol ation 99900271/93-01-EA94-049 (CLOSED1 

Contrary to Section 21.21, "Notification of failure to comply or existence of 

a defect and its evaluation," of the version of 10 CFR Part 21 in effect from 

January 1978 until July 1991, the Measurement Division of Rosemount, 

Incorporated (Rosemount) , failed to adequately establish and implement a 

procedure to effectively implement the provisions of 10 CFR Part 21. 

Rosemount procedure, Quality Implementation Procedure (QIP) 126(N), "Potential 

Defect or Deviation in Products for Nuclear Application," issued March 18, 

1981, did not ensure that deviations, as defined in 10 CFR 5 21.3(e), about 

which Rosemount did not have adequate knowledge of each of the applications at 

NRC licensee facilities, were either evaluated or transmitted to the 

applicable customer or licensee for evaluation. Specifically, between 1984 

and December 1988, Rosemount failed to properly inform licensees of a 

potential for a sensor cell oil-loss problem that could occur in its nuclearsafety-related 1150-series pressure transmitters which could have caused 

safety limits to be exceeded or caused substantial safety hazards to exist at 

1 icensee faci 1 i ti es. 

The 

The NRC inspectors verified that Rosemount's Procedure N-1626, "Implementation 

of lOCFR Part 21 for Deviations and Failures to Comply in Nuclear Products," 

Revision D, dated May 27, 1994, had been appropriately revised since the last 

NRC staff inspection at the Rosemount facility. Additionally, Rosemount 

committed to ensure that its communications with the NRC and its licensees are 

more complete and candid as indicated in its December 14, 1994, letter to NRC 

staff. As discussed in our February 15, 1994, letter, the NRC inspectors 

concluded that the applicable provisions of 10 CFR Part 21 should be 

effectively executed as a result of the revised Rosemount program controls. 

2.2 Viol ation 99900271/93-01-01 (CLOSED) 

Contrary to Section 21.21, "Notification of failure to comply or existence of 

a defect and its evaluation," of 10 CFR Part 21, Rosemount failed to establish 

or implement a procedure to ensure that the provisions of 10 CFR Part 21 were 

executed at its Chanhassen facility. As discussed below in Section 3.1, the 

Chanhassen, Minnesota, facility is currently a separate corporate entity from 

the Eden Prairie, Minnesota, facility. That is, Chanhassen is pdit of the 

Rosemount Measurement Division (RMD) of Rosemount Corporation and the Eden 

Prairie nuclear activities are under Rosemount Nuclear Instruments, 

Incorporated. However, at the time of the identification of the violation, 

Eden Prairie was part of RMD and was responsible for ensuring that Chanhassen 

had adopted procedures to effectively implement the provisions of 10 CFR 

Part 21. Since Eden Prairie is now a separate corporate entity from 

Chanhassen, it would not be responsible for ensuring the adequacy of 

Chanhassen's 10 CFR Part 21 procedure. 

However, during the NRC staff review of the circumstances surrounding 

violation 99900271/93-01-01, the team reviewed a copy of the procedure which 

Chanhassen has adopted to implement the provisions of 10 CFR Part 21, Pressure 

Department Procedure (PDP)-1320-3, "Nuclear Sensor Part 21 Reporting," 

3 

73 
Revision 6, dated January 19, 1995. Since the Chanhassen facility is now a 

separate corporate entity, Procedure PDP-1320-3 will be reviewed during a 

future NRC inspection at Chanhassen and is identified as NRC inspector followup item 99900271/95-01-01. 

2.3 Viol ation 99900271/93-01-02 (CLOSED) 

Contrary to Section 21.6,"Posting requirements," the 10 CFR Part 21 posting at 

Rosemount's Eden Prairie and Chanhassen facilities did not adequately describe 

the 10 CFR Part 21 regulation or the procedure adopted to implement 10 CFR 

Part 21. 

telephone numbers of personnel to whom reports were to be made. The 

inspectors verified that the postings at the Eden Prairie facility were in 

accordance with 10 CFR 4 21.6; therefore, this issue is closed. 

In addition, the postings were found to contain outdated names and 

2.4 Viol ation 99900271/93-01-03 [CLOSED) 

Contrary to Section 21.51, "Maintenance and inspection of records," of 10 CFR 

Part 21, Rosemount records regarding a review of suspect resistors used in 

Rosemount 710 DU products did not contain adequate information to enable the 

team to determine whether Rosemount customers were appropriately informed of 

the deviation. During this 1995 inspection, the team reviewed the documents 

which Rosemount discussed i n i t s April 1, 1994, response letter to the NRC 

letter, dated March 4, 1994, which transmitted the Notice of Violations. The 

April 1st letter refuted violation 93-01-03. Based on its record review, the 

team concluded that Rosemount appropriately informed their customers of the 

subject deviation. Therefore, as discussed in NRC's response dated June 13, 

1994, to Rosemount, this violation has been withdrawn and is closed. 

2.5 Nonconformances 99900271/93-01-04 - 99900271/93-01-09 (OPEN) 

These nonconformances were not reviewed by the NRC inspection team during this 

inspection activity; therefore, each nonconformance will remain open and will 

be reviewed during a future inspection. 

2.6 Inspector Follow-Up Items 99900271/93-01-10 - 99900271/93-01-15 (OPEN). 

These inspector follow-up items were not reviewed by the NRC inspection team 

during this inspection activity; therefore, each inspector follow-up item will 

remain open and will be reviewed during a future inspection. 

3 INSPECTION FINDINGS AND OTHER COMMENTS 

3.1 Rosemount ComDanies 

In the February-March 1993 time period, the Chanhassen and Eden Prairie 

facilities were both a part of the Rosemount Corporation's Measurement 

Division. Subsequent to the 1993 inspection, the responsibility of the 

safety-related transmitter came under Rosemount Nuclear Instruments, 

Incorporated (Rosemount) and the Chanhassen facil ity remained part of RMD. 

During the 1993 inspection, Chanhassen manufactured the pressure transmitter's 

(1) safety-related sensor cell, and (2) the nuclear printed circuit (PC) card 

4 

74 
assembly (basic components as defined in 10 CFR § 21.3). 

sensor cell was manufactured at the Chanhassen facility and the nuclear PC 

card assembly was being manufactured by Rosemount Aerospace, now a Division of 

the B.F. Goodrich Company, during the performance of the January 1995 NRC 

staff inspection. Therefore, Chanhassen (part of RMD) and Rosemount Aerospace 

were characterized and treated by Rosemount as separate sub-tier manufacturers 

of "basic components. 'I Rosemount 1 ocated its nuclear power plant business 

activities at the Eden Prairie facility where it procures pressure transmitter 

parts; and assembles, tests and supplies Rosemount safety-related pressure 

transmitter Models 1152, 1153 and 1154 to the nuclear industry. 

The safety-related 

3.2 Entrance and Exit Meetinqs 

The NRC inspectors conducted an entrance meeting on January 23, 1994, with 

Rosemount staff and discussed the scope of the inspection. On January 26, 

1994, the team leader discussed the NRC inspection results. 

3.3 Backqround 

The silicone oil used in the Rosemount pressure transmitter Models 1153 and 

1154 sensor cells is a Dow Corning Corporation silicone oil. 

1994 review of Rosemount's response to NRC staff inspection findings 

delineated in Inspection Report 999000271/93-01, the staff identified a 

cautionary note in the DC 704 product data sheet which stated: "Caution: Dow 

Corning 704 diffusion pump fluid may crystallize if subjected to temperatures 

below 21°C (69.8"F). 

temperature will return the fluid to its original state." However, in 

conflict with the Dow Corning cautionary note, the NRC staff found that the 

Rosemount transmitter Models 1153 and 1154 product data sheets stated that the 

transmitter's lower operating design limit was t40"F and its lower qualified 

storage 1 imit was -40°F. Therefore, since the Rosemount operating and storage 

lower limits were below the crystallization temperature which Dow Corning 

stated i n i t s DC 704 data sheet, the YRC staff initiated a review. 

During a May 

If crystallization occurs, two hours of 50°C (122°F) 

The staff's concern was based on the postulated consequence that potential 

silicone oil crystallization in a safety-related transmitter could render the 

transmitter inoperable and compromise safety, depending on the number and 

location of affected transmitters. A Rosemount September 28, 1994, letter to 

the NRC forwarded Rosemount's August 12, 1994, analysis of the matter 

(Rosemount Report D9400078). Based on its analysis, Rosemount concluded that 

DC 704 was still the most appropriate fluid for use in the Model 1153 and 1154 

transmitters. The Rosemount analysis discussed Rosemount's initial 

engineering actions in 1980 regarding the DC 704 oil's potential for 

crystallization and referenced its engineering report, Rosemount Report No. 

11805A, which addressed and summarized the DC 704 crystallization issue. The 

Rosemount 1980 report concluded that crystallization was not a concern as long 

as certain conditions were met. As a result of the NRC staff review of the 

September 28, 1994, Rosemount letter, the NRC concluded in a December 9, 1994, 

letter to Rosemount that the manufacturing process (DC 704 preconditioning) 

used by Rosemount will prevent crystallization of the oil in the transmitters 

5 

75 
prior to shipment. Further, Rosemount's analysis of the transmitter oil 

crystallization concern acceptably demonstrated that the potential problem is 

adequately addressed by initial and periodic transmitter performance testing 

by the licensees. 

3.4 Sensor Cell Operation 

Rosemount uses the DC 704 oil in its transmitter's sealed sensor cell 

diaphragm compartments. 

which are separated by a center diaphragm (sensing diaphragm), with isolating 

diaphragms at each end of the cylindrical sensor cell. Each of the three 

diaphragms are circumferentially welded to form the diaphragm's seal. In 

operation, process pressure is transmitted through either of the isolating 

diaphragms to the DC 704 oil which then displaces the sensing diaphragm in the 

center of the sensor cell. The movement and position of the sensing diaphragm 

is detected by capacitor plates located on both sides of the sensing diaphragm 

and a differential capacitance between the sensing diaphragm and the capacitor 

plates is converted to an electrical signal. The amount of sensor cell fill 

fluid on either side of the sensing diaphragm is equal to approximately three 

drops of DC 704 oil. As a result, the properties of the fill fluid 

(viscosity, dielectric constant and density) are important to the performance 

of the transmitter. 

Every sensor cell has two separate oil compartments 

3.5 DC 704 Associated Process Controls 

The NRC staff inspectors observed the Rosemount manufacturing process controls 

associated with DC 704 and reviewed the procedures used to control the 

processes. The inspectors noted that the procedure used to "condition" Dow 

Corning silicone oil (DC200 was the first silicone oil to be outgassed), was 

initially written in 1979. Rosemount Procedure 01153-3045, "Outgas 

Procedure," Revision L, dated June 4, 1992, delineates the oil processing 

steps which are performed by Rosemount on all DC 704 silicone diffusion pump 

fluid received for use in its Models 1153 and 1154 pressure transmitters. 

According to Rosemount staff, the main purpose of the outgassing process is to 

ensure maximum noncompressability, improve the dielectric characteristics and 

to filter out any contaminants. According to the Rosemount staff, moisture is 

removed and the potential for crystallization of silicone oil is greatly 

reduced as a benefit of the outgassing process. 

In the DC 704 conditioning process, the oil is maintained at 150°F for 

approximately 16 hours under a partial vacuum condition at a specific 

pressure. The oil is continuously stirred during the heating and evacuating 

process and the outgassed DC 704 oil is later piped in to sensor cells which 

have been vacuum evacuated at a specific temperature and pressure for several 

hours. After the sensor cells are filled with the DC 704, their fill oil 

tubes are crimped and welded. The team noted that all of the identified 

procedures for the DC 704 oil and the filling of oil in to the transmitter 

cell were in accordance with Rosemount's stated program in their letter to the 

NRC. The team concluded that the procedures have adequate guidance for the 

Rosemount technicians to accomplish the DC 704 oil conditioning to ensure that 

dissolved gases and moisture were removed from the oil. It appeared that 

6 

76 
Rosemount's process was well established, the operator was aware of the 

procedural requirements and the intent of the process control. Additionally, 

all activities that were witnessed appeared to be accomplished in accordance 

with the established procedures. 

3.6 Rosemount Report 11805A 

The team reviewed the Rosemount engineering report, Rosemount Report 11805A, 

"Re-evaluation of Suitability of Dow Corning 704 Oil for Use in Model 1152 or 

1153 Pressure Transmitters, I' to assess its Val idi ty and soundness. According 

to the scope, this report described the results of an engineering evaluation 

undertaken to determine if a problem existed with the use of DC 704 fluid in 

pressure transmitters qualified for Class 1E applications in nuclear power 

plants. 

fluid movement within the transmitter, that the oil volume was very small and 

fluid viscosity is important. Viscosity of the oil would increase at low 

temperatures and resul t in slow response. Crystal 1 i zat i on or incipient 

crystallization (irreversible long-range order) would also show up at low 

temperature as an increase in apparent viscosity which would not disappear 

when the oil returns to room temperature. Measurement of the time constant of 

the transmitter unit therefore provides the prime measure of the occurrence of 

crystallization. 

change of density due to crystallization, the dielectric characteristics would 

a1 so change resulting in a marked zero or span shift as a second indication of 

the occurrence of crystallization. 

testing observations indicated any effect which could be interpreted as due to 

crystallization of DC 704 oil. 

The report stated that the response of the transmitter depended on 

Additionally, the report stated that if there were a marked 

The report concluded that none of its 

The Rosemount Report stated that, crystallization was not a matter for concern 

when using DC 704 oil as a pressure transfer fluid for nuclear qualified 

pressure transmitters, so long as: (a) operating temperature is greater than 

40°F and (b) the present fill method, which insures the absence of air and 

moisture, is maintained. The report indicated that the only time condition 

(a) might be violated would be during storage before installation. As stated 

in the Report, this possibility was tested on a non-operational basis during a 

storage temperature-aging program where Rosemount took 6 Model 1153 Series B 

transmitters (use DC 704 oil) and exposed them to 20 cycles utc.\-en -55°F and 

+135"F with one cycle equal to 3 hours at -55°F and 3 hours at 135°F. 

Rosemount did not identify any anomalies attributed to the cycling, and 

maximum zero or span shift after cycling was approximately one percent. 

Rosemount identified that these test results were within the expected shift. 

3.7 Discussion with Dow Corninq Corporation Staff 

On January 25, 1995, the NRC inspectors conducted a telephone discussion with 

a Dow Corning Corporation technical representative who has been associated 

with the initial research, development and production of Dow Corning silicone 

fluids since the 1950s. The Dow Corning technical representative provided the 

team members with the history of the DC 704 silicone oil development and 

experience over the approximate 40 year existence of the fluid. For example, 

the Dow Corning technical representative indicated that little is known about 

the mechanism of crystallization of the DC 704 fluid, but in their opinion, 

7 

77 
moisture is the suspected cause for oil crystallization due to a molecular 

alignment or orientation. The Dow Corning technical representative stated that 

agitation and geometry of the vessel containing the fluid may have some effects, 

but they are of lesser importance. The team noted that agitation as a possible 

factor for crystallization was addressed in Rosemount's engineering analyses, 

D9400078, Revision A, dated August 12, 1994, and D9400129, Revision A, dated 

November 22, 1994, but was not confirmed by either the experiments conducted by 

Rosemount or Dow Corning. 

The Dow Corning technical representative stated that crystal1 ization can be 

eliminated by heating the oil at or above 122°F for 2 hours and DC 704 fluid has 

a very high thermal stability and heating it to 150"F, as Rosemount does, for an 

extended period of time would not cause degradation. 

radiation on DC 704 were also discussed with the Dow Corning technical 

representative and he stated that polymerization would be expected to occur upon 

exposure to radiation in proportion to radiation dose, and that a high dose would 

be required to cause a noticeable change. 

crystallization is very unlikely. The Dow Corning technical representative also 

stated that DC 704 fluid has high stability to radiation and formation of large 

polymerized molecules which could act as crystallization seeds is not very 

probabl e. 

The effects of gamma 

However, the possibility of 

In summary, the Dow Corning technical representative indicated that the 

probability that a clean (filtered), conditioned DC 704 silicone oil will 

crystalize is very small. That is, the potential of crystallization disappears 

if moisture is outgassed prior to use of the DC 704 and there is no implication 

of any higher probability of crystallization when used in pressure transducers. 

The Dow Corning technical representative also stated that if there was an 

initiation of a fluid phase change in which the DC 704 molecules oriented 

themselves in a crystal lattice pattern prior to being received by the end user, 

such as Rosemount, the heating and outgassing at 150°F for 16 hours that 

Rosemount performs would return the fluid back to a normally stable state. 

3.8 Functional Testincr of Hodel 1153 8 1154 transmitters 

The team reviewed Rosemount's production test procedures and test logs. 

noted that since 1979, Rosemount has been performing functional cycle testing of 

its Model 1153 and 1154 pressure transmitters. 

procedures require a minimum temperature test cycle at 40°F. The lower end 40°F 

test temperature is lower than the identified crystallization threshold 

temperature of 69.8"F. 

testing activities and reviewed the test results. 

team. 

and assemblies. 

which allows various monitoring and technical data retrieval. 

The team 

The team noted that Rosemount 

The NRC team observed one of Rosemount's lower end 

No anomalies were noted by the 

Rosemount has computerized its pressure transmitter test chamber program 

The program is tied into a Rosemount mainframe computer system 

3.9 Trackinq for Future Failures due to Oil Crystallization 

Rosemount, in their letter to NRC dated September 28, 1994, committed that they 

would send copies of their analysis to each of their customers to inform them 

about the potential of oil crystallization. 

Rosemount informed their customers and made them aware of this problem by 

The inspection team verified that 

8 

78 
their customer transmittal letters, dated December 1, 1994. In its letter to the 

NRC, Rosemount indicated that it intended to monitor and track future problems 

relating to oil crystallization. The team reviewed Rosemount's procedures for 

tracking identified crystallization problems and its associated failure analysis 

form. The team found that Rosemount's measures that it has established to track 

fill oil crystal1 ization failures appeared adequate. 

3.10 Sensor Cell Manufacturer 

The team reviewed Rosemount procurement document QA requirements which were 

applicable to the sensor cell manufacturing activities at the Chanhassen facility 

and identified a concern regarding the adequacy of the quality assurance (QA) 

requirements that Chanhassen had applied to its sensor cell manufacturing 

activities. The team found that the quality requirements imposed on Chanhassen 

by Rosemount procurement speci f i cat i on NPS-01 required Chanhassen to establ i sh 

and maintain a quality assurance program that complies with the requirements of 

NPS-01. However, Rosemount allowed an exemption to Sections 1 and 10 which dealt 

with QA program organization and inspection, respectively. The team also noted 

that Chanhassen has adopted International Standard ISO-9001, "Quality systems - 

Hodel for qual i ty assurance i n des i gn-devel opment , product i on, i nstal 1 at i on and 

servicing," Section 4, "Organization," of which requires that inspection, test 

and monitoring be performed by "personnel independent of those having direct 

responsibility for the work being performed." Appendix B has similar 

requirements. Rosemount staff indicated that Chanhassen had not implemented that 

requirement. 

document QA requirement exemptions took precedence over Chanhassen's quality 

system requirements or Appendix B to 10 CFR Part 50 requirements. The inspectors 

will examine the basis for Rosemount's acceptance of Chanhassen's lack of 

implementation of independent inspection and adequacy of its overall program 

related to sensor cell manufacturing during a future NRC staff inspection. 

It was not clear to the team whether Rosemount's procurement 

Additionally, an NRC letter dated March 4, 1994, transmitting a Notice of 

Violation (Report 99900271/93-Ol), irldicated that Rosemount did not establish or 

implement an independent QA inspection or verification function in its sensor 

cell and printed circuit (PC) card manufacturing areas at the Chanhassen 

facility. Since the current absence of independent inspection at Chanhassen may 

be related to lack of adequate corrective action, the NRC staff will review the 

adequacy of the overall QA program corrective action which Rosemount has 

performed. These issues are identified above as inspector follow-up item 95-01- 

02, and will be reviewed during a future inspection. 

3.11 Demonstration of Crystallization 

On January 23, 1995, the NRC inspectors requested Rosemount staff to demonstrate 

crystallization of a DC 704 oil sample and the Dow Corning Corporation 

recommended method of returning the DC 704 oil back to its original state. On 

January 24, 1995, Rosemount staff scraped a few specks of a Dow Corning provided 

sample of crystallized DC 704 and placed the specks in a glass bottle which 

contained a sample of approximately three ounces of new DC 704. 

inspectors placed the sample in a Styrofoam container surrounded by ice cubes to 

reduce the temperature. 

and within 75 minutes the DC 704 was completely crystallized. The sample became 

The NRC 

Within a few minutes, crystallization started to occur 

9 

79 
solid and semi-transparent. 

sample was removed from the ice bath and placed on the inspectors conference room 

table for approximately four hours at room temperature. The sample did not fully 

restore to its pre-crystallization state, but became mushy and appeared to be 

cloudy. Later, the sample was heated at approximately 127°F for 2 hours. At the 

end of two hours, the oil became fluid, clear and it appeared to have returned to 

its original state. The sample was than placed in a refrigerator overnight at a 

maintained temperature of approximately 43°F. On January 25, 1995, the team 

noted that the sample had not returned to the crystallized state and appeared to 

be in its original clear fluid condition. 

At the end of 75 minutes, the crystallized DC 704 

This demonstration indicated that the DC 704 fluid can be returned to its 

original state with as little as two hours of approximately 127°F temperature and 

that recrystallization is not easily obtainable. 

demonstration, it would suggest that performing the Dow Corning recommendation 

would reduce the possibility of future crystallization. 

Although this was only a 

3.12 Conclusions 

Based on the team’s review of the Rosemount procedures, manufacturing process and 

personal interviews with the Rosemount manufacturing and engineering staff, the 

NRC staff concluded that Rosemount’s actions in 1980 regarding the DC 704 

cautionary note adequately addressed its 10 CFR Part 21 responsibilities and its 

engineering basis for its Model 1153 and 1154 design low temperature operation 

were valid. Additionally, the team determined that Rosemount provided their 

customers a summary of their engineering analysis that was provided to the NRC 

and Rosemount had its appl icable manufacturing process controls appropriately 

implemented. The team also concluded that Rosemount’s conditioning of the DC 704 

oil before its use should remove any existing seeds which could cause 

crystallization. Although complete immunity from crystallization may not be 

achievable, the information provided by Dow Corning and the Rosemount testing and 

industry historical data indicate that there is little probability of 

crystallization of DC 704 oil. 

4 PERSONS CONTACTED 

Rosemount Nuclear Instruments. IncorDorated 

G.D. Anderson Nuclear QA Supervisor. 

T. Beaudette Instrument Builder 

Stuart C. Brown Engineering Supervisor 

Marc Bumgarner Product Design Engineer 

Ken Ewald Business Unit Manager 

Tim Layer Marketing Manager 

Rob Lind Manufacturing Engineer 

Jerry Valley Nuclear Q.A. Manager 

Mark Van Sloun Vice President 

and General Manager 

Dow Corninq Corporation 

Arnold R. Huntress (by telephone) Associate Process Consultant 

Materi a1 s Devel opment 

10 

80 
UNITED STATES 

N U CLEA R REG U LATO RY CO M M I SS IO N 

WASHINGTON, D.C. 20555-OOOl 

June 16, 1995 

Mr. Kris Weger 

Senior Vice President 

and General Manager 

Unistrut Corporation 

35680 C1 inton Street 

Wayne, Michigan 48184 

Dear Mr. Weger: 

SUBJECT: NOTICE OF VIOLATION AND DEMAND FOR INFORMATION: 

NRC INSPECTION REPORT 99900362/90-01 AND 

REPORT OF INVESTIGATION CASE NO. 4-89-016 

This letter transmits a Notice of Violation based on the results of the U.S. 

Nuclear Regulatory Commission (NRC) inspection of Unistrut Corporation (UC) , conducted at your facility in Wayne, Michigan, on December 10-14, 1990, the 

inspection report, and the synopsis of an investigation conducted from 

approximately November 1989 until August 1992. The NRC, as part of an ongoing 

scrutiny of fasteners being supplied to the nuclear industry, received 

information that raised concerns pertaining to safety-re1 ated basic components 

that UC supplied to the nuclear industry. 

subject inspection and investigation. 

Therefore, the NRC conducted the 

During the inspection, the NRC staff determined that UC failed to perform the 

required evaluation of deviations or to inform NRC licensees or purchasers so 

that they could perform or cause an evaluation to be performed pursuant to the 

provisions of Part 21 of Title 10 of the Code of Federal Reaulations (10 CFR 

Part 21). This failure to comply constitutes a violation of NRC regulatory 

requirements and, therefore, we are issuing a Notice of Violat’x 

(Enclosure 1). The team a1 so identified several nonconformances related to 

the implementation of the UC quality program; these are discussed later in 

this letter. 

discussion of the areas examined during the inspection and our findings. 

The inspection report (Enclosure 2) contains a detailed 

The Office of Investigations (01) also examined those concerns, as described 

in the enclosed synopsis taken from 01’s Report of Investigation Case No. 

4-89-016 (Enclosure 3). 

with UC actions have recently been completed; the enclosed inspection report 

and Notice of Violation are being issued herewith. 

The 01 investigation and other reviews associated 

On the basis of an 01 interview on June 13, 1990, with UC management (Vice 

President and General Manager , Production Engineering Manager , and Qual i ty 

Assurance Manager) , the NRC staff advised UC that certificates of conformance, 

issued with fasteners supplied to the nuclear industry as safety-related basic 

components, contained invalid certification statements. 

supplied by UC as safety-related basic components certified to comply with NRC 

regulatory requirements when, in fact, the fasteners were commercial-grade because UC had not dedicated them for use as basic components. As a result of 

The fasteners were 

81 
K. Weger -2- 

this NRC action, on June 21, 1990, UC advised all of its authorized service 

centers that on June 15, 1990, UC had suspended certification of products as 

complying with 10 CFR Part 50, Appendix B and 10 CFR Part 21. According to 

UC, changes were made to its quality program and on September 26, 1990, UC 

advised all of its nuclear customers that its quality program (which UC 

asserted complied with Appendix B to 10 CFR Part 50) for safety-related metal 

framing products was "back on line." 

Subsequently, in December 1990, an NRC inspection team, led by Steven M. 

Matthews and comprising the other inspectors named in the report, conducted 

the subject inspection to determine whether the products supplied to the U.S. 

nuclear industry actually complied with the regulatory requirements for basic 

components. The products were supplied with UC certificates of conformance 

that certified compliance with the requirements of 10 CFR Part 50, Appendix B 

and 10 CFR Part 21. 

areas related to UC's manufacturing and quality assurance process. 

During this inspection, the team also reviewed other 

The most significant inspection finding, as described in the enclosed Notice 

of Violation, was the failure of UC to perform the required evaluation of 

deviations or to inform 1 icensees or purchasers so they could perform or cause 

an evaluation to be performed pursuant to the provisions of 10 CFR Part 21. 

The team reviewed a number of purchase orders (POs) for fasteners, issued to 

UC from licensees, which imposed on UC the requirements of 10 CFR Part 50, 

Appendix E? and 10 CFR Part 21. UC supplied those customers with fasteners 

which bore certificates of conformance that certified compliance with the 

licensees PO requirements. 

UC advised the team during the inspection that in October 1990, UC met several 

times with its sole fastener supplier, General Fastener Company (GF) of 

Livonia, Michigan, to discuss the capability of GF to supply UC safety-related 

fasteners that comply with NRC requirements. On October 26, 1990, GF decided 

that it could not supply fasteners to UC certified as safety related because, 

according to GF, it did not have, nor had it ever had, a quality program that 

complied with 10 CFR Part 50, Appendix B. 

not continue to supply fasteners to the nuclear industry that would be 

certified to comply with NRC requirements for safety-related basic components. 

UC subsequently decided it could 

Although UC stated during the inspection that it had not supplied fasteners 

certified as safety related to the nuclear industry since October 26, 1990, UC 

had also neither informed licensees of this decision nor rescinded its letter 

of September 26, 1990. Moreover, as of December 14, 1990, UC had neither 

evaluated its departure from the technical requirements of the licensee POs 

for the fasteners supplied before October 26, 1990, nor had UC informed 

licensees or purchasers of the deviations. 

violation of NRC requirements and, furthermore, demonstrated UC's apparent 

lack of regard for NRC requirements. 

This situation constituted a 

You should be aware that the NRC considers a willful violation to be a 

significant regulatory concern. Following the events described above, the NRC 

issued the regulation in 10 CFR 50.5 that provides for enforcement actions 

82 
K. Weger -3- 

against any individual who, through deliberate misconduct, places or could 

have placed an NRC licensee in violation of NRC requirements. 

aware that, should you engage in deliberate misconduct in the future, you may 

be subject to individual enforcement action pursuant to 10 CFR 50.5. 

You should be 

We have since observed that, even though you stated during the inspection that 

UC had not supplied fasteners certified as safety related to the nuclear 

industry since October 26, 1990, UC was listed under the headings "Products 

and Services" (including the categories of bolts, cable tray hangers and 

supports, fasteners, fittings, nuts, pipe hangers, and studs) and "Suppliers," 

in addition, a full-page advertisement on page 18 in the October 1991 issue of 

Nuclear Plant Journal, Volume 9, No. 6, and under the same headings and 

categories, UC continues to be listed in the November 1994 issue, Volume 12, 

No. 6. 

there is reasonable assurance that UC is conducting these activities under NRC 

jurisdiction in accordance with NRC requirements. Accordingly, pursuant to 

Sections 161c and 1610 of the Atomic Energy Act of 1954, as amended; Section 

206(d) of the Energy Reorganization Act of 1974, as amended; and 10 CFR 2.204, 

you are required to 

Therefore, further information is needed to enable us to determine if 

(3) 

(4) 

Inform the NRC whether or not UC has reentered the business of supplying 

basic components as defined in 10 CFR Part 21 for any period of time 

since October 26, 1990, and if it has, give the extent and nature of 

such business activity, including the dates of such activity. 

If UC has supplied basic components as defined in 10 CFR Part 21 since 

October 26, 1990, submit a detailed explanation of the actions taken to 

ensure that the basic components comply with NRC requirements; include 

an explanation of UC compensatory actions taken to correct the invalid 

certificates of conformance issued before October 26, 1990, as described 

above. 

If UC has not supplied basic components pursuant to 10 CFR Part 21 since 

October 26, 1990, submit a one-time, written notice at least 30 days 

before it resumes such activities to the U.S. Nuclear Regulatory Commission, ATTH: Document Control Desk, Washington, D.C. 20555-0001 

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

Technical Support, Off i ce of Nucl ear Reactor Regul at i on, U . S . Nucl ear 

Regulatory Commission, Washington, D.C. 20555-0001. 

If UC does not agree to submit the notice specified in (3) above, it 

should so state and provide the basis for not submitting the notice. 

You are required to respond to this Demand for Information in accordance with 

10 CFR 2.204 within 30 days of the date of this letter. 

In accordance with the "General Statement of Policy and Procedures for NRC 

Enforcement Actions," 10 CFR Part 2, Appendix C (1990), Violation 90-01-01 

described in the enclosed Notice of Violation has been classified as a 

Severity Level I11 violation (Supplement VII, Subsection C.5) because UC 

83 
K. Weger -4- 

failed to evaluate its departure from the technical requirements of the 

licensee POs, such that, if an appropriate evaluation had been made as 

required, a 10 CFR Part 21 report would have been made. 

Even though other violations related to 10 CFR Part 21 were noted by the team 

in Report 99900362/90-01, only violation 90-01-01 is cited in the Notice of 

Violation for the following reasons: 

(1) The NRC has adequately addressed the safety issues through its issuance, 

on April 1, 1991, of Information Notice 91-25, "Commercial-Grade 

Structural Framing Components Supplied As Nuclear Safety-Related 

Equipment . 'I 

(2) The actions of UC preceded the NRC issuing guidance on commercial-grade 

dedication in Generic Letter 91-05, "Licensee Commercial-Grade 

Procurement and Dedication Programs," on April 9, 1991. 

(i) 

(4) 

Neither the 01 investigation nor the NRC inspection of UC documented in 

Report 99900362/90-01 identified any substandard fasteners. 

Compared to the safety significance of the violation cited in the Notice 

of Wiolation, the other violations had relatively lower safety 

significance, and we believe, little purpose would be served in this 

case by citing the less significant violations. 

During this inspection, the team also found that other elements of your 

quality program did not meet NRC requirements and resulted in the 

identification of several nonconformances described in Report 99900362/90-01. 

The inspection identified that, contrary to the requirements of 10 CFR 

Part 50, Appendix 6, UC had not established adequate measures to control the 

dedication of commercial-grade material. As a result, commercial-grade 

products were supplied to the nuclear industry as safety-re1 ated basic 

components and were certified by UC as complying with 10 CFR Part 50, 

Appendix B and 10 CFR Part 21. The specific findings and references to the 

pertinent requirements for all nonconformances are described in the enclosed 

inspection report. 

A notice of nonconformance for the findings is not being issued for the same 

reasons described above for issuing only Violation 90-01-01. 

the staff recognizes that at the time of this inspection neither NRC licensees 

nor their suppliers had a clear understanding of commercial-grade dedication. 

However, if UC has resumed business with the nuclear industry pursuant to 

10 CFR Part 21, the staff expects UC to carefully review the information in 

the enclosed inspection report and take actions, as appropriate, to correct 

the nonconformances identified and to prevent recurrence and, as part of the 

response required below, provide a summary of your corrective actions taken. 

Additionally, 

You are required to respond to this Notice of Violation within 30 days of the 

date of this letter by filing a written response under oath and affirmation. 

You should follow the instructions specified in the enclosed Notice of 

84 
K. Weger -5- 

Violation when preparing your response. In addition to the instructions 

specified in the enclosed Notice of Violation, send copies of your response, 

at the same address, to the Director, Office of Enforcent, and to the 

Assistant General Counsel for Hearings and Enforcewnt. 

After reviewing your responses to the hand for Information and Notice of 

Violation, including your proposed corrective actions, the staff will 

determine whether further action is necessary to ensure compliance with NRC 

regul atory requi rements. 

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

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

The responses directed by this letter and its enclosures are not subject to 

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

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

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

discuss them with you. 

Sincerely, 

R. Lee S&ssard, Director 

Division of Technical Support 

Off i ce of Nucl ear Reactor Regul at i on 

Docket No. 99900362 

EA-91-020 

Encl osures : 

1. Notice of Violation 

2. Report No. 99900362/90-01 

3. Synopsis: Report of Investigation Case No. 4-89-016 

85 
NOTICE OF WIOLATION 

Unistrut Corporation 

Wayne, Michigan 

Docket No. 99900362 

Report No. 90-01 

EA-9 1 -0 2 0 

During an NRC inspection conducted on December 10-14, 1990, violations of NRC 

requirements were identified. In accordance with the "General Statements of 

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

(1990), the most significant violation is listed below: 

Section 21.21, "Notification of failure to comply or existence of a defect," 

of 10 CFR Part 21 (in effect at the time of this inspection), Subsection 

(a)(l) required, in part, that each individual, corporation or other entity 

subject to the regulations adopt appropriate procedures for either evaluating 

deviations or informing the 1 icensee or purchaser of the deviation. 

Contrary to the above, two examples were identified in which Unistrut 

Corporation (UC), an entity subject to 10 CFR Part 21, failed to perform an 

evaluation of a deviation (i.e., a departure from the technical requirements 

included in a procurement document), or inform licensees or purchasers so they 

could cause an evaluation to be performed. 

Specifically, in the following instances, UC procured commercial-grade 

fasteners from the General Fastener Company (GF) of Livonia, Michigan and, 

without dedicating the commercial-grade fasteners for use as basic components, 

supplied them to the nuclear industry as safety-related basic components, 

issuing certificates of conformance to NRC 1 icensees which certified that the 

fasteners complied with NRC regulatory requirements. 

(90-01-01) 

For Gulf States Utilities Company, River Bend Station purchase order 

(PO) 89-4-77172, dated August 25, 1989, issued to UC for safety-related 

ASTM A307, Grade A, galvanized, hex head bolts, 3/8-inch-16 x 2-1/2 inch 

that imposed on UC the requirements of 10 CFR Part 50, Appendix B and 10 

CFR Part 21; UC supplied River Bend Station with commercial-grade hex 

head bolts (since UC did not dedicate the fasteners) and a certificate 

of conformance dated October 5, 1989, which certified that the fasteners 

met the requirements of the licensee PO, 10 CFR Part 50, Appendix B and 

10 CFR Part 21. 

For Consumers Power Company, Pal i sades P1 ant, Bechtel Constructors 

Corporation issued PO 20557-F1020-Q to UC for safety-related ASTM A307, 

grade A, hex head cap screws, 3/8-inch-16 x 3/4-inch that imposed on UC 

the requirements of 10 CFR Part 50, Appendix B and 10 CFR Part 21; UC 

suppl ied Bechtel Constructors Corporation--Pal i sades P1 ant with 

comnercial-grade hex head cap screws (since UC did not dedicate the 

fasteners) and a certificate of conformance dated October 4, 1990, which 

certified that the fasteners met the requirements of the licensee PO, 10 

CFR Part 50, Appendix B and 10 CFR Part 21. 

-1- Enclosure 1 

86 
These instances have been classified as a Severity Level I11 Violation 

(Supplement VII). 

Pursuant to the provisions of 10 CFR 2.201, UC is hereby required to submit a 

written statement or explanation to the U.S. Nuclear Regulatory Commission, 

ATTN: 

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. 

a "Reply to Notice of Violation" and should include for the violation (a) the 

reason for the violation, or, if contested, the basis for disputing the 

violation; (b) the corrective steps that have been taken and the results 

achieved; (c) the corrective steps that will be taken to avoid further 

violations; and (d) the date when full compliance will be achieved. Your 

response may reference or include previous docketed correspondence if the 

correspondence adequately addresses the required response. Where good cause 

is shown, consideration will be given to extending the response time. Under 

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

be submitted under oath or affirmation. 

Document Control Desk, Washington, D.C. 20555-0001 with a copy to the 

This reply should be clearly marked as 

Dated at Rockvi 11 e , Mary1 and 

this 16th day of June, 1995 

-2- 

87 
ORGANIZATION: UNISTRUT CORPORATION 

WAYNE , MICHIGAN 

I 

REPORT INSPECTION INSPECT1 ON 

NO.: 99900362/90-01 DATE: DECEMBER 10-14, 1990 ON-SITE HOURS: 93 

CORRESPONDENCE ADDRESS: Unistrut Corporation 

35660 Clinton St. 

Wayne, Michigan 48184 

ORGANIZATIONAL CONTACT: 

NUCLEAR INDUSTRY ACTIVITY: Unistrut Corporation (UC) is a manufacturer/suppl ier 

of continuous slotted metal channel products, general fittings, fasteners, and 

mechanical accessories used to create structural framing systems. These 

structural framing systems are supplied as safety-related components and are 

used to support safety-related pipes, duct work, cable trays, and conduit. 

Approximately 10 percent of UC's business is with the nuclear industry. 

Kris Weger, Vice President and General Manager 

TELEPHONE NUMBER: (313) 721-4040 

ate 

ASSIGNED INSPECTOR: 

Steven 14. Matthews. Reactive Inspection Section 

No. 1 (RIS-1) Vend& Inspection Branch (VIB) 

OTHER INSPECTOR( S) : Ann M. Fi tzgerald, NRC I Stewart, L-Magruder , NRC 

2-Pi- 91 

Date 

APPROVED BY: 

Uldis Potapovs, Chief, RIF-1, VIB 

CllUN BAStS AND SCOPE: 

A. 

B. 

BASES : -~ 10 CFR Part 50, Appendix B and 10 CFR Part 21 

SCOPE: 

selected areas such as: 

control of purchased material, equipment and services. 

To assess the UC quality program dnd its implementation in 

control of procurement; design control; and 

PLANT SITE APPLICABILITY: All nuclear power plants with UC products. 

88 Enclosure 2 
ORGAN I ZAT ION : UNISTRUT CORPORATION 

WAYNE, MICHIGAN 

REPORT INSPECTION 

NO. : 999000362/90-01 RESULTS : PAGE 2 of 27 

A. VIOLATIONS: 

1. Contrary to Section 21.21, "Notification of Failure to Coihply or 

Existence of a Defect," of 10 CFR, two examples were identified where 

UC failed to perform an evaluation or to notify the licensees or 

purchasers so they could perform or cause an evaluation to be 

performed regarding commercial-grade fasteners supplied to the nuclear 

industry as safety-related with Certificates of Conformance (C of C) 

that certify compliance with regulatory requirements. (90-01-01) 

a. UC supplied commercial-grade fasteners to the River Bend Nuclear 

Station with a C of C which certified compliance with the licensee's 

PO, Appendix B to 10 CFR Part 50, and 10 CFR Part 21. 

b. UC supplied commercial-grade fasteners to the Palisades Nuclear 

Plant with C of C which certified compliance with the licensee's 

PO, Appendix B to 10 CFR Part 50, and 10 CFR Part 21. 

Contrary to Section 21.31, "Procurement Documents," of 10 CFR, a 

review of documentation packages revealed two examples where UC issued 

purchase orders (POs) to suppliers for safety-related components 

arid failed to invoke the requirements of 10 CFR Part 21 on the POs. 

2. 

(90-01-02) 

a. UC failed to invoke the requirements of 10 CFR Part 21 on PO 

No. P-03779, October 4, 1990, issued for safety-related 

fasteners. 

b. UC failed to invoke the requirements of 10 CFR Part 21 on PO 

No. P-03791, October 8, 1990, issued for safety-related 

fasteners. 

3. Contrary to Section 21.21, "Notification of Failure to Comply or 

Existence of a Defect," of 10 CFR, UC procedure in Section F, "10 CFR 

Part 21 Reporting of Defects and Non-Compliance," of the Unistrut 

Diversified Products Quality Assurance Manual (QAM), Revision No. 3, 

September 4, 1990, did not (1) define the term "potential reportable 

events;" therefore, UC staff did not know what conditions to report to 

their supervisors; (2) ensure the evaluation of deviations or 

departures from the technical requirements included in procurement 

documents; and (3) describe the process or bases of performing 

evaluations for reporting defects to the Nuclear Regulatory Con;rr,ission 

(NRC) or notifying customers of deviations. (90-01-03) 

89 
ORGANIZATION: UNISTRUT CORPORATION 

WAYNE, MICHIGAN 

PAGE 3 of 27 

REPORT 

NO.: 999000362/90-G1 

B . NONCONFORMANCES - : 

1. 

2. 

3. 

Contrary to Criterion 11, "Quality Assurance Program," of Appendix B 

to 10 CFR Part 50, and Section 11, "Quality Assurance Program," of the 

QAM, the NRC inspectors reviewed certain sections of the UC QAM and 

determined that UC failed to establish and implement a program, for 

the quality criteria reviewed, that included those planned and 

systematic actions necessary to provide adequate confidence that basic 

components such as structural framing channel, fasteners, general 

fittings, and rrechanical accessories wi 11 perform their safety-related 

function. Specifically, UC failed to prescribe the following: 

(1) the controls necessary for the completion of internal audit 

corrective actions to assure the quality of inipacted products; (2) the 

quality requirements to be imposed in POs to vendors of safety-related 

products; (3) the controls necessary to assure that UC supplied 

materials met licensee-imposed material specifications; and (4) the 

criteria used by UC to review the licensees' POs to establish 

acceptance of the order and to ensure UC's compliance with the quality 

and technical requirements of the 1 icensees' POs. (90-01-04) 

Contrary to Criterion 11, "Quality Assurance Program," of Appendix B 

to 10 CFR Part 50, UC failed to establish and implement a program, 

before November 1990 to provide for the indoctrination and training 

of personnel performing activit'es affecting quality in the requirements of 10 CFR Part 21 and Appendix B to 10 CFR Part 50. This is 

another occurrence of Nonconformances 86-01-08 and 86-01-09, identified 

by the NRC during the previous inspection and documented in Inspection 

Report ( IR) No. 99900362/86-01. (90-01-05) 

Contrary to Criterion 111, "Design Control," of Appendix I3 to 10 CFR 

Part 50, UC failed in Section 111, "Design Control," of the UC QAM, 

and Quality Assurance Procedure (QAP) No. 20-01-038, "Commercial-grade 

Dedication Program," Revision No. 0, September 4, 1990, to establish 

adequate measures for the dedication of commercial-grade materials 

that were supplied to the nuclear industry as safety-related. 

Specifically, UC (1) had not established an adequate basis for 

accepting the material certifications from its suppliers; (2) did not 

verify the chemical and physical properties of coil material by 

performing chemical and physical tests; (3) until recently, UC did not 

maintain traceability on material in the press shop; (4) did not 

perform testing traceable to each coil of material used in the 

fabrication of channel products; and (5) except for channel nuts and 

90 
ORGANIZATION: 

REPORT 

NO.: 999000362/90-01 i 

UNISTRUT CORPORATION 

WAYNE, MICHIGAN 

- 

INSPECTION 

RESULTS : PAGE 4 of 27 

90-degree angle fittings, did not perform any testing on products 

produced in the press shop. Therefore, UC supplied inadequately 

dedicated parts such as structural framing channel, channel nuts, 

screws, nuts, washers, general fittings, and mechanical accessories as 

safety-related components to the nuclear industry. (90-01-06) 

4. 

5. 

Contrary to Criterion 111, "Design Control ,I' dnd Criterion IV, 

"Procurement DGcument Control," of Appendix B to 10 CFR Part 50, UC's 

requirements in Section 111, "Design Control ,I' and Section IV, 

"Procurement Document Control," of the UC QAM and QAP No. 20-01-010, 

"Control of Purchased Materials, Products, and Services," Revision 

No. 4, November 9, 1990, failed to establish measures to ensure that 

regulatory requirements imposed in licensees' POs are established and 

translated into documents that ensure adequate quality is achieved. 

Moreover, UC failed to ensure that compliance to imposed regulatory 

requirements had been achiaed before issuing C of C that certified 

compliance. This condition led UC to make unsubstantiated certification stateriiepts titi C; c;t C issued to licensees. Specifically, UC 

routinely failed to perform the following: (1) control the receipt 

and review of POs from licensees in order to establish the requirements imposed on UC; (2) establish measures necessary to trdnslate the 

requirements imposed in POs from licensees into specifications or 

procedures; (3) assure compliance to the requirements imposed in POs 

from licensees prior to issuing C of C that certified compliance; (4) 

verify UC's compliance prior to revising previously issued C of C; and 

(5) impose 10 CFR Part 50, Appendix B or similar quality requirements 

in POs to suppliers of safety-related fasteners. 

Contrary to Criterion VII, "Control of Purchased Material, Equipment 

and Services," of Appendix B to 10 CFR Part 50, UC's requirements iri 

Section VII, "Control of Purchased Materials and Services," of the UC 

QAM and QAP No. 20-01-007, "Vendor Qualifications and Audit Requirements," Revision No. 3, August 1, 1990, failed to establish adequate 

measures to assess and document the effectiveness of the quality 

programs of vendors supplying materials and services. Specifically, 

the UC audits failed to discover that the quality program of their 

fastener supplier did not meet 10 CFR Part 50, Appendix B requirements. In addition, UC did not audit their commercial-grade suppliers 

of steel coil material, which furnished material certifications used 

by UC in the commercial-grade dedication process. This is another 

occurrence of Nonconfornlance 86-01-05, identified by the NRC during 

the previous irlspeccicn cind documented in IR No. 99900362/86-01. 

(90-01-07) 

(90-01-08) 

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C. UNRESOLVED ITEMS: 

D. STATUS OF PREVIOUS INSPECTION FINDINGS: 

1. (CLOSED) Violation 86-01-01 

Contrary to 10 CFR 21.6, UC did not post 10 CFR Part 21 procedures for the 

evaluation of deviations, Section 206 of the Energy Reorganization Act of 

1974, and the regulations of 10 CFR Part 21. 

The NRC inspectors determined that UC had properly posted the required 

documents and that corrective actions taken to prevent recurrence were 

I 2. (CLOSED) Violation 86-01-02 

Contrary to 10 CFR 21.21, UC's procedure, "Processing of Reportable 

Events," did not specify or identify a time limit for notification or 

method of notifying the NRC of a defect. 

The NRC inspectors determined that UC's corrective actions taken to prevent 

recurrence were inadequate. Violation 90-01-03, identifies additional 

examples where UC failed to establish adequate procedures for the 

evaluation of deviations. 

tracked under Violation 90-01-03. 

This previous inspection finding will now be 

3. (CLOSED) Violation 86-01-03 I 

Contrary to 10 CFR 21.31, UC failed to impose the requirements of 10 CFR 

Part 21 in POs to suppliers of safety-related materials and services. 

The NRC inspectors determined that UC's corrective actions taken to prevent 

recurrence were inadequate. Violation 90-01-02, identifies additional 

examples where UC issued safety-related POs to niaterial suppliers and 

failed to imposed the requirements of 10 CFR Part 21. 

inspection finding will now be tracked under Violation 90-01-02. 

This previous 

I 4. (CLOSED) Nonconf ormance 86-01-04 

Contrary to Criterion V of 10 CFR Part 50, Appendix B and QAP No. 20-01-003, 

"Receiving Inspection of Customer Returned Material ,'I of the QAPM, UC 

failed to prcrperly secjregate rejected material. 

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The NRC inspectors reviewed the control of rejected material identification 

and segregation during a plant tour and determined that the corrective 

actions taken to prevent recurrence were adequate. 

5. (OPEN) Nonconformance 86-01-05 

Contrary to Criterion V of 10 CFR Part 50, Appendix B and Section XIX, 

"Audits," of the QAM, UC failed to perform external audits of certain 

vendors. 

The NRC inspectors determined that external vendor audits are currently 

performed at scheduled intervals. 

regarding the adequacy and completeness of vendor audits, which are 

discussed in Section E.8.0 of this report and identified as Nonconformance 

However, other concerns were raised 

9 0 -0 1-08. 

6. (CLOSED) Nonconformance 86-01-06 

Contrary to Criterion V of 10 CFR Part 50, Appendix B and Section XIX, 

"Audits," of the QAM, UC failed to perform required interrial audits 

according to schedule. 

The NRC inspectors determined that internal audits are currently performed 

dt scheduled intervals and that the corrective actions taken to prevent 

recurrence were adequate. 

7. (CLOSED) Nonconformance 86-01-07 

Contrary to Criterion V of 10 CFR Part 50, Appendix B and Section XII, 

"Test Control," of the QAM, and QAP No. 20-01-031, "Qualification of NDE 

(PT) Personnel," of the QAPM, UC failed to document objective evidence of 

the annual eye examination for an FIDE Level I1 and a Level 111 examiner. 

The NRC inspectors found that QAP No. 20-01-031, as described above, 

had been deleted from the UC QAPM and Section IX, "Control of Special 

Processes," of the QAM did not address nondestructive examinations. 

UC decided that the requirements for NDE examinations were no longer 

necessary and the quality program had been revised accordingly. 

inspectors determined that the corrective actions taken were adequate. 

The NRC 

8. (OPEN) Nonconformance 86-01-08 

Contrary to Criterion V of 10 CFR Part 50, Appendix B and QAP No. 20-01-024, 

"Quality Personnel Selection and Training," of the QAPM, UC failed to 

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document objective evidence of the indoctrination and traini,ng of a UC 

inspector or maintain complete qualification files for other UC inspectors. 

The NRC inspectors determined that the corrective actions taken to prevent 

recurrence regarding indoctrination ard ti-airling of personnel was 

inadequate, as evidenced by Nonconformance 90-01-05, which identifies other 

inadequacies regarding indoctrination and training described in 

Section E.5.0 of this report. 

9. ( OPEN) Nonconformance 86 -01-09 

Contrary to Criterion V of 10 CFR Part 50, Appendix B and QAP No. 20-01-030, 

"Auditor Qualifications," of the QAPM, UC failed to document objective 

evidence of communicatioi: skills, attendarice at an audit training course, 

and fulfiiling cudit pdrticipation requirements for several lead auditors. 

The NRC itispectors determined that Corrective actions taken to prevent 

recurrence regarding indoctrillation dnd training of lead auditors was 

inadequate, as evidenced by Nonconformance 90-01-05, which identifies other 

inadequacies regardiy indoctrination and training described in Section 

E.5.0 of this report. 

E. INSPECTION FINDINGS AND OTHER COMMENTS: 

1.0 Entrance and Exit Meetings 

The Nuclear Regulatory Commission (NRC) inspectors informed UC staff of the 

scope of the inspection, outlined areas of concern, and established working 

interfaces during the entrance meeting on December 10, 1990. 

1990, the NRC inspectors summarized the inspection findings, observations, 

and concerns to UC management during the exit meeting. 

On December 14, 

2.0 Background 

UC manufactures safety-related continuous slotted metal channel products 

used to create many different structural framing systems to support 

safety-related items. 

nationwide Authorized Service Centers, which are either owned and operated 

by UC or independently owned. The framing systems supplied to the nuclear 

industry consist generally of 1-5/& inch width channel , channel nuts, 

fasteners, fittings, and pipe clamps. 

related components such as pipes, duct work, cable trays, and conduit. 

These systems are supplied through a network of 

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UC has been supplying structural framing products to the nuclear industry 

since 1975, however, UC staff claimed that they have not always certified 

their products as safety-related. 

to 10 CFR Part 50, Appendix B and 10 CFR Part 21 began was of interest to 

the NRC inspectors. review, it had determined that UC first began issuing C of Cs to the 

requirements of 10 CFR Part 21 and Appendix B to 10 CFR Part 50 in August 

1988. The NRC inspectors, however, identified safety-related POs in July 

and September of 1982 with UC C of Cs that certified compliance with the 

customer's PO, which imposed 10 CFR Part 21 and 10 CFR Part 50, Appendix B. 

This issue is discussed in more detail in Section E.7.0 of this report. 

Determining the initial date certification 

UC staff told the NRC inspectors that based on UC's 

3.0 Inspection of 10 CFR Part 21 Compliance 

During an interview with UC's Vice President and General Manager, 

Production Engineering Manager, and Quality Assurance Manager conducted by 

NRC staff or1 June 13, 1990, UC was made aware, by NRC staff, that it had 

supplied safety-related fasteners to the nuclear industry with invalid 

C of C's certifying that the safety-related fasteners supplied met the 

requirements of 10 CFR Part 21 and Appendix B to 10 CFR Part 50. As a 

result, on June 21, 1990, UC advised all UC Authorized Service Centers that 

on June 15, 1990, UC had suspended certificaxion of products as complying 

with Apperidix B to 10 CFR Part 50 and 10 CFR Part 21. After making changes 

to the UC QAM and QAPM, UC advised all nuclear customers on September 26, 

1990, that UC's 10 CFR Part 50, Appendix B program for safety-related metal 

framing products was, "back on line." 

UC held several meetings with their only fastener supplier, General 

Fastener Co. (GF) of Livonia, Michigan, in October 1990 to discuss GF's 

ability to supply safety-related fasteners that fully comply with NRC 

requirements. On October 26, 1990, GF decided that it could not supply 

safety-related fasteners because it did not have, nor had it ever had, a 

quality program that complied with 10 CFR Part 50, Appendix B or 10 CFR 

Part 21. UC subsequently decided it could not continue to supply safetyrelated fasteners to the nuclear industry in compliance with NRC requirements. 

decision or rescinded the letter of September 26, 1990. 

supplied fasteners, certified as safety-related, to the nuclear industry 

since October 26, 1990. 

The NRC inspectors reviewed UC's implementation of and compliance with 

10 CFR Part 21 requirements for licensees' procurement of basic components. 

During this review three violations of NRC requirements were identified. 

The violations are discussed separately as follows. 

As of December 14, 1990, UC had not notified licensees of this 

UC has not 

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3.1 Evaluation and Notification of Deviations 

The following two examples were identified where UC was aware of a 

departure from the technical requirements of 1 icensees' safety-related POs 

and failed to either evaluate the deviation or inform the licensee or 

purchaser of the deviation so they could perform or cause an evaluation to 

be performed. 

area of inspection. 

a. Gulf States Utilities Co. issued PO No. 89-4-77172, August 25, 1989, 

for safety-related ASTM A307, grade A, galvanized, hex head bolts, 3/8 

inch - 16 x 2-1/2 inch, with special head mdrking symbol "U", to 

supply the River Bend Nuclear Station and imposed the requirements of 

Appendix B to 10 CFR Part 50 and 10 CFR Part 21. 

bolts from GF with PO No. P-02694, September 18, 1989, as 

commercial-grade. UC's PO No. P-02694 to GF changed the licensee's 

procurement material specifications and ordered SAE-J429, grade 2, hex 

head cap screws with the special head marking. 

As a result, Violation (90-01-01) was identified in this 

UC purchased the 

GF has been UC's only supplier of fasteners and UC staff told the NRC 

inspectors that the requirements of Appendix B to 10 CFR Part 50 and 

10 CFR Part 21 were not imposed it1 tttt PO tu GF because UC had 

developed an understanding with GF that all fasteners supplied by GF 

were nuclear safety-related. 

understandings between UC and GF are discussed in Section E.8.0 of 

this report. 

The relationship and bases for the 

The NRC inspectors determined, through interviews with GF staff , that 

with regard to supplying fasteriers to UC, GF did not, as of December 12, 

1990, nor has it ever had a quality program that it believed complied 

with the NRC requirements in 10 CFR Part 50, Appendix B or 10 CFR Pdrt 21. 

GF issued a C of C of September 20, 1989, to UC certifying that 

the products met the requirements specified in PG No. P-02694. 

not dedicate the bolts, that are in fact coinniercial-grade, prior to 

certifying them as meeting the requirements of Gulf States Utilities 

Co.'s PO, 10 CFR Part 21, and Appeiidix B to 10 CFR Part 50, as stated 

on the UC C of C of October 5, 1989. UC certified that the fasteners 

met the licensee's required specification, ASTM A307, grade A, even 

though UC ordered them to SAE-J429, grade 2, and received the 

fasteners from GF with a C of C that they met SAE-3429, grade 2. 

had never performed a commercial-grade dedi cation of the fasteners 

purchased from GF. 

UC did 

UC 

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The NRC inspectors requested UC's evaluation of this deviation to 

determine (1) when GF began supplying commercial-grade fasteners that 

UC believed to have been supplied as safety-related; (2) which 

licensees' POs are impacted by this condition; and (3) which licensees 

or purchasers must be notified of the deviation. 

1990, UC had not performed an evaluation of the deviation, which it 

was made aware of on June 13, 1990, or informed licensees or 

purchasers of the deviation so they could perform or cause an 

evaluation to be performed. 

I 

As of December 14, 

b. Bechtel Constructors Corp. issued PO No. 20557-F1020-Q to UC for 

safety-related ASTM A307, grade A, hex head bolts, 3/8- 16 x 3/4 inch, 

and 3/8 inch flat washers. The PO was issued to supply the Palisades 

Nuclear Plant and imposed the requirements of 10 CFR Part 21 and 10 CFR 

Part 50, Appendix B. 

P-03779, October 4, 1990. 

UC purchased the items from GF with PO No. 

UC ordered the items from GF using an "N" prefix to their standard 

catalog number which, UC claims, designates the items as nuclear 

safety-related. 

the new 'IN" prefix, introduced by UC in September 1990, indicated that 

the fasteners were designated as nuclear safety-related. 

No. P-03779 to GF did not impose the requirements of 10 CFR Part 21, 

Appendix B to 10 CFR Part 50, or any other quality requirements. 

UC's PO No. P-03779 to GF changed the licensee's procurement material 

specifications and ordered SAE-J429, grade 2, hex head cap screws. 

GF issued C of C of October 9, 1990, to UC certifying that the fasteners 

for PO No. P-03779 complied with UC's PO, the mater!;' :pecification 

SAE-J429, grade 2, dnd a quality program which conforms to ANSI N45.2 

and 10 CFR Part 50, Appendix B and that 10 CFR Part 21 applied to this 

order. UC then issued C of C to Bechtel Constructors Corp. on October 10, 

1990, certifying that the fasteners met ASTM A307, grade A, and 

were processed in accordance with the purchaser's PO, ANSI N45.2, 

Appendix B to 10 CFR Part 50, and that 10 CFR Part 21 applied to this 

order. 

UC staff told the NRC inspectors that GF knew that 

UC's PO 

GF subsequently advised UC on October 26, 1990, that in fact, GF's 

C of C of October 9, 1990, were riot valid because the fasteners were 

not processed in accordance with Appendix B to 10 CFR Part 50 and 

that 10 CFR Part 21 requirements were not applied to this order. 

Since UC did not perform a commercial-grade dedication of these items, 

the UC C of C of October 10, 1990, issued to Bechtel Constructors 

Corp. were not valid. 

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The NRC inspectors requested UC's evaluation of the deviation or 

evidence that Bechtel Constructors Corp. had been notified of the 

deviation. As of December 14, 1990, UC had not performed an 

evaluation of the deviation, which it was made aware of on October 26, 

1990, or informed the purchaser of the deviation so they could 

perform or cause an evaluation to be performed. 

3.2 Procurement Documents Incorporation of 10 CFR Part 21 

The inspectors reviewed UC's implementation of Section 21.31, "Procurement 

Documents," of 10 CFR and found two examples where UC POs to a supplier of 

safety-related components did not specify that 10 CFR Part 21 would apply. 

As a result, Violation (90-01-02) was identified in this area of 

inspection. 

a. 

b. 

3.3 

On October 4, 1990, UC issued PO No. P-03779 to GF for part No. 

NHCS037075, 3/8 inch - 16 x 3/4 inch, hex head cap screws, per ANSI 

B18.2.1, SAE-J429, grdde 2, and part No. NFLW037 EL, 3/8 inch flat 

washers, type A plain (wide) per ASTM F944-83, and ANSI B18.22.1-81. 

The new "N" prefix to the UC part No. designates the parts as nuclear 

safety-related items. These fasteners were supplied to the Palisades 

Nuclear Plant. UC did not invoke the requirements of 10 CFR Part 21 

and 10 CFR Part 50, Appendix B in the safety-related order to GF. 

On October 8, 1990, UC issued PO No. P-03791 to GF for part 

No. NHCS050125, 1/2 inch - 13 x 1-1/4 inch, hex head cap screws, per 

ANSI B18.2.1, SAE-J429, grdde 2. 

designates the parts as nuclear safety-related items. 

invoke the requirements of 10 CFR Part 21 and 10 CFR Part 50, 

Appendix B in the safety-related order to GF. 

The 'IN" prefix to the UC part No. 

UC did not 

10 CFR Part 21 Procedures Implementation 

The inspectors reviewed Section F of the 'Unistrut Diversified Products 

Quality Assurance Manual (QAM) Revision No. 3, September 4, 1990. 

"10 CFR Part 21 Reporting of Defects and Non-Compliance," of the QAM is 

the UC procedure that implements the requirements of Section 206 of the 

Energy Reorganization Act of 1974 and Section 21.21, "Notification of 

Failure to Comply or Existence of a Defect," of 10 CFR. 

concluded that Section F of the UC QAM failed to comply with the 

requirements of Section 21.21 of 10 CFR as described below. As a result, 

Violation (90-01-03) was identified in this area of the inspection. 

Section F, 

This review 

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a. Section F requires that all employees should notify their immediate 

supervisor of, "potential reportable events." However, Section F did 

not define what constitutes a "potential reportable event'' nor did UC 

provide training for employees that defined the term. 

employees did not know what conditions to report to their immediate 

supervisors. 

Section F failed to ensure the evaluation of deviations or departures 

from the technical requirements included in procurement documents. 

Therefore, 

b. 

c. Section F failed to prescribe the process or bases of performing 

evaluations for reporting defects to the NRC or notifying custmers of 

deviations. 

4.0 Inspection of 10 CFR Part 50, Appendix B, Compliance 

This area of the inspection focused on UC's QAM and QAPM. 

the QAM and QAPM to meet the quality assurance criteria requirements of 

10 CFR Part 50, Appendix B for safety-related components such as structural 

framing channel, channel nuts, screws, nuts, washers, general fittings, and 

mechanical accessories supplied to the nucledr industry. Implementation of 

the UC QAM is supplemented by the QAPM. The NRC inspectors reviewed 

only certain sections of the QAM and QAPM. 

determine whether or not UC had established and implemented a program, for 

the quality criteria reviewed, that comprises all those planned and 

systematic actions necessary to provide adequate confidence that components 

supplied to the nuclear industry will perform satisfactorily in service. 

UC established 

The review was conducted to 

The QAM sections reviewed prescribe only generalized responsibilities and 

purposes and did not prescribe details of a quality program that adequately 

established measures that comply with NRC requirements. The NRC inspectors 

determined from this review that the sections of the QAM, as supplemented 

by the QAPM, did not adequately establish those planned and systematic 

actions necessary to assure the safety and reliability of the components 

and identified the following examples. 

a. The QAM requires in Section 11, "Quality Assurance Program," paragraph 2.7, 

that the status and adequacy of this program will be determined 

through the review of internal audits, and Section XVIII, "Internal/ 

External Audits," paragraph 18.4 provides for the completion of 

corrective action requested within a reasonable time, and QAP 

No. 20-01-027, "Internal Audits," Revision No. 1, July 6, 1987, also 

provides for the completion of corrective action requested within a 

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reasonable time. The QAM and its supplements did not adequately 

prescribe controls for the completion of internal audit corrective 

actions that are necessary to assure that the adequacy of the quality 

of safety-related products, impacted by the corrective actions, is 

achieved. 

b. 

C. 

The QAM requires in Section IVY "Procurement Document Control ,'I 

paragraphs 4.2 and 4.4, that when safety-related products are 

specified (on licensee's POs), this information shall also be called 

out on the PO to the supplier. However, the QAM did not prescribe the 

quality and technical requirements necessary for safety-related POs 

issued to vendors. The QAM referenced QAP No. 20-01-010. The NRC 

inspectors concluded that QAP No. 20-01-010, "Control of Purchased 

Materials, Products, and Services," Revision No. 4, November 9, 1990, 

did not prescribe the quality and technical requirements to be imposed 

on POs to suppliers of safety-related products, inclusive of imposing 

the requirements of 10 CFR Pdrt 21, and did not ensure that the 

rliaterial specifications imposed on suppliers are the same or 

equivtilent to material specifications required by the licensees' POs. 

The QAM requires in Section IW, "Procurement Document Control," 

paragraphs 4.7 and 4.8, that customers' POs requiring certifications 

will not be accepted by Customer Service without written approval from 

the Quality Assurance Department. 

the controls for nuclear safety-related POs. The QAM referenced QAP 

No. 20-01-034. The NRC inspectors concluded that QAP No. 20-01-034, 

"Customer Purchase Order Review," Revision No. 1, October 22, 1990, 

did not include the PO review criteria to meet regulatory 

requirexents, inclusive of verifying whether or riot 10 CFR Part 21 was 

imposed by the licensees' POs. 

However, the QAM did not prescribe 

The inspectors concluded from this review that the QAM, supplemented by the 

QAPM, did not prescribe a quality program in certain sections that provides 

adequate control over those activities affecting the quality of the 

components to the extent consistent with their importance to safety as 

required by regulatory requirements imposed on UC by licensees' POs. 

Specifically, UC failed to prescribe the following: 

necessary for the completion of internal audit corrective actions to assure 

the quality of impacted products; (2) the quality requirements to be 

imposed in POs to vendors of safety-related products; (3) the controls 

necessary to ensure that UC sup lied materials met licensee-imposed 

material specifications; and (4 ! the criteria used by UC to review the 

licensees' POs to establish acceptance-of the order and to ensure UC's 

(1) the controls 

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compliance with the quality and technical requirements of the licensees' 

POs. 

inspection. 

As a result, Nonconformance (90-01-04) was identified in this area of 

5.0 Indoctrination and Trdining 

The NRC inspectors reviewed the training records for UC Quality Control 

Irispectors. 

10 CFR Part 50, Appendix B, for UC personnel whose activities affect 

quality, was held on November 8, 1590. 

that included the requirements of 10 CFR Part 21 and 10 CFR Part 5C, 

Appendix B before November 1990. 

The first training session that included 10 CFR Part 21 and 

There were no reccjrds of training 

Compliance with 10 CFR Part 21 and Appendix B to 10 CFR Part 50 was 

required by the procurement documents of licensees. Training in lG CFE 

Part 21 responsibilities is necessary to ensure that deviations are 

identified and evaluated. Unistrut personnel were never trained, before 

November 1990, in the definition of a deviation as described in 10 CFR 

Part 21 or in the quality requirements of 10 CFR Part 50, Appendix B. 

Therefore, Unistrut personnel who performed activities affecting quality 

did not have the training necessary to identify a deviation as a departure 

from the technical requirements of the procurement documents. 

UC failed to establish and implenitnt a program, before November 1990, to 

provide for the indoctrination and training of personnel performing 

activities affecting quality in the requirements of 10 CFR Part 21 and 

Appendix B to 10 CFR Part 50. This is another occurrence of Nonconformances 86-01-08 and 86-01-09, identified by the NRC during the previous 

inspection and documented in IR No. 99900362/86-01. 

Nonconformance (90-01-05) was identified in this area of the inspection. 

As a result, , 

6.0 Dedication of Commercial-Grade Items 

The dedication process, consisting of the measures UC uses to select and 

review for suitability of application those materials, parts, equipment and 

processes that are essential to the safety-related functions of their 

products, as required by Criterion 111, "Design Control," of Appendix B to 

10 CFR Part 50, was reviewed. 

dedication process which encompasses audits of suppliers, receipt 

inspection, in-process inspection, traceability control, and testing. 

this end, the inspectors were given a detailed plant tour during which 

they had the opportunity to observe in-house processes and talk with 

several inspectors. 

The inspectors evaluated the entire 

To 

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6.1 Dedication of Products Manufactured by UC 

The manufacturing operations at UC are divided into the processing 

department, where channel and tubing are made, and the press shop, where 

channel nuts, fittings, and pipe clamps are made. 

rndterials for these products are purchased commercial-grade, UC has assumed 

the responsibility for dedicating all of the coil matericls that are sold 

as safety-related products. 

Since all the coil 

The following QAPs were reviewed by the inspectors during their evaluation 

of the UC process: 

QAP NO. TITLE REV DATE 

20-01 -002 

20-01-006 

20-01 -GO8 

20-01-009 

20-01-011 

20 -0 1-0 14 

20-01 -020 

20-01-029 

50-01-037 

20-01-038 

20 -0 1-041 

20-01 -042 

20-01-043 

20-01-044 

20-01-045 

First Piece Inspection 3 

Receiving Inspection 5 

Traceability of Channel 4 

and Tubing 

In-Process Inspection 2 

Heat Treat Preparation 2 

and Inspection 

Pull Test 3 

- Charinel Products 

Verification of Vendor 2 

Integrity - Incoming 

Materials, Products, etc. 

Channel Series Nuts 

Traceability of Fittings 0 

Dedication Prograni for C 

Commercial-Grade Items 

Ninety Degree Angle 0 

Fittings-Bend Test 

Torque Testing of 1-5/8" 0 

Channel Series Nuts 

Channel Bend Test 0 

Hardness Test for 0 

SAE 5429, grade 2 

Fasteners 

Alloy Verification Test 0 

Slip Tests of 1-5/8" 3 

7/6/90 

7/31/90 

10/22/90 

7/6/87 

7/6/87 

8/10/90 

7/6/87 

8/21/90 

9/4/90 

9/4/90 

8/18/90 

8/ 2 G/ 90 

8/21/90 

8/29/90 

9/4/90 

In addition to these QAPs, Section C, "Critical Characteristics," of the UC 

QAM was also reviewed. 
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The methods UC uses to purchase coil materials and the receipt inspection 

of this material are the same for both the processing department and the 

press shop. 

commercial-grade. A review of the UC audits of these suppliers, however, 

determined that UC did riot adequately establish a basis for acceptance of 

material certifications from them, as described in more detail in Section 

E.8.0 of this report. 

are distributors who do not independently test the steel that they slit and 

ship to UC, thus further reducing the assurance that material traceability 

had been maintained throughout the procurement and slitting process and 

whether or not the material certification correctly represents the coil. 

UC also did not assure the accuracy or tracedbility of the material 

certifications by performing chemical and physical tests on each coil. 

As previously stated, UC purchases all coil material 

In addition, the niajority of their coil suppliers 

The common receipt inspection procedure requires that all materials be 

weighed or counted and that certification documentation from the supplier 

must accompany the material. This documents is checked to ensure that it 

is in accordance with UC's PO requirements. 

the inspector to inspect all coil material and components in accordance 

with instructions, blueprints, POs, dnd/or engineering specifications 

and to record iili inspection activities performed. 

The procedure also requires 

a. The NRC inspectors determined that, in the processing department, 

traceability is adequately maintained on the coil material from the 

time it enters the rrianufacturiixj process until it is shipped out as 

channel. 

pieces of coil slit to the proper width for the manufacturing process 

bound together by heat number (mc;lts). Upqn receipt, each coil is 

assigned a sequential identification number, which is recorded by the 

Quality Assurance Department in a permanent record book. A tag is 

placed on each coil with the following information: 

vendor, purchase order nuniber, quantity received, coil number, heat 

number, and date. Additionally, each mult not having a tag has the 

coil number marked on it with a permanent black marker. When mults 

are roll-formed into channel or tubing, a number is indented on the 

inside of the channels at a spacing not to exceed 22". This number 

consists of the last digit of the calendar year in which the material 

was rolled, the idetitification letter of the operator who roll-formed 

the material, and the identification number dssigned to the coil by 

the Quality Assurance Department. This method allows identification 

of the channel and tubing when it is cut into shorter sections. 

The coil material is received from the vendor in the forrii of 

part number, 

The process also irtvolves a rdndom visual and dimensional inspection 

and destructive testing of a sample piece of formed channel. This 

103 
ORGANIZATION: UNISTRUT CORPORATION 

WAYNE, NICHIGAN 

ffREPORT INSPECTION 

NO.: 999000362/90-01 RESULTS : PAGE 17 of 27 

L 

In further evaluation of the UC dedication process, the inspectors 

reviewed 4 PO packages from licensees for safety-related products I other than fasteners : 

b. 

UC ORDER NO. DATE LICENSEE PRODUCT 

0450-1944 12/01/87 Diablo Canyon c h a ti 11 e 1 nu t s / b rac ke t s 

0450-2028 03/10/88 Diablo Canyon channel/angle fittings 

0130-1917 0 1 / 14/87 L i mer i c k channel 

0 130 -1944 03/04/87 Limerick channel 

These packages included POs from the licensees, C of C from UC; and 

material certifications from the'coil material suppliers. 

inspectors reviewed the tests performed 011 the safety-related products 

that were shipped to licensees to determine whether or not the 

products were adequately dedicated. The channel product that was 

ordered by the utilities listed above was produced from 16 steel 

coils procured from both Rouge Steel Company and Triad Metals 

Incorporated. 

3 coils were traceable to sample pieces of tested channel product. 

All 3 sample pieces had passed the tests described above. None of the 

material from the 16 coils was subjected to chemical or physical tests 

to confirm the data in the material certificktiw. 

The insuectors consider this to be an inadequate dedication cif 

commercia I-grade material because (1) UC had not established a basis 

for accepting material certifications from its suppliers; (2) UC did 

not verify the chemical and physical properties of coil material by 

performing chemical and physical tests; and (3) LC did not perform 

destructive sample piece testing traceable to each coil of material 

used in fabrication of channel product. 

The process in the press shop, where charwl nuts, pipe clamps, and 

fittings are made, begins with a receipt inspection similar to that 

used in the processing department. The NRC inspectors determined thdt 

since recently implementing a July 31, 1990, revision to QAP 20-01-006, 

"Receiving Inspection," and QAP 20-01-037, "Traceability of Fittings," 

UC did rtiaintain adequate traceability on material in the press shop. 

A lot number is assigned to all incoming shipments and recorded in the 

lot number log book and on a Material Control Ticket. This lot number 

The NRC 

Of these 16 coils, the inspectors determined that only 

I 

104 
ORGANIZATION: UNISTRUT CORPORATION 

WAYNE, MICHIGAN 

REPORT 

NO. : 999000362/90-01 I RESULTS: 

I N S P ECT I 0 M I PAGE 18 01' ;; 

is based on a heat number or batch number supplied by the vendor. If 

Illore than one skid, bundle, lift, or container is received, it may be 

given the same lot number providing the heat number or batch number is 

the same. A materiiil control tag, with the lot number written on it, 

is placed on each skid, bundle, lift, or container. This tag remains 

with the material as it progresses through the press shop operations 

until it emerges as a finished product. 

1. Channel nuts are considered bj UC to be more critical than the 

other products produced in the press shop and are heat treated 

during the manufacturing process. As a result, Chdnnel nuts are 

tested more rigorously than other press shop products. Once per 

shift, at a mininium, a channel nut that has been heat treated is 

tested to determine whether proper case hardening depth has been 

achieved. 

sample of 5 channel nuts is selected for slip testing. 

consists of performing hardness tests 011 the nuts and then 

mounting them in a short length of channel using a special 

fitting, applying a load to the top of the fitting, and recording 

the load at which the assembly begins to slip down the channel. 

A torque test is also performed on a separate sample of 5 channel 

nuts from each lot after hedt treatment. This test consists of 

attaching a fitting to a piece of channel by means of one nut and 

bolt and then torquing the bolt to the recommended torque value 

to ensure that neither the nut nor the bolt exhibit signs of 

failure. 

In addition, after heat trtrutr;ient of each lot, a 

This 

2. Testing of the other products produced in the press shop, pipe 

clamps and fittings, is limited to a bend/crush test of 90 degree 

angle fittings. 

per heat basis, on the conip1ete.G fitting. 

striking the fitting with a hammer until it is bent back upon 

itself and then taking another fitting and hammering it the 

opposite direction. 

they pass the test. 

This test is performed, on a minimum of a once 

It consists of merely 

If the fittings don't crack or break apart, 

The inspectors consider the press shop process tu be an inddequate 

dedication because (1) UC had not established an adequate basis for 

accepting the iiidteriiil certifications from its suppliers; (2) did 

riot verify the chemical and physical properties of coil material by 

performing chemical and physical tests; (3) until recently, did not 

maintain traceability on material in the press shop; and (4) except 

for charinel nuts arid 90-degree angle fittings, did not perform any 

testing of their products. 

105 
ORGANIZATION: UNISTRUT CORPORATION 

WAYNE, PIICHIGAN 

NO.: 999000362/90-01 PAGE 19 of 27 

After reviewing the operations htld prccudures ut both the processing 

department and the press shop, the NRC inspectors concluded that UC did riot 

have an adequate program for dedicating commercial-grade items for use in 

nuclear safety-related applications. This conclusion was reached because 

(1) UC had not established an adequate basis for accepting the material 

certifications from its suppliers; (2) did not verify the chemical and 

roperties of coi 1 material by performing chemical and physical 

tests; (3 P until recently, did nut niaititain traceability on material in 

the press shop; (4) did not perform testing traceable to exh coil of 

material used in the fabrication of channel products; and (5) except for 

channel nuts and 90-degree angle fittings, UC did not perforr,i any testing 

on products produced in the press shop. Therefore, UC supplied inadequately 

dedicated parts such as structural framing ch,inriel, chaiinel nuts, screws, 

nuts, washers, general fittings, and mechanical accessories as safetyrelated cuipurimt5 to the nuclear industry. As a result, Noncotiformance 

(90-01-06) was identified in this area of the inspection. 

6.2 - Dedicaticn of Fasteners 

As stated previously in Section E.3.0 of this report, UC had never 

perfurriled a commercial-grade dedication of fasteners supplied to the 

nuclear industry as safety-related. In addition, UC changed the licensees' 

material specifications imposed in PGs arid ordered fasteners from GF to a 

differelit material specification, even though UC issued C of C to 

licensees that certify compliance to the licensees' material specification. 

GF is only a material supplier of fasteners slid perfornis no manufacturing 

activities. 

suppliers and manufacturers. 

was not examined during this inspection. 

follows. 

GF procures all of the fasteners supplied to UC from other 

The bases of GF's procurement of fasteners 

Examples of these concerns are as 

a. 

b. 

River Bend Nuclear Station's PO No. 89-4-77172, dated August 25, 1989, 

ordered ASTM A3G7, grade A, safety-related fasteners. UC issued 

ccimiier'cial-grade PO No. P-02694, September 18, 1989, to GF ordering 

fasteners to material specification SAE-J429, grade 2. UC issued 

River Bend Nuclear Station a C of C, October 5, 1989, that certified 

that the material met ASTM A307, grade A. 

Bechtel Constructors Corp.'s PO No. 20557-F1020-Q, July 17, 1990, 

ordered ASTM A307, grade A, safety-related fasteners for the Palisades 

Kuclear Plant. UC issued commercial-grade PO No, P-03779, October 4, 

1990, to GF and ordered the fasteners to material specification 

SAE-J429, grade 2. UC issued Palisades Nuclear Pldni C of C, 

October 10, 1990, that certifiei; ;hilt the rtlti.terial met ASTM A307, 

grade A. 

106 
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REPORT INSPECTION 

NO. : 999000362/90-01 RESULTS : PAGE 20 of 27 

The UC, "General Engineering Catalog," No. 11, states, in part, that screws 

conform to SAE-J429, grade 2, and also meet and exceed ASTM A307. 

however, did not document an evaluation of these material specifications to 

subs tanti ate this claim. 

UC 

7.0 Purchase Orders and UC Certificates of Conformance 

The NRC inspectors reviewed the licensees' POs to UC and the UC POs to 

suppliers to determine whether or not UC implemented measures to adequately 

assure that procurement documents include the applicable requirements, 

design bases, and quality program requirements. The review was intended to 

determine whether or not UC had an adequate basis to substantiate the C of Cs 

issued to licensees that certify compliance to their PO requirements. 

UC staff told NRC inspectors that, based on UC's review, it had determined 

that UC first began issuing C of C tu the requirements of 10 CFR Part 21 

and 10 CFR Part 50, Appendix B in August 1988. 

however, identified PO No. MMP2-FPO-15391, July 27, 1982, and PO No. 

NMP2-FPO-16083, September 17, 1982, from Stone & Webster Engineering 

Corporation to UC for safety-related pipe clamps. 

supply Niagara Mohawk Power Corporation's Nine Mile Point Nuclear Station 

and both imposed on UC the requirements of 10 CFR Part 21 and 10 CFR Part 50, 

Appendix B. UC issued C of C August 1982 and November 1, 1982, 

respectively, certifying that, "the parts described hereon have been 

processed in accordance with your purchase order instructions." UC claimed 

that it did not understand that in issuing such a certification that it had 

also certified compliance with 10 CFR Part 21 and 10 CFR Part 50, Appendix B 

requirements imposed in the customers' PO. 

paragraph b., below, and represent a lack of understanding by UC chat 

significantly impacts the bases of evaluating UC C of C that are not valid 

because UC ordered the base material for the clamps as commercial-grade dnd 

failed to perform a dedication of the material. 

The NRC inspectors, 

Both POs were issued to 

These C of C are listed in 

The procurement process starts when licensees place PO'S with UC Authorized 

Service Centers. The Authorized Service Centers then send the order 

information to UC. UC's QAM requires that the Quality Assurance Department 

review and approve all licensees' POs requiring C of C before orders are 

accepted by Customer Service. 

is able to certify compliance with all PO requirements. 

1990, when the current policy was implemented, the Authorized Service 

Centers frequently sent to UC only the part number and a statement that a 

C of C was required. 

PO requirements were not sent to UC. UC personnel interviewed during the 

This practice is intended to assure that UC 

Before July of 

In most cases reviewed by the inspectors, the original 

107 
ORGANIZATION: UNISTRUT CORPORATION 

WAYNE, MICHIGAN 

REPORT 

NO.: 999000362/90-01 PAGE 21 of 27 

inspection stated that it was conimon practice for the Authorized Service 

Centers to filter the PO requirerrreiits, only informing UC that a C of C was 

required, dnd not passing on all of the quality and technical requirements 

imposed in the licenste's PO. 

The inspectors identified various examples described below where UC 

certified compliance to requirements that they were not aware of and did 

not comply with, or revised certifications without verifying prior 

conipl ihnce to requirements imposed. 

a. The following are examples where UC issued C of Cs certifying 

compliance to the requirements of 10 CFR Part 21 and 10 CFR Pdrt 50, 

Appendix B without having received the licensee's PO or estdblishing 

an understanding of the licensee's requirements: 

UC ORDER NO. DATE LICENSEE -~ PRODUCT 

3030-0861 

3050-0083 

3050-0167 

3030-1040 

0450-2231 

3030 -1 08 1 

0450-2265 

0450-2295 

C45G-2550 

0450-2551 

03 10-2388 

0330-1893 

4/8/87 

12/15/87 

7/11/68 

7/22/88 

10 118188 

101 19/88 

1 1 1281 89 

1/10/89 

9/18/89 

9 ,I 1 8 / a9 

121 7 189 

1/5/90 

Browns Ferry 

Browns Ferry 

Browns Ferry 

Sequoyah 

Diablo Canyon 

Browns Ferry 

Diablo Canyon 

Diablo Canyon 

Diablo Canyon 

Diablo Canyon 

Hatch 

Comariche Peak 

pipe clamps 

channel nutslfitting 

channel/f asteners/ 

fittings 

pipe clamps 

charine l/pi pe clamps/ 

f i tt i ngs 

fastenerslpipe clamp 

channel/pi pe clamps/ 

fasteners 

pipe clamps 

pipe clamps/fittings 

pipe clamps/channel/ 

channel ruts 

pipe clamps/channel 

nuts 

channel nuts 

b. The following are examples where UC received a copy of the licenser's 

original PO with the Authorized Service Center's order. The original 

licensees' POs reviewed stated that 10 CFR Part 21 and 10 CFR Part 50, 

Appendix B were applicable to the order. UC issued C of C certifying 

conp1iillicE with the "PO instructio1,s." UC staff claimed they did not 

understand that in doing so it also certified compliance with 10 CFR 

Part 21 and 10 CFR Part 50, Appendix B requirements. 
ORGANIZATION : UNISTRUT CORPORATION 

WAYNE, MICHIGAN 

REPORT INSPECTION 

NO.: 999000362/90-01 RESULTS : PAGE 22 of 27 

UC ORDER NO. 

1070-1245 

1070- 132 5 

2010-1318 

0310-1478 

0130-1917 

0130- 1944 

0370-0024 

0450- 1944 

24 5C-LG27 

0450-2 028 

0450-2057 

0450-2059 

-- DATE LICENSEE 

8/82 Nine Nile 

11/1/82 Nine Mile 

1/2/86 Cal laway 

1/16/87 Vcrgtle 

5/7/87 L 

6/15/87 L 

10/20/87 R 

12/1/87 D 

3/10/88 D 

3110188 D 

PRODUCT 

pipe clanips 

pipe clamps 

channel 

pipe clamps 

merick channel 

merick channel 

ver Bend fasteners 

ablo Canyon channel nuts/ 

fittings 

ablo Canyon channel/pipe clmps/ 

f I -it 1 ng s/ f a s t en er s 

ablo Canyon chamellfittings 

4/20/88 Diablo Canyon fasteners 

4/21/88 Diablo Canyon fi ttingslfasteners 

c. The foliowing are exatiiples where licensees’ POs specifically requested 

certification of compliance to the requirements of 10 CFR Part 21 or 

10 CFR Part 50, Appendix B. 

“Quality control program conforms to ANSI N45.2 and 10 CFR Part 50, 

Appendix B, 10 CFR Part 21 applies to this order.” 

UC issued C of C that certified that the 

LIC GRDER NO. - 

0330-1645 

0310-2208 

0310-2220 

03 10 -2284 

0310-2314 

0330- 1865 

C450-2615 

03 10-2384 

0330-1942 

0310-2574 

0310-2577 

DATE 

12/5/88 

7/19/89 

-- 

7/28/89 

9/21 /89 

10/13/89 

11/16/89 

11/ 17/89 

12/6/89 

3/15/90 

5/8/90 

5/10/90 

LICENSEE -

Comanche Peak 

Sequoya h 

Watts Bar 

Browns Ferry 

S eq u oya h 

Ccnianche Peak 

Diablo Canyon 

Browns Ferry 

Comanche Ped k 

S e q u oy a h 

Browns Ferry 

PRODUCT 

channe 1 /f i tt i ngs/ 

fasteners 

channel nuts/pipe 

ciamps/fittings 

fittings 

channel nuts/fitting 

pipe clamps/channel 

nuts 

beam clamps 

fittings/channel nut 

fasteners/channel 

nu ts/p ipe clanips 

pipe clamps 

pipe cianips 

channel nuts 

-- 

d. The following are examples where the licensee’s PO requested 

certification of compliance to ASME Code Section 111, Subsection 

NF-2600. 

without having a quality program that compiies with ASME Code, 

Section I11 requirements, or an understanding of the ASME Code, 

UC issued C of C certifying conipliance to this requirenieirt 

109 
ORGAN I ZAT I ON : UNISTRUT CORPORATION 

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INSPECTION 

RESULTS : I REPORT 

NO. : 999000362/90-01 I PAGE 23 of 27 

e. 

f. 

Section 111, Subsection NF-2600 technical requirements. 

Gici riot have a copy of the ASME Code requirements. 

In fact, UC 

UC ORDER NO. DATE LICENSEE 

30:6-0942 12/4/87 Watts Bar 

3030-1088 11/17/88 Watts Bar 

PRODUCT 

channel/fittings 

fasteners 

The inspectors identified the following examples where UC was 

requested to revise previously issued C of C to incorporate 

statements certifying compliance with 10 CFR Part 21, 10 CFR Part 5C, 

Appendix By and in one case, ASME Code. UC performed these revisions 

without verifying that they had complied with these requirements at 

the time of original certification. 

UC ORDER NO. ~- 0RIG.DATE REV.DATE LICENSEE PRODUCT 

Stock Items 9/18/87 4/3/89 Watts Bar pipe clamps 

Added compliance statement to the 

requirements of ANSI N45.2, 10 CFR 

Part 50, Appendix By 10 CFR Part 21, 

ASME Code, Section 111, Class 2, 

Subsection NF-2600, and Code 

Case No. 1644. 

(PO NO. 87NBA-36447A) 

3030-0870 6/12/87 7/11/89 Watts 6ar channel 

-- Added compliance statement to the 

requirements of 10 CFR Pdrt Ti. 

The NRC inspectors also reviewed UC's orders to its supplier of. 

fasteners. 

with safety-related fasteners, they did not impose 10 CFR Part 50, 

Appeiidix 8, quality requirements on thesc purchase orders or any other 

quality requirements. The following are examples of POs for fasteners 

that did not impose 10 CFR Part 50, Appendix B requirements on the 

vendor and the fasteiiers were supplied tc the riucledr' industry w5th 

UC C Gf C that certified conipliance with 10 CFR Part 50, Appendix B. 

While UC believed that their supplier was providing them 

PURCHASE ORDER - 

P-02170 

P-03418 

P-03779 

P-03791 

1/24/89 

4/20/90 

10/4/90 

10/8/90 

VENDOR 

GF 

GF 

GF 

GF 

110 
ORGANIZATION: UNISTRUT CORPORATION 

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REPORT INSPECTION 

NO. : 999000362/90-01 RESULTS : PAGE 24 of 27 

From this review, the NRC inspectors concluded that UC failed to establish 

measures to msure that regulatory requirements imposed in licensees' POs 

are established and trdnslated into documents that ensure ddequate quality 

is achieved. Moreover, UC failed to assure that compliance to imposed 

reguliitory requirements had been achieved before issuing C of C that 

certified compliance. This condition led UC to make unsubstantiated 

certification statements on C of C issued to licensees. Specifically, 

UC routinely failed to perform the following: 

review of POs frorii 1 iceriseks it1 order to estdblish the requirenients imposed 

on UC; (2) establish measures necessary to trimslate the requirements 

imposed in POs from licensees into specifications or procedures; (3) ensure 

compliarrct. to the requirements impused in POs from licensees before issuing 

C of C that certified compliance; (4) verify UC's compliance prior to 

revising previously issued C of C; and (5) impose 10 CFR Pbrt 50, Appendix 

B or similar quality requirements in FOs to suppiiers of safety-reletra 

fasteners. 

ared of the inspection. 

(1) contrr;l the receipt and 

As a result, Nonconforiaaiice (90-01-7) was identifiec; in this 

8.0 

The inspectors reviewed UC's control of purchased material with particular 

emphasis on how they assured themselves that purchased material, equipment, 

and services conformed to the procurement documents and whether UC 

effectively assessed the control of qualit;/ of their suppliers. 

supplier files containing cort*espondence, questioiinaires, checklists, and 

audit reports were reviewed. 

UC Audits of Suppliers - 

UC's 

A review of UC's Approved Supplier List, September 4, 1990, revealed that 

all but 2 of the 19 suppliers listed were considered to be coniniercial-grade 

suppliers. Three of these commercial-grade suppliers' files, Rouge Steel 

Co., Stateline Steel Corp., arid Triad Metals Inc. were reviewed. 

out an annual vendor quality audit questionnaire to its suppliers. 

questionnaire consists of a cover shett on which the supplier is asked to 

provide information including the standard, if any, to which their quality 

program complies. 

about the supplier's quality progrm. 

suppliers stated that they had conimercial-grade programs that did not 

comply with the requirements of Appendix B to 10 CFR Part 50. 

UC sends 

This 

The questionnaire also includes 4 pages of questions 

The questimnaire from all 3 

ilC had performed audits of these suppliers on a triennial basis. 

dudits were performed by using the questionnaire as a checklist fc,i the UC 

auditor. 

answered the questions correctly. 

The 

The auditor, using the checklist, verified thdt the vendor had 

These audit checklists/reports did not 

111 
ORGAN I ZAT ION : 

REPORT 

NO.: 999000362/90-01 

UNISTRUT CORPORATION 

WAYNE, MICHIGAN 

INSPECTION 

RESULTS : PAGE 25 of 27 

contain objective evidence to support the conclusions of the auditor and 

did riot provide any detail of the activities audited. 

determined that the audits may be considered adequdte to approve a 

commercial-grade vendor. The audits are not adequate, however, to provide 

UC with a basis to accept material certifications from these vendors, as is 

their current practice. 

The NRC inspectors 

The file of UC's major 10 CFR Part 50, Appendix B supplier of safetyrelated fdsteners, General Fasteners Company (GF), was also reviewed. This 

file revealed that UC utilized the same vendor quality audit questionnaire. 

GF had previously indicated on the questionnaire that their quality program 

had been based on 10 CFR Part 50, Appendix B and until 1990 UC had audited 

GF on a triennial basis. Until the most recent audit, UC performed audits 

using the same questionnaire used for their audits of commercial-grade 

suppliers. 

extensive checklist, in keeping with 10 CFR Part 50, Apperidix B quality 

criteria. 

10 CFR Part 50, Appendix B requirements. 

reports included objective evidetice to support the conclusions of the 

auditors or substantiate UC's use of GF as a supplier of safety-related 

fasteners with a quality program that complied with 10 CFR Part 50, 

Appendix B. As described in Section E.3.0 of this report, GF advised UC on 

October 26, 1990, that it did not have, nor had it ever had, a quality 

program that niet the NRC requirements in 10 CFR Part 50, Appendix B and 

10 CFR Part 21. 

The audit conducted on September 11, 1990 utilized a more 

The auditors concluded that GF had a quality program that met 

None of these audit checklists/ 

Additionally, as part of the inspection at UC, one of the NRC inspectors 

visited GF's fdcility in order to better understand the procurement 

relationship between UC and GF. GF stated that they had an agIw,.c,.t with 

UC to certify compliance to 10 CFR Part 50, Appendix B based on UC's 

reported belief that GF had an adequate quality program. 

personnel at GF admitted, however, that they do not now nor have they ever 

had a quality program that complies with 10 CFR Part 50, Appendix 6 to 

supply nuclear safety-re 1 ated fasteners . 

Senior quality 

The inspectors concluded that UC failed to establish adequate measures to 

assess and document the effectiveness of the quality programs of vendors 

supplying mdterials and services. Specifically, the UC audits failed to 

discover that the quality program of their fastener supplier did not nieet 

10 CFR Part 50, Appendix B requirements. 

their commercial-grade suppliers of steel coil rriaterial, which furnished 

material certifications used by UC in the commercial-grade dedication 

process. This is another occurrence of Nonconformance 86-01-05, identified 

In addition, UC did not audit 

112 
REPORT INSPECTION 

NO. : 999000362/90-01 RESULTS : 

by the NRC during the previous inspection and documented in IR 

No. 99900362/86-01. 

in this area of the inspectioii. 

As a result, Noncunformarice (90-01-08) was identified 

PAGE 26 of 27 

9.0 Licensee Audits 

Audits of UC performed by licensees were reviewed by the NRC inspecturs to 

determine how licensees evaluated UC's quality progrdm, what their findings 

were, and how UC responded to the findings. 

there was a wide range of audit conclusions among licensees. 

inspectors also noted that although UC's quality program changed little 

from 1987 until June 1990, some licensees' audit conclusions during the same 

time period changed significantly. 

As an example, audits perforriled bj Tennessee Valley Authority (TVA) on 

April 10-12, 1989, and May 21-24, and Jurie 26-27, 1990, were reviewed. 

Both the 1989 arid 1990 audit reports were comprehensive and provided 

detailed findings, indicating that the audits had been thorough. 

audit reaffirmed TVA's previous conclusion that UC was capable of supplying 

10 CFR Part 50, Appendix B quality safety-related parts and also meeting 

the requirements of ASME Code, Section 111, even though 10 significant 

findings were identified during the audit. The NRC inspectors noted that 

UC had never been accredited by ASME. 

not know what ASME Code requirements were ilor did they have ASME Code books 

for reference and they did not believe that the UC QAM complied with ASME 

Code requirements. 

The inspectors noted that 

The 

The 1989 

The UC staff stated that they did 

UC's response to TVA's 1989 audit findings was evaluated by reviewing TVA's 

1990 audit report in which TVA rioted that 9 out of the 10 findings iron1 the 

1989 audit had not been corrected. 

downgrading UC from a 10 CFR Part 50, Appendix B supplier to a commercialgrade supplier. 

TVA reacted to UC's response by 

Additionally, audits by Pennsylvania Power and Light on June 5-7, 1990, 

American Electric Power on December 13-14, 1989, and Public Service 

Electric and Gas oil September 22-23, 1987, were reviewed. Based on the 

audit reports, all of these utilities concluded that UC had an acceptable 

quality program, meeting the requirements of 10 CFR Part 50, Appendix B. 

Audits performed by Pacific Gas & Electric (PG&E) on July 2, 1987, and 

July 26-27, 1988, were also reviewed. 

inspectors because the audit conclusions differed from the other utilities 

previously mentioned and because they also changed from one year to the 

These audits were of interest to the 

113 
ORGANIZATION: UNISTRUT CORPORATION 

WAYNE, MICHIGAN 

next. 

the requirements of Criteria I, 11, and V through XVII of Appendix B to 

10 CFR Part 50, 

requirements of Criteria VI1 through XV, and XVII, a significant change 

from the previous year. 

UC as safety-related in the period 1987 through 1989. 

The 1987 audit found that UC's quality program adequately met only 

The 1988 audit found that UC's quality program only niet the 

However, PG&E continued to purchase products from 

The audit reviewed by the inspectors performed by Florida Power & Light 

(FP&L) on May 3, 1989, resulted in audit conclusions that were opposite 

frorn the utilities that considered UC to be a 10 CFR Part 50, Appendix B 

supplier. 

suppl ier. 

FP&L's audit concluded that UC was strictly a commercial-grade 

F. PERSONNEL CONTACTED: - 

#Kris Weger 

#*Charles Nehls 

#*John Springstead 

Del Hamilton 

Rich Farago 

Jerry Dowling 

John Breier 

Richard Hines 

Pat McConalogue 

Vice President and General Manager 

Manager, Production Engineering 

Manager , Qual i ty Assurarice 

Processing Inspector 

Senior Processing Inspector 

Quality Technicidn 

Manager, Customer Service 

Quality Assurance Administrator 

Purchasing Agent 

t Present during the entrance meeting * Present during the exit meeting 

114 
+**** 

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON. D.C. 205554001 

May 6, 1995 

Mr. Clinton W. Deisenroth, President 

Whittaker Electronic Systems 

Whittaker Corporation 

1785 Voyager Avenue 

Simi Valley, CA 93063-3349 

SUBJECT: NRC INSPECTION NO. 99901164/95-01 

Dear Mr. Deisenroth: 

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

(NRC) inspection of Whittaker Electronic Systems at Simi Valley, California, 

conducted by Messrs. R.C. Wilson and R. Mendez of this office on April 3-6, 

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. The inspection consisted of an examination of procedures 

and records, interviews with personnel, and observations by the inspectors. 

The NRC inspectors found that the implementation of your quality assurance 

program failed to meet certain NRC requirements. Specifically, your quality 

assurance surveys of a supplier of hydrogen gas used to calibrate hydrogen 

sensors were not adequate to dedicate commercially procured standards which 

were subsequently used for nuclear safety-related applications. This was an 

example of a generic problem regarding inadequate dedication of calibration 

services until shortly before the NRC inspection. The specific findings and 

references to the pertinent requirements are identified in tne ei~~losures to 

this letter. 

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. 

115 
C. Deisenroth -2- 

In accordance with 10 CFR Part 2.790 of the NRC’s “Rules of Practice,“ a copy 

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

Room. 

Si ncerel y , 

Special Inspection Branch 

Division of Technical Support 

Office of Nuclear Reactor Regulation 

Docket No. 99901164 

Enclosures: 1. Notice of Nonconformance 

2. Inspection Report 99901164/95-01 

116 
NOTICE OF NONCONFORMANCE 

Whittaker Electronic Systems 

Simi Val 1 ey, Cal i forni a 

Docket No.: 99901164 

Based on the results of an inspection conducted on April 3 through 6, 1995, it 

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

NRC requirements as described bel ow. 

Criterion XI1 of Appendix B to 10 CFR Part 50, "Control of Measuring and Test 

Equipment," requires, in part, that measures shall be established to assure 

that testing devices used in activities affecting quality are properly 

calibrated to maintain accuracy within specified limits. 

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

Equipment, Material, and Services," requires, in part, that the effectiveness 

of the control of quality by contractors shall be assessed. 

Section 4.1.1 of the Whittaker Quality Assurance Manual Chapter QAM-07, 

"Control of Purchased Material, Equipment and Services, Revision M, July 29, 

1994, states that Quality Assurance, Production Control, and Purchasing shall 

provide for source evaluation and evaluation of objective evidence of quality. 

Similarly, Section 7.2.1 of Revision L of the Quality Assurance Manual, dated 

September 1, 1989, states that sources to be used are selected based, in part, 

on quality assurance considerations. 

Contrary to the above, at the time of the inspection, documented Whittaker 

qual i ty assurance surveys of Liquid Carbonic Cy1 i nder Gas Products were 

inadequate to dedicate for nuclear safety-related use the hydrogen calibration 

standards provided under a commercial grade purchase order. Furthermore, the 

standards were used in cal i brat i ng a model 113C00264 hydrogen sensor, seri a1 

number 8-2028, shipped to the Cooper nuclear station on April 9, 1993, under 

Nebraska Public Power District Purchase Order No. 362172. (99901264/95-01-01) 

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

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

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

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

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

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

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

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

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

and preventive measures were or wi 11 be compl eted. 

Dated at Rockville, Maryland 

this 5th day of May, 1995 

Enclosure 1 

117 
REPORT NO.: 

ORGANIZATION: 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY: 

INSPECTION DATES: 

LEAD INSPECTOR: 

OTHER INSPECTORS: 

REV I EWED BY : 

APPROVED BY: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION O F T E C H N I C A L S U P P O R T 

99901164/95-01 

Whittaker Electronic Systems 

Whittaker Corporation 

1785 Voyager Avenue 

Simi Val 1 ey, Cal i forni a 93063-8000 

Gary E. King 

Director of Product Assurance 

Suppl ies high-temperature cab1 es and connectors for 

i n-core i nstrumentat i on, and hydrogen and oxygen 

gas analyzers 

April 3-6, 1995 

Richard C. Wilson, Senior Engineer 

Vendor Inspection Section (VIS) 

Special Inspection Branch (TSIB) 

Roger Mendez, Reactor Inspector 

Region I11 

Gregory C. Cwal ina, Chief, VIS/TSIB 

Robert ‘M. Gallo, Chief, TSIB 

Date 

I_’ 7 -. 

Date 

Enclosure 2 

118 
SCOPE OF INSPECTION: 

Whittaker designs and manufactures cables, connectors, and instruments, mostly 

for mil i tarylaerospace and nuclear appl ications. 

primarily high temperature, radiation-resistant cables and connectors for 

applications such as in-core instrumentation and core exit thermocouples, as 

well as hydrogen and oxygen gas analyzers. 

Nuclear products are 

The NRC inspectors reviewed the implementation of selected portions of 

Whittaker’s Appendix B quality assurance (QA) program for supplying safetygrade components, monitored corrections to discrepancies in Whittaker’s 

manufacturing controls that were identified in a previous NRC inspection, and 

reviewed Whittaker’s 10 CFR Part 21 program including reports that have been 

submitted to the NRC. The inspection bases were 10 CFR Part 50, Appendix B, 

and 10 CFR Part 21. 

1.1 Nonconformance 99901164/94-01-01 

Contrary to Criteria XI1 and VI1 of Appendix B to 10 CFR Part 50, which were 

imposed on Whittaker by 1 icensee purchase orders, Whittaker did not assure 

that QA surveys of a supplier of hydrogen gas calibration standards were 

adequate to dedicate the standards for nuclear safety-related use. Furthermore, the standards were used to calibrate a safety-related sensor (see 

Section 3.4 of this inspection report). 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

Although no open findings remained from previous NRC inspections, 

Paragraph E.3 of NRC Inspection Report No. 99901164/89-01 identified certain 

discrepancies in the documentation of manufacturing controls for proprietary 

processes, and stated that the NRC inspector observed that Whittaker had 

already begun to address these discrepancies. As discussed in Section 3.3 of 

this inspection report, the inspectors believe that Whittaker’s manufacturing 

control s are now appropriate. 

3 

3.1 

INSPECTION FINDINGS AND OTHER COMMENTS 

Entrance and Exit Meetinqs 

In the entrance meeting on April 3, 1995, the NRC inspectors discussed the 

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

interfaces with Whittaker management and staff. 

April 6, 1995, the inspectors discussed their findings and concerns with 

Whi ttaker management and staff. 

In the exit meeting on 

3.2 Oualitv Assurance Proqram and Orqanization 

The inspectors selectively reviewed the Whittaker Electronic Resources 

Division QA program established by the QA Manual Revision effective July 30, 

1994. The manual provided the top level requirements, with implementation 

covered in lower tier QA procedures (QAPs) and Quality operating procedures. 

Based on the review of the QAM and QAPs and discussions with the QA manager, 

the inspectors determined that the QA organization had sufficient authority 

2 

119 
and organizational freedom to identify quality problems, initiate solutions, 

and verify implementation of the solutions. 

3.3 Translation of Purchase Order Reauirements 

The inspectors selectively reviewed files for eight purchase orders (POs) to 

determine if the PO requirements were correctly trans1 ated into documented 

specifications, procedures, drawings and testing requirements for fabricating 

safety-related cables, connectors, and gas analyzers and associated 

rep1 acement parts. The inspectors determined that Whi ttaker had an excel 1 ent 

method for translating the PO requirements into the final product. Whittaker 

used a routing card system for documenting fabrication which included 

drawings, procedures, operational steps, operational checks, and final testing 

by quality control. 

control inspector's stamp. 

Each step is verified by the operator's or quality 

The inspector witnessed one cable job in progress and verified that the 

operator was following procedures and that all related procedures and 

instructions were at the job site. In addition, the inspector verified that 

the production personnel followed each step, and that quality control 

documented each inspection or test in the fabrication process. 

concluded that Whittaker's controls effectively ensured that customer PO 

requirements were correctly incorporated into finished products. 

The inspectors 

The inspector, however, noted a discrepancy during a review of Shop Order 

No. 6393A, cables for the San Onofre nuclear power station under ABB 

Combustion Engineering (CE) PO 403770-01 to Whittaker. The inspector observed 

that the configuration control document, which listed documents used in making 

the cable assemblies and which had been checked by document control personnel 

for the current revision status, referenced an outdated procedure. Procedure 

No. MPS 3.4-2, "Torch and Resistance Brazing," Revision B, was checked as 

being the latest applicable revision, although revision C was in effect. The 

cable was fabricated in February 1995, and revision C became effective on 

April 20, 1993. The inspectors reviewed the procedure and found that the 

revision did not affect the fabrication of the cables. 

observed by the inspectors were current. 

All other procedures 

Another concern involved the environmental qualification of the cables and 

connectors in the same shop order. A 1984 CE report establishes environmental 

qualification of the cables and connectors. The qualified connector design 

used a Bi-wax 142 potting compound for pin anchoring, positioning, and spacing 

under the welded sleeve joining the connector and cable. A change order to 

the CE PO directed changing the potting material to Epoxilite 5403, to provide 

better support for the pins during connector mating. CE owns the qualification 

test report, and Whittaker does not certify conformance to it. Thus, responsibility for the environmental qualification of the revised cable and connector 

assemblies lies with CE. According to Whittaker personnel, CE documented 

evaluation of the change in CE-NPSD-968-P, "Qualification Summary Report: 

Qualification of Potting Compounds for Use in MI Cables," December 1994. The 

NRC will review that document during a future inspection at CE. 

The inspectors concluded that Whi ttaker exercises appropriate procedural 

control over manufacturing activities. 

3 

120 
3.4 Accuracv and Calibration of Test Eauipment 

Nebraska Public Power District PO No. 362172 dated February 2, 1993, covered 

testing and rebuilding one hydrogen sensor and one oxygen sensor. The PO 

specified applicability of an Exo-Sensor environmental qualification test 

report, and invoked 10 CFR Part 50, Appendix By and 10 CFR Part 21. Whittaker 

assigned Shop Order (SO) No. 12080A. The inspector's review concentrated on 

the hydrogen sensor, part number 1130002G4, serial number B-2028. The Manual 

Rework Router, which controlled work on the sensor, specified final test per 

Qual i ty Operating Procedure QOP-023, "H, Containment Sensor Performance 

Characterization." QOP-023 specified calibrating the sensor with three 

different calibration gas supplies in the range of 1% to 6% hydrogen in 

nitrogen. 

Whittaker obtained calibrating gas from Liquid Carbonic Cy1 inder Gas Products 

with commercial grade POs. A "Supplier Quality/Inspection System 

Questionnaire" from Liquid Carbonics dated October 12, 1992, was signed off by 

Whittaker QA on October 21, 1992. 

check1 ist completed by the supplier, which does not provide the objective 

evidence at the Appendix B level necessary to dedicate commercial grade 

services for nuclear safety-related applications. Whittaker took no further 

actions to assure the quality of the material supplied by Liquid Carbonics. 

As part of an internal audit, Whittaker added a notation on the questionnaire 

dated October 22, 1993, that an audit was required for gas certifications to 

support the listing of Liquid Carbonics on the Approved Suppliers List (ASL). 

Nonetheless, Whittaker continued to obtain hydrogen without an audit; the 

inspector observed three cylinders with calibration certifications dated 

August and October 1994. The inspector noted that the March 23, 1995, ASL 

still stated that an audit was required. A whiteboard in the QA area also 

included Liquid Carbonics in a list of suppliers to be audited, but no date 

was defined. 

inspectors to perform the audit by May 1, 1995. 

The questionnaire was a military-type 

The Whittaker director of product assurance committed to the NRC 

The inspector reviewed a recent Whittaker audit report for another supplier, 

EIL Instruments, Inc., dated March 27, 1995, and found that report to contain 

the essential elements of a survey to dedicate commercial grade calibration 

services. 

Criterion XI1 of 10 CFR Part 50, Appendix B, requires proper control of 

testing devices affecting quality, and Criterion VI1 requires assessing the 

effectiveness of the control of quality by suppliers. Whittaker's failure to 

perform a commercial grade survey of Liquid Carbonics which was adequate to 

dedicate the helium gas calibration standard constitutes Nonconformance 

NO. 99901164/95-01-01. 

The inspector also reviewed the control and calibration of Whittaker's 

in-house measuring and test equipment. Records showed all the instruments in 

the program to be calibrated within the scheduled intervals. The Whittaker 

Metrology Report data sheet used to document calibration of standards 

contained a "Test Results" section with four columns titled Function Tested, 

To1 erance, Received Val ue and Re1 eased Value. Numerical data were not entered 

in these columns on the sheets observed by the inspector. Whittaker personnel 

stated that numerical data are not recorded unless an instrument is found out 

of tolerance. Whittaker personnel did place checkmarks in a section 

indicating the results of the calibration as 0-50510, 50-75%, or 75-100% of the 

4 

121 
acceptance range, to provide general trending information. However, the 

inspectors noted that instruments such as multimeters had a number of 

functions with different ranges and accuracies, and checking of one of the 

boxes may not apply to all of the ranges. 

Whittaker personnel indicated that they interpreted their calibration 

procedures to require calibration of all of,the ranges of an instrument, using 

the manufacturer’s manual and the accuracy criteria stated in the manual, 

although the procedures did not specifically so state. 

product assurance agreed, by May 1, 1995, to ensure that the documentation of 

in-house calibrations clearly defines the specific ranges tested and the 

acceptance criteria for each calibration. One possibility is to change the 

procedures to specify that all ranges will be tested against the accuracy 

specifications of a designated revision of the manufacturer’s manual. 

The director of 

3.5 10 CFR Part 21 Proqram 

The NRC inspector reviewed Whittaker’s procedure for reporting activities in 

accordance with 10 CFR Part 21, QA Procedure (QAP) 15-02, “Substantial Safety 

Hazard Reporting,“ Revision A, dated September 8, 1994. The inspector 

discussed strong and weak points of the procedure with the Whittaker director 

of product assurance and the QA engineer who wrote the procedure. 

QAP 15-02 focused on evaluating deviations and failures to comply with the 

technical requirements of procurement documents, and only briefly addressed 

the identification and reporting of such concerns. 

specifically encourage or require reporting by the employees most likely to 

identify potential deviations or failures to comply, except that the section 

on discovery procedure (5.2.1) states that “Discovery occurs when a material 

or equipment deviation or discrepancy is documented, e.g., on a Defect Report, 

MRT [Material Removal Ticket], etc.; or a failure to comply is documented.” 

The inspector pointed out that this statement does not cover all possible 

sources of potential deviations, and suggested revising the procedure to 

emphasize such reporting. 

QAP 15-02 did not 

The inspector also noted that Whittaker would rarely have the plant-specific 

information necessary to perform the required evaluation of deviations, and 

suggested instead that the procedure concentrate on the five-day notification 

of customers addressed in 10 CFR 21.21(b), so that customers can perform the 

evaluation. The suggested changes would result in a shorter procedure that 

would better address the Part 21-related activities that Whittaker normally 

performs. 

Review of selected specific issues, as detailed below, indicated that 

Whittaker’s Part 21 reporting program was functioning properly. The inspector 

noted that copies of QAP 15-02 and 10 CFR Part 21 were posted in several 

locations, and a copy of section 206 of the Energy Reorganization Act of 1974 

was posted on one bulletin board. The inspectors concluded that Whittaker’s 

activities with respect to 10 CFR Part 21 appeared to be acceptable. 

3.6 Review of SDecific Part 21 Reports 

Hydrogen analyzer sensor low sensitivity - Interim Report dated 

February 3, 1994, and Final Report dated June 9, 1994 

5 

122 
The concern, originally identified at the Prairie Island nuclear power 

station, was determined by Whittaker to involve a manufacturing lot also 

affecting the H.B. Robinson and San Onofre stations. The manufacturing lot 

was produced by a newly-certified technician and a trainee, using a procedure 

which required repetitively adding electrolyte, evacuating, and venting a 

sensor until bubbles no longer came out of solution. Evaluation of returned 

sensors showed low electrolyte levels, indicating incomplete initial fill. 

The manufacturing procedure was revised to add a specification for a minimum 

electrolyte volume. 

notifications, and corrective actions to be acceptable. 

The inspector considered Whittaker’s analysis, 

Hydrogen analyzer flow switch wire insulation - Written notification dated 

March 8, 1991 and Licensee Event Report No. 91-003 dated March 4, 1991, 

from the Fermi 2 nuclear power plant. 

The concern involved the use of Teflon wire insulation, which could suffer 

radiation degradation and cause a short circuit, instead of the qualified 

Tefzel insulation for a reed switch in a hydrogen analyzer flow switch. The 

concern arose from the acquisition of Exo-Sensor, the original manufacturer, 

by Whittaker. 

supplied commercial grade by Chem-Tec with Teflon wire insulation; the ExoSensor drawing contained a note specifying Tefzel, which was available from 

the switch supplier as an option. 

was to include drawing notes in purchase orders. Thus, Whittaker’s first 

procurement of these switches failed to specify Tefzel. The drawing was 

changed to assign a unique part number to the switch using Tefzel. 

inspector considered Whi ttaker’s analysis, notifications, and corrective 

actions to be acceptable. 

The reed switch specified on Exo-Sensor’s drawing was normally 

Exo-Sensor’s practice, but not Whittaker’s, 

The 

Excessive play in control rod step counters - Licensee Event Report (LER) 

No. 93-005, Revision 1, dated February 22, 1993, from the Sequoyah unit 2 

nuclear pl ant. 

Although this concern was not reported as a formal Part 21 notification, 

the inspector pursued it as having reportability potential. 

that the malfunctioning step counters were manufactured by Whittaker and 

obtained from Scientific Sales, Inc. The LER failed to state that the 

counters were supplied by Whittaker as commercial grade items. 

had no Part 21 responsibilities for the counters. 

The LER stated 

Thus Whittaker 

3.7 Commercial Grade Item Dedication 

One of the eight licensee PO files reviewed by the NRC inspectors included 

dedication of a commercial grade power supply for nuclear safety-related use. 

The dedication was accomplished by functional testing in accordance with a 

five-page test procedure. The NRC inspector considered the dedication to be 

adequate. The inspector noted that Whittaker’s current dedication form 

appeared to focus on the essential elements of a dedication: equipment 

description, safety function, failure modes, critical characteristics, and the 

verification requirement for each characteristic. 

The inspector also reviewed files for two licensee inspections of Whittaker, a 

surveillance by an individual utility in January 1994, and an audit by an 

industry group in September 1994. Both inspections identified findings for 

inconsistencies in Whittaker’s implementation of its dedication program, 

6 

123 
particularly with respect to completeness of the critical characteristics 

list. 

with implementation of the dedication program. 

These licensee activities appear to be adequately addressing concerns 

4 PERSONNEL CONTACTED 

Whittaker: 

t 

+ 

+ * 

* 

* 

C.W. Deisenroth, President 

G.E. King, Director of Product Assurance 

F. P. Rudek, Program Manager 

L.L. Nickoley, Quality Engineering Manager 

G.D. Griffis, Quality Assurance Engineer 

R. Stuart, Quality Assurance Engineer 

J. Williams, Quality Administrator 

K. Johnston, Quality Control Manager 

F. Accario, Calibration Laboratory Manager 

R. Mauldin, Chief Engineer 

J. Fuller, Project Engineer 

6. Battin, Technician 

M. Erixon, Director of Offsite Operations 

D. Briggs, Business Area Manager 

R. McKinnay, Production Control Manager 

C.J. Hutchison, Contract Administrator 

J. Torres, Manufacturing 

+ Attended the entrance meeting on April 3, 1995 

* Attended the exit meeting on April 6, 1995 

7 

124 
Selected Generic Correspondence on the Adequacy of 

Vendor Audits and the Quality of Vendor Products 

Identifier Title 

Information Notice 95-22 Hardened or Contaminated Lubricants Cause Metal 

C1 ad Circuit Breaker Failures 

Information Notice 95-26 Defect in Safety Related Pump Parts Due to 

Inadequate Heat Treatments 

125 
CORRESPONDENCE RELATED TO VENDOR ISSUES 

126 
Mr. Randy P. Smith, Pr ident 

Automatic Switch Company 

50-56 Hanover Road 

Florham Park, NJ 07932 

SUBJECT: ASCO RESPONSE TO A NOTICE OF NONCONFORMANCE CONTAINED IN NRC 

INSPECTION REPORT 99900369/95-01 

This letter addresses the Automatic Switch Company (ASCO), May 19, 1995, 

response letter to the U.S. Nuclear Regulatory Comnission (NRC) regarding the 

Notice of Nonconformance (NON) issued to ASCO on May 15, 1995, with NRC 

Inspection Report 99900369/95-01. 

The NRC review of your Hay 19, 1995, NON response letter found that the ASCO 

statements were generally responsive to our concerns. However, as discussed 

with Hessrs. S. Casadevall and G. Plechy of your staff on May 31, 1995, some 

additional cl ari f icat i ons to your response were required to ensure that your 

corrective action responded to the NON. As a result of our conversation, it 

is our understanding that ASCO will ensure that its new HV-270000 SOV series, 

which has replaced ASCO’s NP8323 SOY series for certain main steam isolation 

valve (MSIV) applications, will not be manufactured using any greases or 

lubricants internal to the SOV. 

Your proposed corrective action to Nonconformance 95-01-01, as a result of our 

discussion, was found to be acceptable and further communication regarding 

tbis issue is not required. We will review the implementation of your 

corrective action during a future NRC inspection to determine the adequacy of 

your corrective action and ensure that full compliance has been achieved and 

will be maintained. 

In accordance with 10 CFR 2.790 of the NRC’s “Rules of Practice,” a copy of 

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

Sincerely , 

(Original signed by) 

Robert M. Gallo, Chief 

Speci a1 Inspection Branch 

Division of Technical Support 

Office of Nuclear Reactor Regulation 

Docket No. 99900369 

cc: See next page. 

127 
cc: 

Hr. F. R. 8. Mattheus, Vice President 

Product Development and Techno1 ogy 

Valve Department 

Automatic Switch Company 

50-56 Hanover Road 

Florham Park, NJ 07932 

Mr. David B, Tompsen, Quality Assurance Director Automatic Switch Company 

50-56 Hanover Road 

Florham Park, NJ 07932 

Mr. William C. Kirchhoff, NED Superintendent 

LaSalle Site Support Engineering 

LaSal 1 e County Nucl ear Station 

RRfl Post Office Box 220 

2601 North 21st Road 

Marsei 11 es , IL 61341 

128 
***if*‘ 

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 205550001 

April 24, 1995 

Mr. Scott Akers, Jr., Chief Executive Officer 

Cardinal Industrial Products, 

Limited Partnership 

3873 West Oquendo 

Las Vegas, Nevada 89118 

SUBJECT: EVALUATION AND CONCLUSION REGARDING THE REPLY TO THE NOTICE OF 

NONCONFORMANCE (NRC INSPECTION REPORT NO. 999001076/94-01) 

Dear Mr. Akers: 

I am responding to your January 30, 1995, reply to the Notice of 

Nonconformance (NON) that was issued by the U.S. Nuclear Regulatory Commission 

(NRC) on December 30, 1994, following an NRC inspection of Cardinal Industrial 

Product’s (Cardinal’s) facility in Las Vegas, Nevada. The Notice of 

Nonconformance identified Cardinal activities which failed to comply with the 

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

Requlations (10 CFR Part 50). 

In your response to Nonconformance 99901076/94-01-03, you state that you have 

been unable to establish a relationship or applicability of the NRC-cited 

requirements to Cardinal’s documented Quality Systems Manual (QSM) 

requirements. You further state that Cardinal’s QSM does, in fact, require 

that procured material and/or product comply with the invoked procurement 

document requirements. Additionally, you state that Criterion VI1 of 10 CFR 

Part 50, Appendix B does not address the use of sampling plans for 

veri f i cati on of compl i ance to the appl i cab1 e procurement document 

requirements, and that you consider that Cardinal’s sampling plan, which was 

reviewed during the inspection, provides reasonable assurance that the product 

complies with the applicable procurement requirements. 

NRC staff comments concerning the re1 ationship between NRC requirements and 

Cardinal’s sampling practices are summarized in the following paragraphs. 

These comments are appl icabl e to procurements where the purchase documents invoke 10 CFR Part 50, Appendix B, as the basis for product acceptance. Where 

specific acceptance criteria or sampling plans are specified by the customer, 

it is expected that the customer (licensee) will assure the compliance with 

Appendix B for the intended application. 

10 CFR Part 50, Appendix B, Criterion I1 requires, in part, that a 

quality assurance program be established to provide control over 

activities affecting the quality of structures, systems, and components. 

Conformance with this Criterion would generally require Cardinal to 

impose the applicable provisions of 10 CFR Part 50, Appendix B, on their 

suppliers and subsuppliers and assure compliance with these provisions. 

129 
Mr. S. Akers Jr. -2- 

10 CFR Part 50, Appendix B, Criterion VI1 states, in part, that 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. 

Society for Testing and Materials (ASTM) fastener specifications 

referenced in most customer purchase orders involving Cardinal are 

considered part of the procurement documents. 

The American 

For certain applications, the NRC staff has accepted commercial grade item 

(CGI) dedication as an alternative method to the programmatic quality 

assurance measures prescribed in Criteria I1 and VI1 of 10 CFR Part 50, 

Appendix B, for assuring that items purchased from commercial sources will 

meet the applicable purchase specification requirements. For organizations 

such as Cardinal that choose to supply material in this manner, it would be 

acceptable to follow NRC CGI dedication guidance. Such guidance has been 

published in NRC Generic Letters 89-02 and 91-05. With respect to sampling 

plans, NRC Inspection Manual Procedure 38703 inspection guidance ctates that 

such plans should be controlled and have an adequate technical basis and also 

discusses the application of different sampling plans based on lot/batch 

homogeneity. 

1 etter. 

These documents are provided to you as enclosures to this 

Cardinal's destructive testing sample plan, which is based on the Shipping Lot 

Method sampling plan described in ASTM Specification A-325, "Specification 

for High Strength Bolts for Structural Steel Joints," does not provide 

reasonable assurance that materi a1 procured from commerci a1 manufacturers or 

distributors (heat/lot traceability not established) will conform to the 

specified requirements. The lack of assurance can be demonstrated by a 

probability calculation assuming a lot of 280 fasteners with 50% of the 

fasteners (140) being nonconforming with respect to chemical composition. 

Since, according to Cardinal's sampling plan, only 2 fasteners are tested for 

chemistry, the probability of the two fasteners being acceptable is 

(140/280)~(139/279)=25%. 

will be rejected. Thus, the confidence level is only 75% that a lot of 

fasteners with 50% of them being nonconforming, will be rejected. 

This implies that there is a 75% chance that the lot 

The above example assumes testing a random population of fasteners. 

staff recognizes that additional assurance of adequate fastener performance 

could be achieved by combining Cardinal's sampling approach with other factors 

such as test results of other parameters which may be indicative of heat/lot 

traceability and homogeneity and the historical performance of fasteners 

obtained from a particular supplier. When such factors are considered, the 

rationale for their use and the basis for their contribution to the overall 

confidence level can be explained in the sampling plan. 

of the NRC inspection nor in its submittal has Cardinal provided documentation 

how, if at all, such factors were considered in the technical basis for its 

sampl ing pl an. 

The NRC 

Neither at the time 

Additionally, the A-325 Shipping Lot Method sampling plan, as used by 

Cardinal, does not appear to be an appropriate application to assure product 

130 
Mr. S. Akers Jr. -3- 

compliance to other fastener specifications when sampling a wide range of 

fastener products obtained from several commercial manufacturers and 

distributors. 

"Specification for Alloy-Steel and Stainless Steel Bolting Material for High 

Temperature Service," require more rigorous sampling as a part of the 

manufacturing process. 

tension test for studs, bolts, and screws for each diameter of every heat of 

material from a given lot of fasteners. Thus, a distributor such as Cardinal 

selling fasteners obtained from an unqualified source would not be able to 

assure conformance to ASTM A-193 by relying on the Shipping Lot Method of 

sampl ing as described in ASTM A-325. 

Other fastener specifications such as ASTM A-193, 

Section 11.1.3 of that specification requires one 

For safety related items, the NRC staff would generally accept a confidence 

level of 90-95% that no more than 5-10% of items sampled are nonconforming. 

The most common sampling standard used by NRC licensees that is accepted by 

the NRC staff is MIL-STD-105D. 

and provides sampling procedures ana tables for inspection by att-ibutes. 

This standard contains various samDling plans 

In your response to Nonconformance 99901076/94-01-03 you also state that all 

of your customers have accepted your existing sampling programs. As discussed 

in Inspection Report 99901076/94-01, the inspectors reviewed the report of the 

last audit by the Nuclear Utilities Procurement Issues Committee (NUPIC) which 

was performed in April, 1994. 

sampling methods used by Cardinal during its commercial grade dedication 

process. Therefore, licensees using the NUPIC audit as the basis for 

qualifying and procuring safety related items from Cardinal would be unaware 

of the use of Cardinal's destructive testing sample plan in its commercial 

grade item dedication process. 

That NUPIC audit report did not address 

The NRC staff has reviewed your response to Nonconformance 99901076/94-01-04 

and finds that it adequately addresses our concerns. We will review your 

corrective actions and actions to prevent recurrence during a future 

inspection. 

You are requested to address the NRC staff comments above concerning your 

January 30, 1995, response to Nonconformance 99901076/94-01-03 and provide 

additional response within 30 days of your receipt of this letter. 

In accordance with Section 2.790 of 10 CFR, a copy of this letter will be 

placed in the NRC's Public Document Room. 

131 
Mr. S. Akers Jr. -5- 

If you should have any questions concerning the inspection or the additional 

information requested by this letter, we will be glad to discuss them with 

you. 

Si ncerel y , (Original signed by) R. Lee Spessard, Director 

Division of Technical Support 

Office of Nuclear Reactor Regulation 

Docket No. : 99901076 

Encl osures : 

1. NRC Generic Letters 89-02 and 91-05 

2. NRC Inspection Manual Procedure 38703 

132 
' ****+ 

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, 0. C. 20555 

;*larch 21, 1389 

TO: ALL HOLDERS OF OPERATING LICENSES AND CONSTRUCTION PERMITS FOR 

NUCLEAR POWER REACTORS 

SUBJECT: ACTIONS TO IMPROVE THE DETECTION OF COUNTERFEIT AND FRAUDULENTLY 

MARKETED PRODUCTS (GENERIC LETTER 89- 02) 

Recent instances of counterfeit and fraudulently marketed vendor products have 

heightened the PIRC's concerns for licensees' capability to assure the quality 

of procured products and to reduce the likelihocd of the use of count.t.feit or 

fraudulent products in nuclear power plants. During recent NRC inspections of 

licensees and vendors, the NRC has observed a wide variety of practices and 

progrm for procurement, receipt inspection, testing and dedication of equipment and material for safety-related applications. letter is to share with all licensees some of the elements of programs that 

appear to be effective in providing the capability to detect counterfeit or 

fraudulently marketed products and in assuring the quality of vendor products. 

The staff is aware of and encourages the industry working group efforts to 

develop guidance in these areas. 

Three characteristics of effective procurement and dedication programs have 

been identified during these NRC inspections. These characteristics are (1) the 

involvement of engineering staff in the procurement and product acceptance process, (2) effective source inspection, receipt inspection, and testing programs, 

and (3) thorough, engineering based, progrsms for review, testing, and dedication of commercial-grade products for suitability for use in safety-related 

applications. 

ments were generally effective in providing enhanced capability to detect 

counterfeit or fraudulently marketed products and in assuring the quality of 

procured products, both in safety-related and other plant systems. 

Licensees may want to consider the lpplicability of these characteristics to 

their programs to reduce the likelihood of the introduction of counterfeit or 

fraudulent products into their plants and to assure the quality of procured 

vendor products. 

The purpose of this generic 

NRC has found that programs that embodied the above three eleIC should be noted that the NRC staff conditionally endorses the guidelines contained in EPRI NP-5652, "Guideline for the Utilization of Commercial-Grade 

Items in Nuclear Safety-Related Applications (NCIG-07) ," that was issued by 

EPRI in June 1988 for evaluating commercial-grade prcducts for suitability for 

use in safety-related applications. 

Background : 

Numerous instances have been identified by the NRC during the past 2 years in 

which the nuclear industry received, accepted, and installed items of hardware 

133 
Flu 1 t iple Addrkssees -2- 

that were not of the quality purported by the manufacturer or suppiier due to 

apparelit misrepresentation. 

in the programs for dedicatiiig commercial-grade products for use in safetyreldted applications. 

Significant deficiencies have also been identified 

The use in nuclear facilities of products which are counterfeit or fraudulently 

marketed increases the likelihood that some plant equipment may not perform 

ds expected. 

regarding this matter.) 

(See the enclosed list of NRC Information Notices and Gulletins 

Discussion: 

Procurement quality assurance (Ck) controls for products to be used in safetyrelated applications are established in Appendix B to 10 CFI? Part 50, and in 

Regulatory Guides 1.28, 1.33, and 1.123. 

vides criteria for QA programs and does not specifically address fraudulent 

activities; however, an effectively implemented licensee QA program would 

incredse the likelihood of detectinq fraudulently marketed vendor 2roducts. 

Although a properly implemented QA program nay inore readily deteci substandard 

products than wi 11 the commercial-grade component upgrade process, a 1 icensee's 

commercial-grade dedication process, ds described in paragraph C., will greatly 

enhance the effectiveness of current upgrade practices. The actions described 

in paragraphs A. and B. have also prgved useful in detecting substandard, 

counterfeit or fraudulently marketed products intended for use in systems 

needed for the safe operation of the facility. 

It is recognized that Appendix B proA. Engineering Involvement in the Procurenient Process 

Appropriate engineering involvement is warranted during the procurement and 

product dcceptance processes, including testing, for products used in 

nuclear power plants. Inadequate engineering involvement has been a 

common weakness in 1 i censees' procurement programs, particuiarly when commercial-grade procurements were involved. Involvement of a licensee's 

engineerin staff in an effective procurement process would normally 

include (1 4 development of specifications to be used for the procurement 

of products to be used in the plant, (2) deterrdnation of the critical 

ch~racteristics of the selected products that are to be verified during 

product acceptance, (3) determination of specific testing requirements 

applicable to the selected products, and (4) evaluation of test results. 

The extent of necessary engineering involvement is dependent on the nature 

and use of the products invclved. 

E. Product Acceptance Programs 

Experience indicates that reliance on part number verification and Certification documentation is insufficient to ensure the quality of procured 

products. Licensees with effective product acceptance programs have 

included receipt/source inspection dnd appropriate testing criteria, 

effective vendor audits, special tests and inspections and post-instal latioil tests in their programs. These licensees have applied the inspection 

12A 
Nultiple Addressees -3- 

C. 

and testing criterid to products procured for use in safety-related systems 

and for all commercial-grade products being evaluated for suitability for 

use in safety-related systenis. 

have required identification and verification of the products' critical 

characteristics. 

consideration may be given to the safety significance, complexity, and 

application of the various products. 

grams, as confirmed by licensee audits, samplinq plans are often utilized 

to perform the required inspections and tests. 

receipt/source inspections and tests, effective licensee programs normally 

verify the traceability to the original manufacturers of procured 

materials, equipment, and components in those cases where original 

manufacturer's certifications are elements of the safety-related procurement or commercial-grade dedi cation program. 

included consideration of audit approach, depth of audit, and audit team 

compociticn and have included appropriate engineering/technical representatives. 

be effective. 

The inspection and testing criteria also 

In selecting the critical characteristics to be verified, 

For suppliers with acceptable QA proIn addition to these 

Effective audits have 

Comprehensive multi-licensee audit teams have also been found to 

Dedication Programs 

It is each licensee's responsibility to provide reasonable assurance that 

nonconforming products are not introduced into their plants. Dedication 

programs that ensure the adequacy of critical parameters of products used 

in safety-related applications can also contribute to the identification 

of counterfeit or fraudulently marketed vendor products. 

The NRC staff believes that licensees Mho use methods similar to those 

described in EPRI hP-5652 "Guideline for the Utilization of ComercialGrade Items in Nuclear Safety-Related Applications (NCIG-07) ,It to verify 

the critical characteristics cf commercial-grade items intended for safetyrelated a~p; i,ations have the basis for effective dedication programs. 

50 

1. 

2. 

Properly implemented, the EPRI guidelines, as modified below, establish 

methods which satisfy existirig requirements of Appendix B to 10 CFR Part 

as they apply to the dedication process of commercial-grade items. 

Acceptance Method 2, "Commercial-Grade Survey of Supplier," 

should not be employed as the basis for accepting items from 

suppliers with undocumented commercial quality control programs or with programs that do not effectively implement their 

own necessary controls. Likewise, Method 2 should not be 

employed as the basis for accepting items from distributors 

unless the survey includes the part manufacturer(s) and the 

survey confirms adequate controls by both the distributor 

and the part manufacturer( s). 

Acceptance Method 4, "Acceptable Supplicr/Item Performance 

Record," should not be employed alone unless: 

135 
flu 1 ti ple ddrdssees -4- 

a. The established historicdl record is based 011 industrywide perforirlance data that is directly applicable to the 

item's critical characteristics and the intended safetyreliited application; and 

b. The manufacturer's IiieGsures for the control of design, 

process, aiid material changes have been adequately iniplernented as verified bj audit (multi-licensee team audits 

are acceptable). 

The NRC stdff believes that if licensees' procurement programs effectively 

implerneilt the elements discussed iii pdragraphs, A., B., and C., they will 

reduce the likelihood of the introduction of counterfeit or fraudulent products into their plants. 

Although no response ti; this letter is required, if you have any questicjns 

reydrdiliy this matter, please coritact the technical contact listed below. 

Si ncere ly , 

/Acting Associ atebbrector for Projects 

Office of Nuclear Reactor Regulation 

Enclosures: 

1. List of Bulletins arid Information Rotices 

'2. List of Recently Issued Generic Letters 

CONTACT : 

E. Oilliam Brach, NRR 

( 301 ) 492 -096 1 

136 
Enclosure 1 

BULLETINS AND INFORMATION NOTICES CONCERNING NONCONFORMING 

MATERIALS AND EQUIPMENT AND INSTANCES OF INADEQUATE 

CEDICATIOL OF EQUIPMENT FOR SAFETY-RELATED APPLICATIONS 

Eulletin No. 

87-02* 

Title 

Fastener Testing to Determine Conformance 

with Applicable Material Specifications 

87-02, Supplement 1" 

87-02, Supplement 2* 

88-05* Nonconforming Materials Supplied by Piping 

Supplies, Inc. , at Folsum, New Jersey, and 

blest Jersey Manufacturing Company at 

Williamstown, New Jersey 

88-05, Supplement l* 

88-05 , Supplement 2* 

8840* Nonconforming Molded-Case Circuit Breakers 

Information Notice No. 

87-66 

88-19" 

88-35 

88-4G* 

Inappropriate Application of Comnercial-Grade 

Components 

Questionable Certificarion of Class 1E 

Components 

Inadequate Licensee Performed Vendor Audits 

Licensee Report of Defective Refurbished 

Ci rcui t Breakers 

86-46, Supplement 1* 

88-46, Supplement 2" 

Date 

11/06/87 

04/22/8& 

06/ 10/88 

05/06/88 

06/15/88 

08/03/88 

11/22/88 

12/3 1/87 

34/26/88 

06/03/88 

07/08/88 

07/21/88 

12 / 30/ 88 

88-48* Licensee Report of Defective Refurbished Valves 07/12/88 

88-48, Supplement 1* 08/ 24/88 

88-97 Potentially Substandard Valve Replacement Parts 12/16/88 

*These i terns reflect instances in which suppliers and manufacturers of safetyrelated material may have intentionally eluded QA requirements to misrepresent 

the quality of their products. 

blem was brought to NRC's attention by either a licensee or a nuclear supplier. 

In the instances mark?d by an asterisk, the pro137 
Enclosure 2 

LIST OF RECENTLY ISSUED GENERIC LETTERS 

Generic Date of 

Letter No. Subject Issuance Issued To 

89 -01 IMI 

PR( 

FOI 

TE( 

IN 

COf 

TE( 

ANI 

PR( 

RE1 

CAL 

THE 

88- 20 

88-19 

88-18 

88- 17 

88- 16 

88- 15 

PLEMENTATION OF 1/31/89 ALL LICENSEES HOLDING 

3GRAMMATIC CCNTROLS OPERATING LICENSES 

? RADIOLOGICAL EFFLUENT AND CONSTRUCTION 

2HN I C AL SPEC I F I CAT I O NS 

ITROLS SECTION OF THE 

ZHNICAL SPECIFICATIONS 

1 THE RELOCATION OF 

ICEDURAL DETAILS OF 

.CULATION MANUAL OR TO 

! PROCESS CONTROL PROGRAM. 

PERMITS FOR NUCLEAR 

THE ADMINISTRATIVE POWER REACTOR FACILITIES. 

rs TO THE OFFSITE DOSE 

INDIVIDUAL PLANT 

EXAMINATION FOR SEVERE 

10 CFR 50.54(f) 

ACCIDENT VULNERABILITIES - 

US€ OF DEADLY FORCE BY 

LICENSEE GUARDS TO PREVENT 

THEFT OF SPECIAL NUCLEAR 

MATERIAL 

PLANT RECCRD STORAGE ON 

OPTICAL DISKS 

11/23/88 

10/ 28/ 88 

1 O/ 2 0/88 

LOSS OF DECAY HEAT REMOVAL 

10 CFR 50.5S(f) 

10/17/88 

REMCIVAL OF CYCLE-SPECIFIC 

PARAMETER LIMITS FROM 

TECHNICAL SPECIFICATIONS 

ELECTRIC POWER SYSTEMS - INADEQUATE CONTROL OVER 

DESIGN PROCESSES 

10/04/88 

09/12/88 

ALL L!CF' XES HOLDING 

OPERATIE.; LICENSES 

AND CONSTRUCTION 

PERMITS FOR NUCLEAR 

POWER REACTOR FACILITIES 

ALL FUEL CYCLE FACILITY 

LICENSEES WHO POSSESS , USE, IMPORT, EXPORT, 

OR TRANSPORT FORMULA 

QUANTITIES OF STRATEGIC 

SPECIAL NUCLEAR FIATERIAL 

ALL LICENSEES OF 

OPERATING REACTORS 

AND HOLDERS OF 

CONSTRUCT ION PERM ITS 

ALL HOLDERS OF 

OPERATING LICENSES 

OR CONSTRUCTION 

PERMITS FOR 

PRESSURIZEC WATER 

REACTORS 

ALL POWER REACTOR 

LICENSEES ANG 

APPLICANTS 

ALL POWER REACTOR 

LICENSEES AND 

APPLICANTS 

138 
TO: 

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, 0. C. 20555 

April 9, 1991 

ALL HOLDERS OF OPERATING LICENSES AND CONSTRUCTION PERMITS FOR 

NUCLEAR POWER REACTORS 

SUBJECT: LICENSEE COMMERCIAL-GRADE PROCUREMENT AMD GEDICATION PROGRAMS 

(GENERIC LETTER 91 -05) 

This generic letter notifies the industry of the staff's pause in conducting 

certain procurement inspection and enforcement activities and identifies a 

nuniber of failures in licensees' commercial-grade dedication programs 

identified during recerlt team inspections performed by the U.S. Nuclear 

Regulatory Commission (NX). The pause, which began in March of 1990, will 

end ill late surer Gf 1991. The purpo;e of the pause is to allow licensees 

sufficient t--?, to fully understand and irriplement guidance developed by 

industry to improve procurement and comercial-grade dedication programs. 

This generic letter expresses staff positions regardirig certain aspects of 

licensee commercial-grade procurement and dedication progranis which would 

provide acceptafile methods to meet regulatory requirements. 

During the period from 1906 to 1989, the NRC conducted 13 team inspectioris of 

the licensees' procurement and commercial-grade dedication programs. During 

these inspections, the NRC staff identified a common, programmatic deficiency 

in the licensees' control of the procurement and dedication process of 

commercial-grade items for safety-related applications. 

the staff found that licensees had failed to adequately maintain programs as 

required by 10 CFR Part 50, Appendix B, to assure the suitability of 

commercially procured arid dedicated equipment for its intended safety-related 

applications. 

quality installed in the licensees' facilities. 

In a number of cases, 

In addition, the staff identified equipment of indeterminate 

Because of a decrease in the number of qualified nuclear-grade vendors, the 

NRC staff is aware that there has been a change in the industry's procurement 

practices. Ten years ago, licensees procured major assemblies from approved 

vendors who maintained quality assurance programs pursuant to Appendix B of 

Part 50 of Title 10 of the Code of Federal Regulations (10 CFR). Currently, 

due to the reduction in the number of qualified nuclear-grade vendclrs, 

1 icensees are increasing the numbers of commercial-grade replacement parts 

that they procure and dedicate for use in safety-related applications. 

is a substantial change from the environment in which 10 CFR Part 50, 

Appendix B was promulgated. This has necessitated an increased emphasis by 

licensees and the lVRC staff to maintain procurerirent and dedication programs 

that adhere to the requirements of 10 CFR Part 50, Appendix B, and thus assure 

the quality of items purchased and installed in safety-related applications. 

Therefore, dedication processes for commercial-grade parts have increased in 

importance and NRC inspections have determined that a riumber of licensees have 

not satisfactorily performed this procurement and dedication process. 

This 

9104030126 139 
GENERIC LETTER 91-05 -2- 

The industry has been made fully aware of the NRC's concerns in this program 

area. 

affected licensees and to the industry of NRC's findings, concerns, and 

expectations in the implementation of procurement and dedication programs. 

In the past, escalated enforcement cases have provided notice to the 

Further, the NRC staff continues to participate in numerous industry meetings 

and conferences at which the NRC's positions in this area have been presented. 

The Nuclear Utility Management and Resources Counci 1 (NUMARC) Board of Directors recently approved a comprehensive procurement initiative as described in 

NUMARC 90-13, "Nuclear Procurement Program Improvements ,I' which commits 

licensees to assess their procurement programs and take specific action to 

enhance or upgrade the program if they are determined to be inadequate. 

initiative on the dedication of commercial-grade items, which is part of 

NUMARC 90-13, was to be implemented by January 1, 1990. The staff is monitoring implementation of licensee program improvements by conducting assessments 

of their procurement and commercial-grade dedication programs and maintaining 

close interaction with the nuclear industry through participation in conferences, panels, and meetings. 

The 

The staff will cmtinue to perform reactive inspections relating to plant 

specific operational events or to defective equipment and, as required, will 

continue to initiate resultant enforcement actions. In addition, the staff 

will continue to perform inspections of vendors. 

procurement and dedication inspection activities in the late summer of 1991. 

These resumed inspections will be conducted using 10 CFR Part 50, Appendix B 

(not the NUMARC initiatives) as the applicable regulatory requirement. 

Licensee programs must assure the suitability of commercially procured and 

dedicated equipment for its intended safety-related application. 

The staff position is that the staff will not initiate enforcement action in 

cases of past programmatic violations that have been adequately corrected. In 

addition. the staff does not expect licensees to review all past procurements. 

However, if during current procurement activities, licensees identify 

shortcomings in the form, fit, or function of specific vendor products, or if 

failure experience or current information on supplier adequacy indicates that a 

component may not be suitable for service, corrective actions are required for 

all such installed and stored items in accordance with Criterion XVI of 10 CFR 

Part 50, Appendix B. 

determine programnatic causes when actual deficiencies in several products from 

different vendors are identified during current procurement activities and 

these deficiencies lead to the replacement of installed items as part of the 

corrective action. In such cases, a further sampling of previously procured 

commercial-grade items may be warranted. 

In NRC Generic Letter (GL) 89-02, "Actions to Improve the Detection Of Counterfeit 

and Fraudulently Marketed Products 

good practices in procurement and dedication and provided the NRC's conditional 

The staff expects to resume 

Also in accordance with Criterion XVI, licensees must 

the staff described its perspective on 

140 
GENERIC LETTER 91- 05 -3 - 

endorsement of an industry standard (EPRI NP-5652) on methods of conmercialgrade procuremetit and dedication. 

discussed in Enclosure 1, indicate that licensees have failed to include 

certaitr key activities, as appropriate, ir! the implenierltatiotr crf the dedication 

process. The NRC staff's positions on the successful implementation of 

licensees' proyranis for conmercial-grade dedication with respect to critical 

characteristics and like-for-like replacements zire as follows. (These are also 

included in Enclosure 1.) 

A nutntrer of recent inspection findings, as 

The term "critical characteristics" is not contained in Appendix B hrrd has no 

special regulatory significance beyond its use and definition in various 

industry guides atid standards. 

characteristics in GL 89-02 as constitutirig those characteristics which need to 

be identified snd verified dbring prciduct acceptat,ce as part of the procurwent 

process. 

be considered to be critical characteristics 2s defined and used in 

EPRI HP-5652. Rather, as st6ted in Appendix By Criterion 111, licensees must 

assure the suitability of all parts, materials, and services for their intended 

safetj-related applications (i.e. , therr needs tcj be assurance thct the iteni 

will perform its intended sarety functiclr, when required). The licensee is 

responsible for itientifying the iniportatit design, material, and performance 

characteristics for each piirt, material , and service intended for safetyreldted applications, esthbl ishing acceptance criteria, and providin5 

reasonable assurance of the conformance of -items to these criteria. 

The hKC first used the term critical 

The NRC has not tcketi the psition that ti11 design requirements nust 

A like-for-like rEplacement is defined as the replacenlent of an item with an 

item that is identical. 

if it was purchased at the same time from the same vendor as the item it is 

replacing, or if the user can verif!: that there have been no changes in the 

design, materials, or iiianufacturifig process since procurmrnt of the item beins 

replaced. 

repl5cement item, then the item is not identical, but similar to the item being 

repliced, and an evaluatiori is necessary to deterniine if any changes in design, 

material , or the manufacturing process could inipact the functional 

characteristics and ultimately the cmponent's ability to perforni its required 

Safety function. If the licensee can demonstrate that the replacenlent iter;i is 

identical, then the licensee need not identify the safety function or review 

and verify the design requirements arid critical characteristics. 

irtvc,lvenient is necessary itr the above activities. Reliance on part number 

verification and certificaticn documentation is irisufficient to etisure the 

quality of commercially procured prodwts. 

For example, the replacement item would be identical 

If differences from the original iteL: are identified in the 

Engineerins 

The other matters discussed in Enclosure 1 do not ccxstitute NRC staff 

positions, but provide information on inspection findings and clarify the 

characterization of effective procurenrent and dedication programs previously 

described in GL 89-02. 

BACKFIT DISCUSSION: 

Based GI: Fast inspection findings and tht: resulting Enforcement actions, the 

f4RC staff has determinea that 1 i censee comnercial-grade procurement and 

141 
GENERIC LETTER 91-05 

aeaicction progrms needeti to be improved to complj with the existing NRC 

requirements as described in 10 CFR Part 50, Appendix B, Criterion I11 (Design 

Cofitrol) , I\/ (Procurement Document Ccintrol), VI1 (Control of Purchased 

Materia:, Equipnient and Services), and XVIII (Audits). 

have fiiled to adequately nihiritain progranis to assure the suitability of 

connicrcial ly prbcured and dedicated equipment for its intended safety-related 

application. 

iFjp7enientation of existit15 regulatcjry requirements, as contained in Appendix Ei 

to 10 CFR Part 50, the staff has concltided, that this is a compliance backfit 

ziid has prepdred the gerieric letter in accordance with 10 CFR 5G.109 (a)(4)(i). 

In light of the inadequacies identified in the procurement and dedication 

program of a large number of licensees, the issuance of this generic letter is 

nccessary tr; express the staff's position on the key element that licensees 

must incluct as part of the dedication process, specifically that commercialgrade procurement ana ckdication programs vjust cssure the suitability of 

equipment for its intecded safety-related 6pplicatiorr. This generic letter is 

also intended to clarify the elements of effective prccurement and 

coniniercial-grade dedication program that were previously provided to licensees 

in GL 69-02'. Since licensets' procurement and deGication progranis may contain 

programmatic deficiencies, the staff has itlcluded in the generic Ieitgr the 

necessary licet,see corrective actiorl to address shortcomings identified in 

specific vendor products or components that directly :ead to the ccmponent not 

being suitahle for safety-related service. 

Specificzlly, licensees 

Since the generic letter presuts staff pcsitions resiirdiny 

Although no response to this letter is required, if JOU have any questions 

regardiriy this mtter, please cctntact the persons listed below. 

Sincerely , 

n 

d-hes G. Partlow 

Office of Nuclear Reactcr Regulation 

A i sociate Director for Projects 

Enclosures: 

1. Characteristics of Effective Coamercial-Grade 

2. 

Procurement and Dedi catioti Prqrams List of Recently Issued Generic Letters 

Technical Contacts: Richard P. NcIntyre, NRR 

(301) 492-3215 

Uldi s Potapovs, NRR 

(301) 492-0959 

142 
Enclosure 1 

CHARACTERISTICS OF EFFECTIVE COMMERCIAL-GRADE 

PROCUREMENT AND DEDICATION PROGRAMS 

Background 

Appendix B to 10 CFR Part 50 contains the NRC’s regulation; for procurement 

quality assurance (QA) and quality control (QC) for products to be used in 

safety-related applications. In addition, the NRC has provided further 

guidance in Regulatory Guides 1.28, 1.33, and 1.123. 

guides, if properly implemented, provide a measure of assurance for the 

suitability of equipment, including commercial-grade items for use in 

safety-related systems. 

select and review for suitability of application materials, parts, equipment, 

and processes that are essential to the safety-related functions of the 

structures, systems, and components. Criterion IV requires that procurement 

documents specify the applicable requirements necessary to ensure functional 

performance. Criterion VI1 requires licensees to assure that the following are 

sufficient LO identify whether specification requirements for the purchased 

material and equipment have been met: 

objective evidence of quality, inspection of the source, and examination of 

products upon delivery. The process used to satisfy these requirements when 

upgrading commercial-grade items for safety-related applications is commonly 

called “dedication.” The process of ensuring compliance with 10 CFR Part 50, 

Appendix B, must include all those activities necessary to establish and 

confirm the quality and suitability of commercially procured and dedicated 

equipment for its intended safety-related application. 

activities may occur early in the procurement cycle, before ‘the item is 

accepted from the manufacturer. 

the Detection of Counterfeit and Fraudulently Marketed Products,” discussed 

commercial-grade dedication in terms of engineering involvement in the 

procurement process, product acceptance, and the dedication process as 

identified in the EPRI NP-5652 guidelines. 

the characteristics of effective procurement and dedication programs previously 

discussed in GL 89-02 and provides examples of specific failures by licensees 

to effectively implement these characteristics for dedicating and ensuring the 

suitabi 1 ity of cowmercial-grade products for safety-related applications. 

Appropriate implementation of these characteristics would have avoided many of 

the failures to meet 10 CFR Part 50, Appendix B requirements in licensee 

procurement and commercial-grade dedication programs which were identified 

These requirements and 

Criterion 111 of Appendix B requires licensees to 

source evaluation and selection, 

Some of the dedication 

Gen3ric Letter (GL) 89-02, “Actions to Improve 

This enclosure further discusses 

during past NRC inspections. 

Inspection Observations and Findings 

From 1986 to 1989, headquarters and regional personnel conducted 13 team 

inspections of licensees’ procurement and dedication programs. These inspections have identified a common, broad programmatic deficiency in licensees’ 

control over the process of procurement and dedication of commercial-grade 

143 
items. 

the suitability of equipment for USE' irl safety-related applications as 

required by 10 CFR Part 50, Appendix B, Criterion 111. These 13 inspections 

resulted in findincjs with significant sdfety impliccitions. 

eight findings that were consiticred to be Severity Level I11 violations and 

three findings that were Severity Level IV violations. At one plant, the staff 

did not assicjn a severity level to individual violations. Instead, the staff 

considered the entire group to be a Severity Level 111 probleni and used enforcement discretion, as provided under the enforcement pblicy, based on the 

licensee's corrective actions (see 10 CFR Part 2, Appendix C, Section V.G.2). 

Only one of the plants that were inspected did not receive violations in this 

program area. 

In a number of cases, licerlsees have not nrzlirltiiined programs to ensure 

The staff identified 

:n GL 89-02, tho hRC has conditionally endorsed the dedication methods 

described in EPRI NP-5652 guidelines. The staff believes that licensees who 

implenient these dedication methods, in accordance with the NRC's endorsement, 

can establish a basis for satisfying the existing requirements of Appendix E 

to 10 CFR Part 50 i;s these requirements spp'ij to the dedication process for 

commercial-grade items. 

must include prrisions to demonstrate that a dedicated iteni is sLiitable for 

safety-related tipplications. 

ity, certain key activities must be perfortlied, as appropriate, as part of the 

dedication process. 

approaches described in GL 89-02. 

An effective comniercial-grade dedication program 

For a licsnsee to acequately establish suitabilThis generic letter is intended to clarify the aeclication 

During each of the 13 inspections, the staff idmtified a cc;imon element in 

each of the inspection findings. This element was the failure of the licensee 

to assure that a commercially procured and dedicated item was suitable for the 

intended safety-related application. A dedicated commercial-grade item must 

be equivalwt in its ability to perform its intended safety function to the 

same item procured utider a 10 CFR Part 50, Appendix B QA program. The following is a list of the 13 licensees inspected and the inspection report numbers. 

A sunciclry of the general inspection findings and NRC Gbservations on these 

findings follows the list of licensee inspections. 

LICENSEE and PLANT 

1. Tennessee Vallej Authority (Sequoyah) 

2. Southern Cblifornia Edison (San Onofre) 

2. Alabama Power (Farley) 

4. Louisiana Power and Light (Waterford) 

IhSPECTiON REPORT NO. 

50-327/86-61 

50-328/86-61 

50-206/87-02 

50-361/87-03 

50-362/87-04 

50-348/b7-11 

50-364/87-i1 

50-382/87-19 

144 
-3 - 

LICENSEE and PLANT INSPECTION REPORT NO. 

5. Sacramento Municipal Utility District (Rancho Seco) 50-312188-02 

6. Maine Yankee Atomic Power (Maine Yankee) 50 -309 /88-200 

7. Northern States Power (Prairie Island) 

8. Portland General Electric (Trojan) 

50 -282/88-201 

50-306/88-201 

50-344188 -39 

50-344188-46 

9. Connecticut Yankee Atomic Power (Haddam Neck) 50-213/89-200 

10. Washington Public Power Supply System (GINP-2) 50-397189-21 

50-397189-28 

11. Florida Power (Crystal River) 50-302189-200 

12. Gulc rtates Utilities (River Bend) 50-458/89-200 

13. Commonwealth Edi son (Zion) 50-295/89-200 

50-304/89-200 

1. Inspection Findings 

a. Failure to identify the methods and acceptance criteria for verifying the critical characteristics, such as during receipt inspection, 

dedication process, or post-installation testing. 

b. Failure to establish verifiable, documented traceability of complex 

comnercial-grade i terns to their original equipment manufacturers in 

those cases where the dedication program cannot verify the critical 

characteristics. 

c. Failure to recognize that some commercial-grade items cannot be 

fully dedicated once received on site. 

tured using special processes, such as welding and heat treating. 

Dedication testing of these items as finished products would destroy 

them. For these items, licensees may need to conduct vendor surveil lances or to witness certain activities during the manufacturing 

process. 

Certain items are manufacD i scus si on 

The NRC staff has met on several occasions with NUMARC and licensee 

representatives to discuss "critical characteristics" as used in the 

context of commercial-grade procurement and dedication. The term "critical characteristics" is not contained in Appendix B and has no special 

regulatory significance beyond its use and definition in various industry 

145 
-4- 

3 

L. 

yuides and standards. 

tics in GL 89-02 as constituting those characteristics which need to be 

identified and verified during product acceptance as part of the 

procurement process. 

requirements must be considered to be critical characteristics as defined 

and used in EPRI NP-5652. Rather, as stated in Apperldix By Criterion 111, 

licensees must assure the suitability of all parts, materials, and 

services for their intended safety-related applications (i.e. , there needs 

to be assurance that the item will perform its intended safety function 

when required). The licensee is respmsible for identifying the important 

design, material , and performance characteristics for each part, niaterial, 

and service intended for safetj-related tipplications, establishing 

acceptance criteria, and providing reasonable assurance of the conformance 

of items to these criteria. There is no nrininm or maximuni number of 

critical characteristics that need to be verified. 

characteristics for an item may vary from application to applicatiorl 

depending on the desigt-i and performance requirenients unique to each 

appl icatiw. 

The NRC fjrst used the term critical characterisThe NRC has not taken the position that all design 

Further, the critical 

A licensee GiaY take different approaches for the verification of the 

critical characteristics, depending en the complexity of the item. 

niany cases, the iicensee can verify tht critical characteristics of each 

item auriiig receipt inspectior; testing. However, for a complex item 

with internal parts which receive special processing during manufacturing, 

the licensee may need to ccjnduct a swrce verification of the manufacturer during production to verify the critical characteristics 

identified as necessary for the item to perform its safety function. When 

these methvds cannot verify the critical characteristics related to 

special processes and tests, certification by the origirlal equipment 

manufactlirer may be an acceptable alternative provided dccumented, 

verified traceability to the original equipmerit manufacturer has been 

established and the purchaser has verified bj audit or survey that the 

orisinal equipment manufacturer has iiriplenented adequate quality controls 

for the -tivity being certified. 

In 

For items with critical characteristics that can be verified ,*”. 

thE rimst severe or limiting plant application, the licensee might prefer 

to identify and verify the item’s critical characteristics to qualify that 

item for a;l possible plant applications. For complex items that would be 

purchased for specific plant applications, it may be appropriate to 

address the acceptance criteria for each item individually. Engineering 

involvement is important in either method because the technical evaluation 

wi 11 identify the critical characteristics, acceptance criteria, and the 

methods to be used for verification. 

Ifispection Findings 

a. Failure to demonstrate that a like-for-like replacement iter11 is 

identical in form, fit, and function to the item it is replacing. 

Part number verification is not sufficient because Gf the probability 

of undocumented changes in the design, material , or fabrication 

of commercial-grade items using the same part number. 

146 
b. 

C. 

d. 

Fai1l;r-E to evaluatc changes iri the design, mterial, or manufacturins process for the effect of these chanses on safety function 

perforr,ii,nce (particularly under aesiyri basis event ccilai tions) of 

rep1,iccenient items that are sirdar as opposed to identical to the 

itenis being replkced. 

Failure to enxrt that items will function under all design requirements. 

grade i tern would fknction brlcer normal clpwation cmditions. 

On some occasicns, 1 icensees mly ensured thh'i the commerciclFai lure to verify the validity of certificates of curiforinaiice 

received from vendors not on the licensee's list of approved vendors/ 

suppliers. 

grade vendor is riot sufficiwt. 

An uiiverified certificate of conformafice from a comercialZiscussion 

k like-for-like replacement is dcfined as tlic replacement of an iten with 

iil: item that is identica;. For example, the replacement item wc;uld be 

identir.21 if it was purihased at the same time fron: the silrile vendor as tht 

item 

changes in the design, matrrials, or clanufacturirlg process since 

procurenierit of the item being replaced. iter.1 tre identified in the replacemeiit item, then the item is i&t 

identicG1, but similar to the item being replaced, and evaluation is 

necessary to drtermint: if any changes irl design, material, Gr the 

nlanufacturing process could inipact the functional Lharacteristics and 

ultiniztely the cciirponent's cbility to pcrform its required safexy functiw 

identical, then the licensee need not identify the safety 'functiou or 

revitw and verifj the design requirements and critical characteristics. 

t is replacing, or if tne user can verifj that there have been nc 

11: differences from the Griginal 

If the licensee can aenicnstrate thtt the replccement item is 

Encjineering involvement is necessary iri the abGve activities. 

uf this iilvolN:einent is dependelit or the nature, coiriplexit\*- and use of 

thc items to be dedicated. ParticipEtion of engirmring persotinel is 

apprcpriate in the procurerileiit process, irnd product acceptance, tcj 

deve:op purchase specifichticns, determine specific testing requirements 

itpplicable to the prodl;cts, and evoluate the test results. 

neering persuinel specify design requirements for inclusion on the 

pl;rchase docunients for replacement components, they need riot reconstruct 

and reverify design adequacy for procurement purposes, but need only 

ensure that the existins design requirements (which may reference the 

original design basis) are properly translated into the purchase order. 

Reliance 011 part iicnber verificdtion and certification documentation is 

insufficient to ensure the quolity of commerciallj procured prGducts. 

Effective prc;cluct acceptance progrms have as tleriients , receipt an6 sburce 

inspection, approprihte testiris criteria, effective vendor audits aid 

survei 1 lances (including witness/hold pcirrts as appropriate), special 

tests and inspections, and post-instal latiori tests. 

adequate qLilificaticr,s ana trhiiiirig for implementing personnel are alsr, 

necessary factors in successful iiiip1enientatiCrri. 

The extent 

Nhen engiProcedures and 

147 
LIST OF RECENTLY ISSUED GENERIC LETTERS Lficl(3sul’e L 

Gener i c Date of 

Letter No. Subject Issuance Issued To 

VEILLANCE INTERVALS TO ACCOMMODATE A 24- 

MONTH FUEL CYCLE (GENERIC LETTER 91-04) 

1-04 CHAm IN Tt CHmrCAL SF -Trrm6tRs - OF OL 

OR CONSTRUCTION PERMITS FOR NUCLEAR 

POWER RFACTORS 

91-03 

9 1-02 

91-01 

90-09 

89-10 

SUPP. 3 

90-08 

90-07 

89-10 

SUPP. 2 

REPORTING OF SAFEGUARDS 

EVENTS 

03/06/91 

REPORTING MISHAPS INVOLVING 12/28/90 

LLW FORMS PREPARED FOR 

DISPOSAL 

REMOVAL OF THE SCHEDULE FOR 

THE WITHDRAWAL OF REACTOR 

VESSEL MATERIAL SPECIMENS 

FROM TECHNICAL SPECIFICATIONS 

01/04/91 

ALTERNATIVE REQUIREMENTS FOR 12/11/90 

SNUBBER VISUAL INSPECTION 

INTERVALS AND CORRECTIVE 

ACTIONS 

CONSIDERATION OF THE RESULTS 10/25/90 

OF NRC-SPONSORED TESTS OF 

MOTOR-OPERATED VALVES 

SIMULATION FACILITY 

EXEMPTIONS 

08/ 10/90 

OPERATOR LICENSING NATIONAL 08/10/90 

EXAMINATION SCHEDULE 

AVAILABILITY OF PROGRAM 

DESCRIPTIONS 

148 

08/03 /90 

ALL HOLDERS OF OLs 

OR CPs FOR NUCLEAR 

POWER REACTORS AND 

ALL OTHER LICENSED 

ACT1 V IT1 ES INVOLV I NG 

A FORMULA QUANTITY 

OF SPECIAL NUCLEAR 

MATERIAL (SNM) 

ALL OPERATORS OF 

ACTIVE WASTE (LLW) 

DISPOSAL SITES 

WASTE PROCESSORSy 

& ALL HOLDERS OF 

LICENSES FOR NUCLEAR 

FUELS, NUCLEAR 

MATERIALS & NUCLEAR 

POWER REACTORS 

LOW-LEVEL RADIOALL HOLDERS OF OLs 

OR CPs FOR NUCLEAR 

POWER PLANTS 

ALL LIGHT-WATER 

REACTOR LICENSEES 

AND APPLICANTS 

ALL LICEPSEES 3F 

OPERATING NUCLEAR 

POWER PLANTS AND 

TION PERMITS FOR 

NUCLEAR POWER PLANTS 

HOLDERS OF CONSTRUCALL HOLDERS OF 

OPERATING LICENSES 

OR CONSTRUCTION 

PERMITS FOR NUCLEAR 

POWER REACTORS 

ALL POWER- REACTOR 

LICENSEES AND 

APPLICANTS FOR AN 

OPE RAT I NG L I CENSE 

ALL LICENSEES OF 

OPERATING NPPs AND 

HOLDERS OF CPs FOR 

NPPs 
. . . ,., , ,,

.. , .. .. _---,_,-.-.I . -- - 

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-ooO1 

NRC INSPECTION MANUAL RVIB 

INSPECTION PROCEDURE 38703 

COMMERCIAL GRADE DEDICATION 

PROGRAM APPLICABILITY: 2515 

SALP FUNCTIONAL AREA: ENGINEERING (SOETS-0) 

38703-0 1 INSPECT ION OBJECTIVES 

01.01 To determine whether the failure of a safety-related system, ;tructure, 

component (SSC), or part to perform its intended safety function was the result 

of a deficient comnercial grade item (CGI) dedication process. 

01.02 To verify that the licensee's process for dedicating CGIs, as implemented, 

meets the applicable portions of Appendix B to 10 CFR Part 50 and provides 

reasonable assurance that CGIs will perform their intended safety function. 

38703-02 INSPECTION REQUIREMENTS 

02.01 Reactive InsDection Reauirements 

a. Jnitial Evaluation. After reviewing the licensee's evaluation of the 

failed item, determine if the failed item was procured as a CGI and 

dedicated for safety-related applications. If the failed item was 

dedicated, review the complete procurement and dedication records to 

determine if the comnercial grade dedication process was sufficiently thorough. 

b. Further Assessments. If it is determined that the dedicated item failed 

as the result of certain critical characteristics not being identified and/or properly accepted, the inspector should perform the fol 1 owing 

assessments: 

1. Determine if other CGIs from the same accepted lot or batch as the 

failed dedicated CGI have been similarly dedicated and installed in 

other safety-related applications. If yes, determine if the 

licensee has evaluated the operability of the systems or components 

where these CGIs are installed. The inspector also should review 

1 icensee-provided data, if avail able, for some CGIs (non-dedicated) 

that failed in applications that were not safety-related. Explore 

the possibility that the same CGIs also may have been used 

(following dedication) in a safety-related application and may have 

the potential to affect the safe operation of a SSC. 

Issue Date: 11/08/93 -1- 38983 

149 
2. If possible select and evaluate, as in step 1 above, at least three 

other dedicated CGIs having similar applications and critical 

characteristics as the CGI(s) that resulted in the identified 

fai 1 ures. 

3. If, after performing step 2 above, it is determined that there were 

weaknesses in the commercial grade dedication processp the inspector 

should perform a more comprehensive inspection of the ’licensee’s 

dedication process in accordance with the inspection requirements in 

Section 02.02 below. 

02.02 Proaramnatic InsDection Reauirements 

a. Review of Proaram and Proc edures. Using the inspection guidance 

contained in Section 03.02 and Appendix A to this procedure, review the 

1 icensee’s program and procedures for the procurement and dedication of 

CGIs in order to understand the basic operation of the licensee’s 

program. 

b. 

C. 

d. 

Selection of Dedication Packaaes. Select approximately 20 dedication 

packages for evaluation from a list of comnercially dedicated items 

provided by the licensee. Request that the licensee provide (or make 

available for review) a complete package of the pertinent procurement and 

dedication records for each item. 

Evaluation of Dedication Packaaes. Using the inspection guidance 

contained in Section 03.01 of this procedure, perform a detailed 

evaluation of the dedication packages selected in item b above. 

Evaluation of Trainina Effectiveness. If the inspector’s evaluation of 

comnercial grade dedication activities indicates there are weaknesses in 

the way these activities are being performed, the inspector should 

investigate further to determine if weaknesses within the licensee’s 

training program may have contributed to the cause. The inspector should 

determine if the licensee is implementing an effective training program. 

38703-03 INSPECTION GUIDANCE 

GENERAL GUID ANC 

Backaround. Licensees are required to ensure the qual ity of items purchased and 

installed in safety-related applications. In the past, 1 icensees procured major 

assembl ies from approved vendors who maintained quality assurance (QA) programs 

pursuant to Appendix B to 10 CFR Bart 50. Because of the decrease in the number 

of qual i fi ed nucl ear-grade vendors, 1 i censees are increasing the numbers of 

commercial grade replacement parts that they procure and dedicate for use in 

safety-related applications. 

Since commercial grade dedications have increased in number, the Nuclear 

Wegul atory Commission (WWC) has deweloped this inspection procedure to provide 

guidance to assist the inspector in assessing the effectiveness of the 

implementation of the licensee’s commercial grade procurement practices and 

provide for early identification of any adverse trends or emerging problems. 

?he Vendor Inspection Branch, of the WRC’s Office of Nuclear Reactor Regulation, 

is available to assist with specific questiorts that arise during the performance 

of this procedure. 

38703 -2- 

150 

Issue Date: 11/08/93 
Schedul inQ the InsDection. This inspection procedure should be considered for 

implementation when there is reason to believe that the failure of a SSC or part 

to perform its intended safety function was the result of weaknesses in CGI 

dedication. This inspection procedure may be implemented independently or it may 

be used as a supplement to other major team inspections. Such inspections may 

include maintenance, modification, or system-specific inspections where review 

of failed SSCs or parts is appropriate, or an augmented inspection team 

investigating fai 1 ures. 

The NRC should contact the appropriate NRC and licensee personnel to schedule the 

inspection. When practical, inform the licensee of the objectives of the 

inspection 4-6 weeks before the inspection is to begin and advise them of 

information that will be needed, such as a list of items that the licensee 

purchased as comnercial grade after July 1990 and subsequently dedicated for use 

in safety-related applications. Before the beginning of the onsite inspection, 

the inspector should request and review the licensee’s program and procedures to 

become familiar with the licensee’s procurement and dedication process. Also 

explore with the licensee the possibility of obtaining a list of recent component 

failures. Request this list only if the licensee states this type of information 

would be easily retrievable. The list of component failures can be used by the 

inspector during the selection of dedication packages for review described in 

Section 02.02 of this inspection procedure. 

This inspection procedure is consistent with the Nuclear Management and Resources 

Council (NUMARC) initiative for improving the utilization of CGIs in nuclear 

safety-related applications that was implemented in July 1990. The methods used 

to commercially dedicate items procured by licensees before that date will not 

necessarily meet the guidance contained in this inspection procedure. If the 

inspector encounters a significant failure of a comnercially dedicated item, 

which was dedicated before July 1990, the inspector may review the dedication of 

that item with the understanding that the licensee was not expected to meet the 

current guidelines. 

SPEC IF IC GUIDANCE 

03.01 Pea ctive Inspection 

a Jnitial Evaluation. A failure resulting from general weaknesses in the 

commercial grade dedication program may occur when the important design, 

material, and performance characteristics that are necessary to provide 

reasonable assurance that the dedicated CGI will perform its intended 

safety function are not addressed during dedication. For example, 

failures of safety-related bolting have occurred when the dedication 

process did not verify that the material composition and/or mechanical 

properties met the specified requirements and nonconforming material was 

suppl i ed . 

Review and discuss with 1 icensee personnel the fail ure/root-cause 

analysis when required or applicable for the failed CGI. The inspector 

should attempt to determine if the failure was due to a design 

deficiency, failure unrelated to the item’s safety function, or normal 

wear, and eliminate these from further review. The inspector should 

focus on the inspection of failures that appear to Be due to weaknesses 

in the commercial grade dedication process. If none of the failures are 

due to weaknesses in the commercial grade dedication process, then the 

inspector should not continue using this inspection procedure. If the 

inspector decides to change the focus of the insTection to examine other 

issues related to the failures, such as the adequacy of corrective 

Issue Date: 11/06/93 -3- 38703 

151 
actions, other procedures should be used, such as NRC Inspection 

Procedure 92720, "Corrective Action." Once the failure mode and cause of 

failure have been postulated or determined, review the dedication package 

as described in Section 03.01a(l) to determine if appropriate critical 

characteristics had been identified by the licensee. Appendix A to this 

inspection procedure should not be interpreted as inspection requirements 

but only as a discussion of dedication issues including guidance on 

selection and verification of critical characteristics. Appendix A, if 

properly imp1 emented, represents an acceptable means of complying with 

regulatory requirements. Individual 1 icensee's may develop a1 ternate 

methods of achieving Appendix B compliance. Appendix B provides 

definitions of terms used for commercial grade dedication activities, and 

Appendix C provides the typical contents of a dedication package. 

The goal of the review of the dedication packages Is to provide 

reasonable assurance that the CGIs dedicated for safety-re1 ated 

applications will perform their intended safety functions. Inspection 

effort should be directed towards the identification of weaknesses in the 

dedication process that could potentially render SSCs or parts 

inoperable. When reviewing 1 icensee's operability determinations for 

dedication of CGIs, the inspector should refer to the "Technical 

Guidance" section of NRC Inspection Manual, Part 9900, for further 

guidance. 

all) Review of Dedication Packaqes. After becoming familiar with the 

1 i censee's procurement and dedi cation program and procedures, perform a 

detailed review of the dedication package as described below. 

e 

e 

e 

0 

Determine if the safety function of the item for its intended use 

has been identified by reviewing the documents associated with the 

technical evaluation including, as applicable: 

- classification of the item 

- consideration of credible failure modes - item equivalency/substitution evaluations 

Determine if the important design, material, and performance 

characteristics relevant to the safety function have been 

identified. Determine whether the licensee verified the 

characteristics necessary to provide reasonable assurance that the 

item will perform its intended safety function. If appropriate, 

take into account post-installation testing and periodic 

surveillance testing and inspection. Review the basis for 

engineering judgment when it is used as the basis for selection or 

verification of critical characteristics. 

Determine whether the item is an equivalent replacement or a new 

item replacement of an obsolete item. 

Determine if the item is or may be used in a different safetyrelated application than previously evaluated in which different 

design, material, and performance characteristics may be applicable. 

This is especially applicable for generic dedications of bulk items 

and stock material. Determine if the dedication ensures those 

design, material, and performance characteristics relevant to the 

safety function. 

38703 -4- 

152 

Issue Date: 11/08/93 
Determine why the item is being replaced. Have there been repeated 

failures? Is the degraded performance a result of adverse 

environment? Did it fail because it was a refurbished or fraudulent 

item? General information on similar activities subject to Appendix 

B to 10 CFR Part 50 is provided in American National Standards 

Institute (ANSI) ANSI N45.2-1977, "Quality Assurance Program 

Requirements for Nuclear Power Plants, " Section 17, "Corrective 

Act i on. " 

Determine how the identity of the item is controlled from the time 

it is receipt inspected until the time it is installed. General 

information on similar activities subject to Appendix B to 10 CFR 

Part 50 is provided in ANSI N45.2-1977, Section 9, "Control of Parts 

and Components. " 

Determine if information learned during the dedication process is 

fed back to the appropriate persons to evaluate existing stock 

items, or installed items, and for future use in surveys and source 

verifications. This information could include positive and adverse 

findings obtained during surveys and source verifications. General 

information on similar activities subject to Appendix B to 10 CFR 

Part 50 is provided in ANSI N45.2.13-1976, "Quality Assurance 

Requirements for Control of Procurement of Items and E=rvices for 

Nuclear Power Plants," Section 9, "Corrective Action." 

Refer to the discussion of significant dedication issues in Appendix A 

for guidance during the review of dedication packages. Also refer to the 

specific guidance for each of the four dedication methods provided below. 

Focus should be on those activities that are likely to affect the 

performance of the items being dedicated. It is not necessary to review 

the licensee's programmatic compliance to the 18 criteria of Appendix B 

to 10 CFR Part 50 as they may not apply to the activities reviewed. 

Appendix B to 10 CFR 50 does not apply to commercial grade activities 

which occur prior to dedication for use in a safety-related system. It 

also should be recognized that this appendix provides for the application 

of QA to safety-re1 ated systems and components consistent with their 

importance to safety (graded qual i ty approach). 

A1 though guidance concerning the appl ication of graded qual i ty assurance 

is discussed in the first paragraph of Appendix A to this inspection 

procedure, it is expected that the inspector will need to exercise 

considerable judgment in determining the adequacy of controls applied to 

a specific activity. 

The following are the four acceptance methods that can be used to accept 

CGIs. These methods provide, either individually or in combination, a 

means to reasonably ensure that a CGI that is received meets the 

requirements of the item specified. The results of employing each method 

should be documented. 

Method 1 - Special Tests and InsDections 

General information on similar activities subject to Appendix B to 10 CFR 

Part 50 is provided in ANSI Nt5.2.13-1976, Section 10, "Acceptance of 

Item or Service." Use the following approach to review packages that 

were dedicated using this method: 

Issue Date: 11/08/93 -5- 

153 

38703 
To the extent practicable, attempt to witness receipt inspections 

and tests of in-process dedication of CGIs that are similar to that 

of the fafled item to verify the identified critical 

characteristics. 

0 Review receiving records and associated tests and inspections. 

* Review post-instal 1 at i on test records. 

Verify that the tests and inspections specified for acceptance 

Verify that sampling plans are controlled and have adequate 

technical basis, considering lot traceab 11 ity and homogeneity, 

complexity of the item, and adequacy of supplier controls. 

adequately verify the identified critical characteristics. 

Verify that CGI receiving inspection activities are adequately 

controlled under a quality program regardless of whether they are 

being performed in conjunction with other plant receipt inspection 

act i vi t i es . e Verify that receipt inspection activities establish and maintain 

traceability of CGIs by capturing and appropriately relating 

traceability documents through identification and monitoring of 

CGIs. 

Verify that measuring and test equipment was properly calibrated, 

that approved vendors were used to perform tests, and that personnel 

were qualified to perform the tests. 

Method 2 - Commercial Grade Survey 

Use the following guidance to review packages that were dedicated using 

this method: 

Determine if the guidance of Generic Letter 89-02, or an appropriate 

alternate, is included in the appropriate procedures. Specifically, 

confirm that (1) the documented commercial quality program was 

effectively implemented and (2) the surveys were conducted at the 

1 ocat i on necessary to verify that adequate control s were exerci sed 

on distributors as we1 1 as manufacturers. 

Through interview, determine if the persons who perform vendor 

surveys are knowledgeable in the following: 

- 

- screening third-party surveys 

- 

the use of performance-based surveys 

processing and evaluating adverse findings resulting from the 

review of third-party surveys to ascertain if those findings 

affect CGIs already installed or stored in the warehouse 

awaiting future instal 1 at i on 

General information on similar activities subject to Appendix B to 10 CFR 

Part 50 is provided in ANSI N45.2.12-1977, "Requirements for Auditing of 

Qual i ty Assurance Programs for Wucl ear Power P1 ants, " and ANSI N45.2.23- 

1978, "Qualification of Quality Assurance Program Audit Personnel for 

Nuclear Power P1 ants. " 

38783 -4- Issue Date: 11/08/93 

154 
e 

e 

e 

e 

e 

e 

Verify that the supplier's comnercial quality controls are imposed 

i n the procurement documen t s . 

Determine if the critical characteristics that are to be verified by 

the survey team are accurately and completely incorporated in the 

survey plans. 

Determine if the validity of supplier documentation, relied on in 

the dedication of the item, is verified during the survey. 

Determi ne if surveys of comnerci a1 grade suppl i ers are performance 

based as opposed to programnatic. Specifically, verify that the 

critical characteristics for the CGIs being surveyed are controlled 

by the supplier's quality activities. 

Determine if survey teams include technical and qual i ty personnel, 

as appropriate, that are knowledgeable in the operation of the 

item(s) and the associated critical characteristics to be verified, 

including any special processes such as welding and heat treatment 

that are specific to the critical characteristics. 

Determine if surveys are conducted at appropriate times relative to 

the procurement. Are surveys required to be updated on a regular 

basis to support dedication? 

Determine if the control of subvendors is adequately addressed by 

the surveys so that the supplier has an adequate basis to accept 

test results and certifications from the subvendor. 

Determine if pertinent information about a supplier or its products 

is used to plan, conduct, and report results of surveys and source 

verifications. Such information could have been available from 

source verifications, receiving inspections, the dedication process, 

suppl ier/product performance history, or outside sources such as NRC 

information notices and bull et i ns , nucl ear pl ant re1 i abi 1 i ty data 

system reports, or Nuclear Uti1 ity Procurement Issues Comnittee 

(NUPIC) comnercial grade survey reports. 

Method 3 - Source Verification 

General information on similar activities subject to Appendix B to 10 CFR 

Part 50 is provided in ANSI N45.2.13-1976, Section 10.3.2, "Acceptance by 

Source Verification." Use the following approach to review packages that 

were dedicated using this method: 

e 

e 

Determine if source verifications involve witnessing the supplier 

performing quality activities on the actual items being procured and 

adequately verify the item's critical characteristics. 

Determi ne i f personnel who participated in the source veri f i cat i on 

surveys were qualified for their specific assignment. 

Determine if appropriate hold points are imposed in the purchase 

orders. This would include a hold point to verify design, material, 

and performance characteristics relevant to the safety function 

that cannot be verified after the item has been completely 

manufactured. 

Issue Date: 11/08/93 -7- 

155 

38703 
Determine if the results of the source verifications were adequately 

documented. 

Method 4 - AcceDtable SUDD~ ier/Item Performance Record 

Use the following guidance to review packages that were dedicated using 

this method: 

e 

e 

e 

Determine if the guidance of Generic Letter 89-02, or an appropriate 

alternate, has been incorporated. Specifically, (1) the established 

historical record is based on industry-wide performance data that is 

directly applicable to the item’s critical characteristics and the 

intended safety-re1 ated application and (2) the manufacturer’s 

measures for the control of design, process, and material changes 

have been adequately implemented as verified by survey (multi1 i censee team surveys are acceptabl e). 

Determine if information pertinent to the CGI’s quality of 

performance, obtained from outside sources (e.g., operational event 

reports, NRC, vendor equipment and technical information program, 

and Institute of Nuclear Power Operations) and from comnercial grade 

surveys, source veri f i cat i ons , recei pt i nspecti or - , previous 

dedication or qualification and operational history, is factored 

into the dedication process. 

Determine if the item or manufacturer is included in the licensee’s 

performance trending program. 

b. Further Assessments 

1. No inspection guidance. 

2. From the list of dedicated items provided by the licensee, the 

inspector should select for review approximately three other dedication 

packages having similar applications and critical characteristics as the 

CGI(s) that resulted in the identified failures. After the selections 

have been made, the inspector should request that the licensee compile a 

complete package cf all the procurement and dedication records for each 

item. Typical contents of a dedication package are described in Appendix 

C of this inspection procedure. The inspector should review the 

dedication packages as described in Section 03.01a(l) of this inspection 

procedure. 

3. No inspection guidance. 

03.02 Prowamnatic InsDection 

a. Review of Proqram and Procedures. The review of the program and 

procedures should be performed to familiarize the inspector with the 

licensee’s CGI dedication process. For cases in which problems are 

identified with the 1 icensee’s CGI dedication process, the inspector may 

decide to perform a more extensive review of the program and procedures 

to determine if these problems are the result of inadequate procedures. 

The inspector’s review should include procedures that control : 

procurement activities; material control; the dedication of CGIs, 

including receipt inspection and acceptance testing; surveys of 

38703 -8- 

156 

Issue Date: 11/08/93 
comnercial grade suppliers; classification of components; training of 

personnel ; trending of suppl i er performance; and equipment fai 1 ures. 

Attempt to identify any apparent weak areas to concentrate on during the 

evaluation of the program implementation. 

After arriving onsite, the inspector should request that the 1 icensee 

explain its commercial grade dedication process and conduct a walkthrough 

of areas associated with it. Areas in the walkthrough could include the 

engineering, receipt inspection, component testing, and warehouse. The 

inspector should become familiar with key licensee personnel involved in 

the comnerci a1 grade dedication process. These key personnel should 

include the responsible engineer(s) who developed the dedication 

package(s) and systems engineers, procurement engineers, receipt 

inspectors, qual i ty assurance engi neers and inspectors, and warehouse 

personnel. The inspector should discuss the commercial grade dedication 

process with these key personnel to gain a better understanding of the 

process, including: 

How processing of CGI procurement documents is controlled under the 

quality program and how they receive review and approval. General 

information on similar activities subject to Appendix B to 10 CFR 

Part 50 is provided in ANSI N45.2.13-1976, Section 3, "Procurement 

Document Preparation, Review, and Change Control. " 

flow technical personnel participate in the preparation, review, and 

approval process of procurement documents. 

How consistency and coordination is maintained between corporate 1 evel , engineering/support 1 evel , and site 1 evel programs and 

imp1 ementing procedures. 

b. Selection of Dedication Packases. As discussed in the general guidance 

section above, the NUMARC initiative for the utilization of CGIs in 

nuclear safety-related applications was not implemented until July 1990. 

Therefore, the methods used to perform commercial grade dedication of 

items procured or dedicated by licensees before that date will not 

necessarily meet the guidance contained in this inspection procedure. 

The selection process should be performance oriented (e.g., weighted 

toward the review of dedication packages for equipment, components, or 

parts that have experienced failures). To accomplish this, the inspector 

should request from the licensee approximately 20 packages for review 

using the two-step approach described below. The licensee should be 

given sufficient lead time to prepare the 20 packages and make them 

available for the first day of onsite inspection. 

Step 1: Review the licensee's records available at the plant site to 

identify recent failures (approximately the last 2 years) of equipment, 

components, or parts. Review these failures to determine if any were 

CGIs dedicated for use in safety-related applications. If available, 

select approximately 75 percent of the total sample from CGI failures. 

Step 2: From the list of dedication packages supplied by the licensee, 

under the "Inspection Guidance" section of this procedure, select the 

remainder of packages for review. The total sample size including 

packages from steps 1 and 2 should be approximately 20 packages. 

However, the inspector can select a larger or smaller sample depending on 

the complexity of the packages and the time available. The inspector 

Issue Date: 11/08/93 -9- 

157 

38703 
C. 

d. 

38703 

should select these packages on the 

considerations: 

e 

f) 

basis of the following 

The inspector should select packages for items whose failure would 

have the most effect on the abil ity of the plant to safely operate, 

safely shutdown from an adverse condition, or maintain a safe 

shutdown condition. If time permits, review the plant-specific 

probabilistic risk assessment, indiwidual plant examination, and 

risk-based inspection guides that provide information on the risk 

significance of safety-related plant equipment. 

The inspe 

sel ect i on 

that have 

ctor should take a performance oriented approach to the 

process by incl’uding in the sample packages those items 

been problems ,in the past. Revidw available sources of 

information to identify any known failures of CGIs that were used in 

safety-re1 ated appl i cati ons . These sources of information coul d 

i ncl ude : 

component fai 1 ure 1 i sts or 1 i sts of i tems requiring frequent 

maintenance or 

misrepresented or Praudul ent 1 tems reported in NRC information 

notices 

licensee trending of equipment and supplier performance 

previous history of component failures or malfunctions as 

reported in 1 icensee event reports or plant nonconformance 

reports 

replacement as provided by the licensee 

The inspector should include both simple and complex packages in the 

sample as well as packages that include a variety of dedication 

methods (e.g., Methods 1 through 4) described in Section 03.01a(l) 

above. 

In addition to selecting packages based on the above considerations 

(safety significance, complexity, and failures), the inspector 

should attempt to select samples from each of the following areas: 

electrical , instrumentation and control, mechanical equipment, and 

materi a1 s. 

Svaluatix :? Dedication Packaaes. Perform a detailed review of the 

dedication packages as described above in Section 03.01a(l). 

Evaluation of Trainina Effectiveness. Experience gained during the 

procurement assessments and pilot inspections suggested that training of 

personnel involved in CGI dedication activities was a very important 

factor in the development of a good CGI dedication program. The CGI 

dedication process generally requires more highly qual ified/trained 

personnel than specified in Appendix B to 10 CFR Part 50 procurement. 

Personnel involved In this process need to be familiar with current 

industry and NRC guidance and have a strong interface with the licensee’s 

design/engi neering organizations. The training expectati ons, however, 

should not exceed what is required by the existing licensee’s QA program. 

As applicable to their job function, select and review the training 

records for individuals involved in the following areas: 

e Determining the safety classification of an item. Training in this 

area is appropriate when the job function includes reclassification 

- 10 - Issue Date: 11/08/93 
e 

e 

e 

of items or establishing safety classification of piece parts of 

safety-re1 ated components. 

Specifying design, material, and performance characteristics 

relevant to the safety function and establishing the acceptance 

criteria for these characteristics. 

Specifying or performing comerci a1 grade surveys, source 

verifications, and tests and inspections, including enhanced postreceipt verification testing or inspection. 

The preparation and review of procurement documents. 

Through observation, interviews, and a review of records of work 

performed by the individuals: 

e 

e 

e 

Determine if the individuals selected have adequate knowledge to 

perform the specific tasks assigned to them. Attend a training 

course, if available, or review the lesson plans for selected 

training courses. 

Determine if training inadequacies contributed to any of the 

deficiencies that may be identified during the inspection. 

Determine if the personnel are familiar with the program 

requirements and procedures and if they have been properly trained 

in the dedication process. 

It should be noted that alternatives to a formal training program may be 

adequate to ensure satisfactory program implementation (e.g.., on the job 

training). Additional information in this area is provided in NRC 

Inspection Procedure 41500, "Training and Qualification Effectiveness. " 

38703-04 INSPECTION RESOURCE ESTIMATE 

The estimated number of onsite inspection hours required to complete all 

inspection requirements is 144 hours when both the reactive and programmatic 

options are implemented. This estimate is for broad resource planning and is not 

intended as a quota or standard for judging inspector or regional performance. 

The on-site hours can be expected to vary significantly depending on the specific 

circumstance and scope of each inspection. 

38703-05 REFERENCES 

The following documents are listed for the inspector's information only and are 

not considered regulatory requirements unless the 1 icensee has formally committed 

to implementing any of these documents for application to safety-related 

activities. The inspector may wish to review these documents to become familiar 

with comercial grade dedication issues. 

ANSI N45.2-1977, "Quality Assurance Program Requirements for Nuclear Power 

Plants," as endorsed by NRC Regulatory Guide 1.28, Revision 2. 

ANSI N45.2.13-1976, "Quality Assurance Requirements for Control of Procurement 

of Items and Services for Nuclear Power Plarlts," as endorsed by NRC Regulatory 

Guide 1.123, Revision 1. 

Issue Date: 11/08/93 - 11 - 

159 

38703 
Electric Power Research Institute (EPRI) NP-5652, "Guide1 ines for the Uti1 ization 

of Comnercial-grade Items in Nuclear Safety Related Applications (NCIG-O7)," as 

conditionally endorsed in NRC Generic Letter 89-02. 

Generic Letter 89-02, "Actions to Improve the Detection of Counterfeit and 

Fraudulently Marketed Products" (microfiche 48960-001). 

Generic Letter 91-05, "Licensee Comnercial Grade Procurement and Dedication 

Programs" (microfiche 57468-264). 

NRC Inspection Procedure 41500, "Training and Qual ffication Effectiveness." 

SECY-90-304, "NUMARC Initiatives on Procurement" (microfiche 55277-049). 

SECY-91-291, "Status of NRC's Procurement Assessments and Resumption of 

Programnatic Inspection Activity" (microfiche 59490-079). 

END 

Appendices: 

A. Dedication Issues 

B. Definitions 

C. Contents of Dedication Packages 

38703 - 12 - 

160 

Issue Date: 11/08/93 
APPENDIX A 

DEDI CAT ION I SUES 

BASIS FOR THE SELECTION AND VERIFICATION OF CRITICAL CHARACTERISTICS 

1. 

2. 

3. 

4. 

Consideration of Item’s Safetv Function 

Critical characteristics should be based on the item’s safety function. 

The licensee is responsible for (a) identifying the important design, 

material, and performance characteristics that have a direct effect on 

the item’s ability to accomplish its intended safety function and (b) 

selecting from these characteristics a set of critical (or acceptance) 

characteristics that, once verified, will provide reasonable assurance 

that the item will perform its intended safety function. Criterion I1 of 

Appendix B to 10 CFR Part 50 provides for the application of quality 

assurance over activities affecting the quality of structures, systems, and components to an extent consistent with their importance to safety. 

This graded quality approach can be used in the selection and 

verification of critical characteristics for comnercial grade items 

(CGIs). When an existing eqyipment specification is avat;able that 

contains adequate technical requirements for the i tern being purchased, 

that specification can be used to select the critical characteristics for 

this item. 

Graded 0 ualitv Assurance 

The application of graded quality assurance to the CGI dedication process 

should include consideration of the item’s importance to safety and other 

factors specific to the item being procured. Certain items and services 

may require extensive controls throughout all stages of development while 

others may require only a limited quality assurance involvement in 

selected phases of development. The following factors should be 

considered in determining the extent of quality assurance to be applied: 

(a) The importance of malfunction or failure of the item to plant safety, 

(b) the complexity or uniqueness of the item, (c) the need for special 

controls and surveillance over process and equipment, (d) the degree to 

which functional compliance can be demonstrated by inspection and test, 

and (e) the quality history and degree of standardization of the item. 

Additional guidance on the use of graded quality assurance can be found 

in the nonmandatory appendix to ANSI N45.2.13-1976. 

Consideration of Failure Modes 

An evaluation of credible failure modes of an item in its operating 

environment and the effects of these failure modes on the item’s safety 

function may be used in the safety classification of an item and as a 

basis for the selection of critical characteristics. 

Reasonable Assurance 

The dedi cation process represents an acceptable method of achieving 

compliance with Appendix B to 10 CFR Bart 50 with the purchaser assuming 

many of the responsibilities for ensuring quality and functionality of an 

item that had previously been the responsibility of the vendor. In this 

context, reasonable assurance consists of the purchaser controlling or 

Issue Date: 11/08/93 A- 1 

161 

38703, Appendix A 
verifying the activities affecting the item’s quality to an extent 

consistent with the item’s importance to safety or ensuring that these activities are adequately controlled by the supplier. For more complex 

items, dialogue with the original equipment manufacturer may be necessary 

to identify the design and functional parameters of specific piece parts. 

Once the dedication process is completed, the quality assurance and/or 

other measures applied to those aspects of the item that directly affect 

its safety function should result in the same level of performance as for 

a like item manufactured or purchased under a quality assurance program 

of Appendix B to 10 CFR Part 50. 

5. Enuineerinu Judament 

Engineering judgment can be used in selecting ‘hose important design, 

material, and performance characteristics that are identified as the 

item’s critical characteristics. The bases for engineering judgment for 

this application should be documented. 

TRACEABILITY 

Haterial/Items Purchased From Distributor2 

Traceability can be defined as the ability to verify the history, location, or 

application of an item by means of recorded identification. Where the item’s 

acceptance is based entirely or partially on a certification by the manufacturer, 

the traceabil ity must extend to the manufacturer. The purchaser should ensure 

by survey or by other means that the manufacturer has established adequate 

traceability controls and that these controls are effectively implemented. For 

situations in which intermediaries (distributors) are included in the supply 

chain, the activities of these organizations may need to be surveyed to ensure 

that traceability and proper storage conditions are maintained. A survey of the 

distributor mzy not be necessary if the distributor acts only as a broker and 

does not warehouse or repackage the items or in cases where traceability can be 

established by other means such as verification of the manufacturer’s markings 

or shipping records. 

SAMPLING 

1. Established Heat Traceabilitv (Materials1 

When heat traceability of metallic material has been established and each 

piece of the material is identified with the material heat number, 

chemical analysis and destructive testing required for the acceptance of 

this material may be performed on one piece of the material. The same 

rationale may be used for the acceptance of containers of nonmetallic 

materials such as lubricants providing that traceability has been 

established and each container is identified with a unique mix or batch 

number. 

2. Established Lot/Batch Control (Items) 

When lot/batch (defined as units of product of a single type, grade, 

class, size, and composition, manufactured under essentially the same 

conditions and at essentially the same time) control is established 

through a commercial grade survey, the party performing dedication of 

such items can use sampling prescribed by standard statistical methods 

38703, Appendix A A- 2 

162 

Issue Date: 11/08/93 
that are based on homogeneous product lots. Such sample plans should be 

identified and should provide for the verification of the critical 

characteristics with confidence level consistent with the item’s 

importance to safety. Other means of demonstrating adequate lot/batch 

control may include satisfactory performance history and the results of 

receipt inspection/testing. When such methods are used as a basis for 

developing product sampling strategy, they should be supported by 

documented objective evidence. 

3. Material and Items With No Lot/Batch Control 

When lot/batch control cannot be established, sampling plans need to be 

considered on individual, item-specific basis and ensure that they are 

capable of providing a high level of assurance of the item’s suitability 

for service. There may be situations where each item needs to be tested. 

COMMERCIAL GRADE SURVEYS 

1. Verification of Vendor’s Control of Swcific Characteristics 

A comnercial grade survey should be specific to the scope of the CGI(s) 

being purchased. The vendor’s controls of specific critical 

characteristics to be verified during the survey should be identified in 

the survey plan. The verification should be accomplished by reviewing 

the vendor’s program/procedures control 1 ing these characteristics and 

observing the actual implementation of these controls in the manufacture 

of items identical or similar to the items being purchased. 

2. 

3. 

Jdentification of Applicable Proaram/Procedures 

The vendor must have a documented program and/or procedures to control 

the critical characteristics of the item or items being procured that are 

to be verified during the survey. When many items are being purchased, 

a survey of a representative group of similar items may be sufficient to 

demonstrate that adequate controls exist. If the vendor’s controls are 

determined to be satisfactory, purchase orders for these items should 

invoke these controls as contract requirements by referencinq the 

applicable program/procedure(s) and revision. If multiple working level 

procedures are applicable to the vendor’s activities, which affect the 

item’s critical characteristics and these procedures, in turn, are 

controlled by a higher level document, it may be appropriate to reference 

that document in the purchase order. It is important to ensure that the 

specific controls reviewed and accepted during the survey be applied 

during the manufacturing process. Upon completion of the work, the vendor 

should certify compliance with the purchase order requirements. 

Documentation of Survev Results 

Comnercial grade survey documentation should include the identification 

of the item or items for which the vendor is being surveyed, 

identification of the critical characteristics of these items that the 

vendor is expected to control, identification of the controls to be 

applied (program/procedure and revision), and a description of the 

verification activities performed and results obtained. Critical 

characteristics that are not adequately controlled should be addressed by 

contractually requiring the vendor tc institute additional controls or by 

utilizing other verification and acceptance methods. 

Issue Date: 11/08/93 A-3 38703, Appendix A 

163 
4. Survey Freauency 

Comnercial grade surveys should be conducted at sufficient frequency to 

ensure that the process controls applicable to the critical 

characteristics of the item procured continue to be effectively 

implemented. Factors to be considered in determining the frequency of 

commercial grade surveys include the complexity of the item, frequency of 

procurement, receipt inspection, item performance history, and knowledge 

of changes in the vendor's controls. The survey frequency should not 

exceed the audit frequency established for 10 CFR Part 50, Appendix B, 

suppl iers. 

ACCEPTANCE OF CERTIFIED MATERIAL TEST REPORTS (CMTRs) AND CERTIFICATES OF 

COMPLIANCE (CoCs) 

Validity Verified Throuqh Vendor/SuDD lier Audit or Testi nq 

When the verification of critical characteristics is based on vendor CMTRs or 

CoCs, the validity of these documents should be ensured. This can be 

accomplished through a commercial grade survey or, for simple items, periodic 

testing of the product on receipt. Such verifications should be conducted at 

intervals commensurate with the vendor's past performance. If the item's supply 

chain includes a distributor, a survey of the distributor's acti\:,ies may be 

necessary (see "Traceability"). 

USE OF INDUSTRY GUIDANCE 

The Electric Power Research Institute (EPRI) NP-5652, "Guide1 ine for the 

Uti1 ization of Commercial Grade Items in Nuclear Safety Related Applications 

(NCIG-07) ," defines critical characteristics as "identifiable and measurable 

attributes/variables of a CGI, which once selected to be verified, provide 

reasonable assurance that the item received is the item specified." NRC's 

conditional endorsement of EPRI NP-5652 by Generic Letter 89-02 was based on 

interpreting that in the EPRI definition of critical characteristics the "item 

specified" encompassed those attributes that are essential for the performance 

of the item's safety function. This interpretation is consistent with the 

definition of "critical characteristics for acceptance" found in EPRI NP-6406, 

"Guidelines for the Technical €valuation of Replacement Items in Nuclear Power 

Plants," which notes that critical characteristics for acceptance are a subset 

of "critical characteristics for design." The EPRI NP-6406 definition of 

"critical characteristics for design" includes those attributes that ensure the 

performance of the item's design function. 

Published NRC guidance does not differentiate between design and acceptance 

critical characteristics and thp CGI dedication guidance provided in Generic 

Letters 89-02 and 91-05 does not suggest that all design requirements of an item 

need to be werified during the dedication process. Rather, the licensee is 

expected to identify the item's design, material, and performance characteristics 

that have a direct effect on the item's ability to accomplish its intended safety 

function and select from these characteristics a set of critical (or acceptance) 

characteristics that, once verified, will provide reasonable assurance that the 

item will perform that function. Consistency in the definition of critical 

characteristics can be improved by equating the NRC's definition of critical 

characteristics to the EPRI definition of "critical Characteristics for 

acceptance. ' 

END 

38703, Appendix A A-4 

164 

Issue Date: 11/08/93 
APPENDIX B 

DEFINITIONS 

The following terms are listed to provide the inspectors with working definitions 

of important terms used during the procurement and dedication of comnercial grade 

items (CGIs). These terms are defined only in the context of the CGI dedication 

process and are solely to aid the inspector in the inspection process. 

Basic Comment - A plant structure, system, component, or part thereof necessary 

to ensure one of the following: 

the integrity of the reactor coolant pressure boundary 

capability to shut down the reactor and maintain it in a safe shutdown 

condition 

the capability to prevent or mitigate the consequences of accidents that could 

result in offsite radiation exposures comparable to those referred to in 

10 CFR Part 100.11 

Certificate of ComDliance - A written statement attesting that the materials are 

in accordance with specified requirements. 

Certified Material Test ReDort - A document attesting that the material is in 

accordance with specified requirements, including the actual results of all 

required chemical analyses, tests, and examinations. 

Comercial Grade Item - An item satisfying the following criteria: 

not subject to design or specification requirements that are unique to nuclear 

facilities 

used in applications other than nuclear facilities 

ordered from the msnufacturerlsuppl ier on the basis of specifications set 

forth in the manufacturer’s pub1 ished product description (e.g., catalog) 

Comercial Grade Survey - Activities conducted by the purchaser or its agent to 

verify that a supplier of CGIs controls, through quality activities, the critical 

characteristics of specifically designated GIs, as a method to accept those 

i terns for safety-re1 ated use. 

Critical Characteristics - Those important design, material, and performance 

characteristics that, once verified, will provide reasonable assurance that the 

item will perform its intended safety function. 

Dedication - The process by which a CGI is designated for use as a basic 

component. This process includes the identification and verification of critical 

characteristics. (Also refer to definition in 10 CFR Part 21.3(4)(c-l)) 

Enaineerins Judament - A process of logical reasoning that leads from stated 

premises to a conclusion. This process should be supported by sufficient 

documentation to permit verification by a qualified individual. 

Issue Date: 11/08/93 B- 1 

165 

38703, Appendix B 
Source Verification - Activities witnessed at the suppliers’ facilities by the 

purchaser or its agent for specific items to verify that a supplier of CGIs 

controls the critical characteristics of that item as a method to accept the item 

for safety-re1 ated use. 

Traceability - Is the ability to verify the history, location, or application of 

an item by means of recorded identification. 

END 

38703, Appendix B B- 2 

166 

Issue Date: 11/08/93 
APPENDIX C 

CONTENTS OF DEDICATION PACKAGES 

The dedication packages compiled by the licensee may contain the following items, 

as applicable, depending on the item chosen and the dedication methods used. 

Purchase requi si ti ons and purchase orders. 

Other pertinent vendor/l i censee correspondence. 

Design specifications - original and updated to verify certain important 

parameters, such as original design pressure of a system or degraded pickup 

voltage of a solenoid or relay. 

Catalog specifications. 

Procurement basis evaluation - 1 ike-for-1 i ke, equivalency, plant design change 

packages, drawing and specification updates. 

10 CFR Part 50.59 safety evaluation, if required. 

Material receiving reports, packing lists/invoices, and other shipping 

documents. 

Receipt inspection reports and any related test reports. 

Other documents to trace the item from the time it was dedicated to the time 

Certificates of conformance/compl iance/qual i ty. 

it was installed, tested, and accepted. 

Vendor test and inspection reports. 

Third-party or subvendor test and inspection reports. 

Shelf life information. 

Vendor dedication/partial dedication information. 

Design/material/process change history information. 

Completed comnercial grade dedication document including: 

- safety classification 

- 

- identification of critical characteristics 

- identification of verification methods and acceptance criteria for the 

critical characteristics 

identification of safety functions/application requirements 

- evaluation of credible failure modes (if applicable) 

Any deviation from design, material, and performance characteristics relevant 

Issue Date: 11/68/93 c- 1 

167 

38703, Appendix C 
to the safety function (nonconformance dispositions). 

Documents showing objective evidence: 

- special test and inspection procedures and results 

- comnercial grade survey reports - item, design, material, and specific 

performance characteristic (relevant to safety function) 

- source inspection reports 

Completed post-installation test procedure and results. 

Completed stock or material issue forms and installation work orders or 

reports. 

Historical performance information. 

END 

38703, Appendix C c- 2 

168 

Issue Date: 11/08/93 
***** 

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-0001 

June 29, 1995 

Mr. Scott Akers, Jr., Chief Executive Officer 

Cardinal Industrial Products, 

Limited Partnership 

3873 West Oquendo 

Las Vegas, NV 89118 

SUBJECT: CARDINAL’S MAY 30, 1995 REPLY TO NOTICE OF NONCONFORMANCE 

99901076/94-01-03 (NRC INSPECTION REPORT NO. 999001076/94-01) 

Dear Mr. Akers: 

Thank you for Mr. Hathcock’s letter of May 30, 1995, in response to our letter 

of April 24, 1995, which documented our evaluation and conclusions regarding 

your January 30, 1995 response to Notice of Nonconformance 99901076/94-01. 

Accordingly, as agreed in a June 26 telephone conversation between Messrs. 

Potapovs, of this office and Hathcock, of your staff, we would like to have a 

telephone conference with you on July 6, 1995, to further discuss the concerns 

identified in this letter. 

We agree that NRC Generic Letters 89-02 and 91-05 do not apply to 

manufacturers and suppliers. 

letters does not constitute regulatory requirements for NRC 1 icensees. 

of these documents were provided to you for information only. 

Additionally, the guidance contained in these 

Copies 

Our April 24, 1995, letter to you outlined the relationship between NRC 

requirements and the activities of manufacturers and suppliers that provide safety related items for use in licensed facilities. In summary, when an NRC 

licensee invokes the quality assurance program set out in 10 CFR 50, 

Appendix By in its puxhase order to a supplier, the supplier is expected to 

certify conformance with this program. 

program is inadequate to demonstrate conformance, a notice of nonconformance 

may be warranted for the supplier and a notice of violation may be warranted 

for the licensee. The responsibility of the supplier would include assuring 

that the applicable provisions of 10 CFR 50, Appendix By are passed on to 

lower tier vendors whose activities affect the product’s conformance with the 

appl i cab1 e speci f i cat i on. 

To the extent that the supplier’s 

A material supplier does not need to use any sampling plans in order to 

provide material in conformance with the provisions of 10 CFR 50, Appendix B. 

Methods of supplying material in this manner include purchasing the source 

materi a1 from qual if i ed manufacturers or supplying such materi a1 in accordance 

with the qual ity program of the ASME Code which has been endorsed by the NRC 

as complying with the NRC quality assurance provisions. 

As stated in our April 24 letter, for certain applications, the NRC staff has 

accepted commercial grade item (CGI) dedication as an alternative method to 

the programmatic quality assurance measures prescribed in Criteria I1 and VI1 

169 
S. Akers Jr. -2- 

of 10 CFR Part 50, Appendix B, for assuring that the items purchased from 

commercial sources will meet the applicable purchase specification. This 

method of achieving conformance with NRC quality assurance provisions relies 

on the identification and verification of the CGI's critical characteristics 

which are those characteristics that are necessary for the item to perform its 

safety function. 

The use of CGI dedication is most practical when the dedicating entity knows 

the end use of the item, since the end user controls the item's application 

and can, therefore, identify those characteristics of the item which are 

essential for the performance of the item's safety function. The end user of 

the item may also be able to adjust the Acceptable Quality Level (AQL) of a 

sampling plan based on the safety significance of the item's application when 

sampling is used to verify the item's critical characteristics. When a 

material manufacturer or supplier elects to use CGI dedication to provide 

safety-related material, he generally does not know the end application of the 

material and, therefore, is not able to identify a specific set of 

characteristics required for the material to perform its safety function. 

order to certify such material as conforming with 10 CFR 50, Appendix 6, the 

supplier has the option of assuming that all material requirements specified 

in the procurement documents are critical and must be verified. Alternatively, the supplier can request the purchaser of the item to identify 

specific product acceptance requirements in the procurement documents and 

certify conformance only to those requirements. 

In 

The NRC has not published or endorsed any material sampling plans for generic 

use. The references to MIL-STD-105D in our April 24, 1995, letter were for 

the purpose of recognizing a sampling standard that has been used by NRC 

licensees and accepted by the NRC for specific applications. 

generic letters and NRC Inspection Manual Procedure 38703 which were 

enclosures to our April 24, 1995, letter, can be used by vendors for guidance 

in developing CGI dedication and sampling strategies if they elect to supply 

safety related items using the dedication process and to certify such items as 

conforming with 10 CFR 50, Appendix B. This guidance provides a rationale for 

developing sampl ing strategies based on material traceability considerations 

and 1 ot/batch control. 

However, the 

In our April 24, 1995, letter, we also detailed our concerns regarding 

Cardinal's destructive testing sampling plan. Your letter of May 30, 1995, 

stated that the use of certain factors, such as "trending process and 

additional testing (i.e., chemical, mechanical, dimensional, etc.)" in 

selecting a sampling plan, were discussed with the inspectors but not 

considered in the NRC inspection report. 

99901076/94-01, and reiterated in our April 24, 1995, letter, the NRC 

inspectors were not provided any documented bases for the sampling plan during 

that inspection. 

As stated in Inspection Report 

We have considered the suggestion in your May 30, 1995, letter, that the NRC 

should provide 1 i censees and suppl i ers with a cl ear def i nit i on of "reasonabl e 

assurance" and "acceptable sampl ing pl ans. I' Based on the results of recent 

170 
S. Akers Jr. -3- 

NRC vendor inspections, including the inspection of Cardinal, the NRC staff 

has initiated a study of industry practices and existing guidance in this 

area. 

sampl i ng approaches. 

This study, when completed, may result .In NRC’s endorsement of specific 

In accordance with Section 2.790 of 10 CFR, a copy of this letter will be 

placed in the NRC’s Public Document Room. 

No response to this letter is required. 

Sincerely, 

J 

R. Lee Spessard, Director 

Division of Technical Support 

Office of Nucl ear Reactor Regul at i on 

Docket No.: 99901076 

171 
4RC FORM 335 

2-891 

IRCM 1102, 

201,3202 

U.S. NUCLEAR REGULATORY COMMISSION 

BIBLIOGRAPHIC DATA SHEET 

(See instructions on the reverse) 

I. TITLE AND SUBTITLE 

Licensee Contractor and Vendor Inspection 

Status Report 

Quarterly Report 

April - June 1995 

i. AUTHOR(S) 

1. REPORT NUMBER 

(Aulgned by NRC. Add Vol., SUPP., Rw., and Addendum Numben, If any.) 

NUREG-0040 

Vol. 19, No. 2 

3. DATE REPORT PUBLISHED 

MONTH YEAR 

August ' 1995 

4. FIN OR GRANT NUMBER 

6. TYPE OF REPORT 

Quarterly 

April 1995 - June 1995 

7. PERIOD COVERED l/nc/usrveDares) 

I. PERFORMING ORGANIZATION - NAME AND ADDRESS 111 NRC, provide Division. Office or Region. U.S. Nuclear Rrgufarory Commission, andmailing address: if conrracror, provide 

name and mailing addmd 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

I. SPONSORING ORGANIZATION - NAME AND ADDRESS 111 NRC. rype "Same as above': if conrracror,provide NRC Division, Office or Region, U.S Nuclear Regularory Commission, 

and malling address.1 

Same as above. 

0. SUPPLEMENTARY NOTES 

. . . ,__ .. . . ._ . . ___ . . -. __ . .. ... 

This periodical covers the results of inspections performed by the NRC's Special 

Inspection Branch, Vendor Inspection Section, that have been distributed to the 

inspection organizations during the period from April through June 1995. 

12. KEY WORDSIDESCR!PTORS (Lis1 words orphrases :ha: v~illassistresearchers in locating rhemporr.I 

Vendor Inspection 

3. AVAILABILITY STATEMENT 

Unlimited 4. SECURITY CLASSIFICATION 

[This Page) 

Uncl ass i f i ed 

(This Report) 

Unclassified 

5. NUMBER OF PAGES 

6. PRICE 

JRCFORM 335 (2691 
Federal Recycling Program