NUREG0040 

Vol. 21, No. 1 

Licensee Contractor 

and Vendor Inspection 

Status Report 

Quarterly Report 

January-March 1997 

Manuscript Completed: July 1997 

Date Published: July 1997 

Division of Inspection and Support Programs 

Ofice of Nuclear Reactor Regulation 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 
NUREG-0040, VoL 21, No. 1 has been 

reproduced from the best available copy. 
DISCLAIMER 

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

States Government. Neither the United States Government nor any agency thereof? nor 

any of their employees, make any warranty, express or implied, or assumes any legal liability or resporsibiiity 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, procss, or service by 

trade namq trademark, manufacturer, or otherwise does not necessarily constitute or 

imply its endorsement, reutmmedation, or favoring by the United States Government or 

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

DISCLAIMER 

ABSTRACT 

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

Spec a1 Inspection Branch, Vendor Inspection Section, that have been 

distributed to the inspected organizations during the period from January 1997 

through March 1997. 

iii 

CONTENTS 

PAGE 

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

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

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

ACCUTECH 

Las Vegas, NV 

(99901307/96-01) ......... 2 

Caldon, Inc. 

Pittsburgh, PA 

(99901311/97-01) ......... 32 

Ellis & Watts 

Batavia, OH 

(99901308/96-01) ......... 40 

Limi torque Corporation 

Lynchburg, VA 

(99900100/97-01) ......... 53 

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

59 

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 

pl ants. 

for the inspection of commercial nuclear facilities to provide for multiple 

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

participates in a quality verification process in compliance with requirements 

prescribed by the NRC’s rules and regulations (Title 10 of the Code of Federa7 

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

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

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

industry within the framework of quality verification programs. 

The Federal government and nuclear industry have establ i shed a system 

The licensee is responsible for developing and maintaining a detailed quality 

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

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

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

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

codes, and standards. 

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

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

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

laboratories performing equipment qualification tests, and holders of NRC 

construction permi ts and operating 1 i censes i n vendor-re1 ated areas. 

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

problems are determined and appropriate corrective actions are developed. The 

inspections also review vendors to verify conformance with applicable NRC and 

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

coordination between licensees and vendors. 

These 

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

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

and review of equipment problems found during operation and their corrective 

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

the inspected organization is required to take appropriate corrective action 

and to institute preventive measures to preclude recurrence. 

imp1 ications are found, NRC ensures that affected 1 icensees are informed 

through vendor reporting or by NRC generic correspondence such as information 

notices and bulletins. 

When generic 

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

during the calendar quarter for which it is published. 

report lists the nuclear facilities inspected. This information will also 

alert affected regional offices to any significant problem areas that may 

require special attention. Appendices 1 ist selected bulletins, generic 

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

correspondence i nvol vi ng vendor i ssues . 

Each vendor inspection 

viii 
INSPECTION REPORTS 

1 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 2CtW-0001 

March 4, 1997 

Mr. Roy P. Reindl, Branch Manager 

ACCUTECH 

A Division of B&G Manufacturing Co. Inc. 

3873 W. Oquendo Road 

Las Vegas, NV 89118 

SUBJECT: NRC INSPECTION REPORT 99901307/96-01 AND NOTICES OF VIOLATION AND 

NONCONFORMANCE. 

Dear Mr. Reindl : 

On January 30, 1997, the U.S. Nuclear Regulatory Commission (NRC) completed an 

inspection at the, ACCUTECH Division of B&G Manufacturing Co. Inc. 

enclosed report presents the results of that inspection. 

During this inspection, the NRC inspectors found that certain of your 

activities appeared to be in violation of NRC requirements. 

contrary to the requirements of 10 CFR 21.21 of the Code of Federal 

Reaulations, you did not perform an adequate evaluation of certain deviations 

identified by your customers in order to determine their relationship to the 

conclusions of your previous evaluations and to determine whether any 

additional notifications were required. You also failed to identify the 

existance of additional potentially defective material that required an 

evaluation. In a separate instance, after your evaluation determined the 

existence of a defect, you failed to notify the NRC within the time limits 

required by the regulation. The failure to perform an adequate evaluation of 

potentially reportable conditions is of special concern since it is an 

essential part of the process that assures that NRC licensees are informed of 

situations where they may have been supplied with basic components that 

contain defects which could create a substantial safety hazard. 

The 

Specifically, 

These violations are cited in the enclosed Notice of Violation (NOV), and the 

circumstances surrounding the violation(s) are described in detail in the 

enclosed report. 

the instructions specified in the enclosed NOV when preparing your response. 

The NRC will use your response, in part, to determine whether further 

enforcement action is necessary to ensure compliance with regulatory 

requirements. 

You are required to respond to this letter and should follow 

In addition, the NRC inspectors determined that the implementation of your 

quality assurance program failed to meet certain NRC requirements imposed on 

you by your customers. Specifically, for material supplied to the 

requirements of Section I11 of the ASME Code, you failed to provide your 

customers with all of the documentation as required by the ASME Code. 

Additionally, in at least one instance, your files did not contain 

documentation to show that all tests required by the material specification 

had been performed on material obtained from a supplier, or that the supplier 

of this material had been qualified to the ASME Code requirements. 

2 
Furthcp for non-Code material supplied tr the requirement3 nf 10 Part CFR 50, 

Appendix 8 (Appendix B), the inspectors identified instances where the 

available documentation did not demonstrate that the material manufacturers 

had been surveyed and audited in accordance with the requirements of Appendix 

I?. Although ACCUTECH performed limited physical testing of this material, the 

sampling plan used for this testing did not have a documented basis to 

demonstrate that this material was equivalent tg material purchased from a 

supplier q'ralified to Appendix 6 requirzments. The inspectors also determined 

that, contrary to your QA manual and implementing procedure requirements, 

there was no evidence to indicate that material and documents acquired from 

Cardinal Industrial Products as warehouse stock had been reviewed and approved 

before its sale by ACCUTECH as safety related material. 

Additionally, the inspectors identified that you did not initiate a corrective 

action report for n-?conformances described in NRC Report 99901076/94 -01. 

Although the nonconformances were issued to Cardinal Industrial Products, 

corrective actions, including corresponaence with the NRC continued after 

B&G's acquisition of the company and ACCUTECH continued to supply material 

using processes (sampling) related to a previously identified 

nonconformance. 

needs additional response from ACCUTECH. 

This nonconformance (9990; 376/94-01-03) is considered open and 

These nonconformances are cited in the enclosed Notice of Nonconformance 

(NON), and the circumstances surrounding them are described in detail in the 

enclosed report. 

should follow the instructions specified in the enclosed NON when preparing 

your response. 

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

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

Room (PDR). 

You are requested to respond to the nonconformances and 

Sincerely, 

0RIGI.W SIGm BY 

Robert M. Gallo, Chief 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Docket No. 99901307 

Encl osures : 1. Notice of Vi ol at i on 

2. Notice of Nonconformance 

3. Inspection Report 99901307/96-01 

3 
AC C UT EC H 

Las Vegas, Nevada 

NOTICE OF VIOLATION 

Docket No.: 99901307 

During an NRC inspection conducted at ACCUTECH's Las Vegas facility on 

November 12 through 14, 1996, and January 28 through 30, 1997, violations of 

NRC requirements were identified. In accordance with the "General Statement 

of Pol icy and Procedure for NRC Enforcement Actions," NUREG-1600, the 

violations are 1 isted below: 

A. 10 CFR 21.21, "Notification of fai!ure to comply or existence of a 

defect and its evaluation," requires, in part, that a director or 

responsible officer must notify the Commission when he or she obtains 

information reasonably indicating a defect in a basic component. 

Notification to the NRC must be made be facsimile or by telephone within 

two days following the receipt of the informdtion and written 

notification to the NRC within 30 days foliowing receipt of the 

information. 

Contrary to the above, ACCUTECH, after determining that heat lot TS7 

contained defective fasteners, did not notify the NRC within two days of 

obtaining information of a defect affecting a basic component and did 

not provide the written notification within the required 30 days of 

obtaining information of a defect affecting a basic component. 

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

(99901307/96-01-01) 

8. 10 CFR 21.21, "Notification of failure to comply or existence of a 

defect apd its evaluation," requires, ir, part, that deviations and 

failures to comply be evaluated in order to identify a reportable defect 

or failure to comply that could create a substantial safety hazard were 

it to remain uncorrected. 

Contrary to the above, ACCUTECH did not adequately evaluate whtther the 

identified defects in heat lots TS7 and K7 affected the validity of the 

conclusions documented in the November 1995, Final Report to determine 

if additional notifications were required. 

eval uate avai 1 ab1 e information re1 ated to the heat treatment of heat 1 ot 

K7 to identify the existance of an additional suspect heat lot (M2). 

ACCUTECH also did not 

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

(99901307/96-01-02) 

Pursuant to the provisions of 10 CFR 2.201, ACCUTECH, is hereby required to 

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

Comnission, ATTN: Document Control Desk, Washington D.C. 20555, with a copy 

to the Chief, Special Inspection Branch, Division of Inspection and Support 

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

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

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

4 
for disputing the violation, (2) the corrective steps that have been taken and 

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

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

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. 

Because your respon- will be placed in the NRC Public Document Room (PDR), to 

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

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

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

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

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

withholding the information from tne public. 

Dated at Rockville, Maryland 

this 4th day of March 1997 Enclosure 1 

5 
NOTICE OF NONCONFORMANCE 

ACCUTECH Division of B&G Mfg. Co.,Inc. 

Las Vegas, Nevada 

Docket No.: 99901307 

Based on the results of an inspection conducted on November 12 through 14, 

1996, and January 28 through 30, 1997, 1996, it appears that certain of your 

activities were not conducted in accordance with NRC requirements. 

A. Criterion VII, "Control of Purchased Material, Equipment, and 

Services,"of Appendix B to 10 CFR Part 50, requires,in part, that 

measures shall be establ i shed to assure that purchased materi a1 conforms 

to procurement documents. 

Paragraph NCA-3862.l(b) of Section I11 of the ASME Code states that 

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

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

approved supplier's certification for the operations performed shall be 

furnished as an identified attachment to the Certified Material Test 

Report. 'I 

Paragraph NCA-3853.3(a) of Section I11 of the ASME Code states that the 

Materials Organization shall be responsible for establishing and 

verifying that the suppliers controls applicable to the activities 

performed. are adequate by surveying and auditing the suppl ier's 

established quality program which is consistent with the requirements of 

this Subarticle or having the supplier perform the activities with 

controls established by the Material Organization's program. 

The fol1 owing four examples demonsrate ACCTECH's fai 1 ure to comply with 

the requirements and consi tute Nonconformamce 99901307/96-01-03. 

1. Contrary to the above, ACCUTECH's Certified Material Test Report 

(CMTR) for material supplied to PECO Energy Company (PECO) under 

Purchase Order (PO) LS 605919, dated January 18, 1996, certifiJ 

that this material was supplied in accordance with their ASME 

Quality Systems Certificate (QSC), but did not reference or include 

the mill heat analysis as an identified attachment. 

2. Contrary to the above, ACCUTECH's CMTR for material supplied to PECO 

under PO LS 607801, dated March 3, 1996, certified that this 

material was supplied in accordance with their QSC, but did not 

include or reference the mill heat analysis or certification for 

heat treatment performed by an approved supplier as identified 

attachments. 

3. Contrary to the above, ACCUTECH's CMTRs for material supplied to 

Consumers Power Company under PO G 0183280, dated December 7, 1995, 

and to Entergy under PO MP 96L093, dated November 8, 1995, certified 

that the material was supplied under their QSC, but did not include 

or reference reports of mill heat analysis or mechanical testing 

performed by approved suppliers. Additionally, documentation for 

6 
this material d 

(Macroetch) had 

d not verify that all tests required to be performed 

been completed . 

4. Contrary to the above, for material described in 3, above, ACCUTECH 

could not provide documentation that organizations supplying thi s 

material had quality assurarce programs that complied with the 

requirements of NCA 3800 or with contrcl s establ i shed under 

ACCUTECH's program. 

B. Criterion VII, "Control of Purchased Material, Equipmeiit, and Services," 

of Appendix B to 10 CFR Part 50, requires,in part, that measures shall 

be establ i shed to assure that purchased material conforms to procurement 

documents. 

Paragraph 6.1 of ACCUTECH's Quality Systems Manual assigns to QA 

Department the responsibility for work order review prior to 

certification and states that, as a minimum, this review shall include 

verification that customer PO requirements are met and the material 

conforms to the specification requirements. 

Contrary to the above, ACCUTECH's QA Department did not provide adequate 

verification that material supplied to PECO under their PO LS 606821, 

dated February 15,1996, and to Wisconsin Electric Power Co. under their 

PO 4500021861, dated November 1, 1996, complied with the PO 

requirements. 

suppliers of this material had been surveyed and audited to verify their 

conformance with the applicable provisions of 10 CFR, Appendix B. 

Although ACCUTECH performed 1 imited physical testing of this material, 

the sampling plan used for this testing did not have a documented basis 

to demonstrate that the material supplied under these POs was equivalent 

to material purchased from suppliers qualified under the requirements of 

Appendi) B. (Nonconformance 99901307 /96-01- 04) . 

Specifically, ACCUTECH could not demonstrate that the 

C. Criterion XVI, "Corrective Action" of Appendix B to Part 50 states, in 

part, "Measures shall be established to assure that conditions adverse 

to quality, cuch as failures, malfunctions, deficiencies, ,deviations, 

defective material, and equipment, and nonconformances are promptly 

identified and corrected. 

condition adverse to quality, the cause of the condition, and the 

corrective action taken shall be documented and reported to appropriate 

1 eve1 s of management. I' 

The identification of the significant 

ACCUTECH Standard Operating Procedure (SOP) 17.001, "Corrective Action, I' 

Revision 5, dated November 15, 1995, states, in part, in Section 3.1 

that a failure of the QA program to comply with an approved regulation 

or commitment (ASME, ASTM, ANSI, MIL-STD etc.), is one of the conditions 

considered as adverse to quality. Section 3.2 states that conditions 

adverse to quality may be identified during an Internal Audit, 

Management Audit, Supplier Audit or by other means. Section 3.3 states 

that the inaividual (usually a Lead Auditor or Auditor) identifying what 

appears to be a condition adverse to qual'ity shall initiate a Corrective 

Action Report (CAR) . 

7 
Contrary to the above, ACCUTECH failed to identify two nonconformances 

described in NRC Inspection Report No. 99901076/94-01, i .e. failures to 

comply with a regulation (10 CFR 50, Appendix B), as conditions adverse 

to quality and did not enter them in the corrective action process by 

initiating a CAR. Although these nonconformances were issued to 

Cardinal Industrial Products, corrective actions, includ'ng 

correspondence between ACCUTECH and NRC, continued after ACCUTECH's 

acquisition of Cardinal Industrial Products. (Nonconformance 

99901307/96-01-05) 

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

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

prescri bed by documented instructions , procedures , or drawings , of a 

type appropriate to the circumstances and that such activities be 

accomplished in accordance with these instructions, procedures, or 

drawings . 

ACCUTECH Quality System Manual Section 22.0, "Interface Activities for 

Material and Associated Documentation Which Has Been Legally Transferred 

From Cardinal Industrial Products Corporation (CIPC) to ACCUTECH. 

(Note: This material has not been removed from the warehouse facilities 

or control of ACCUTECH's Qual i ty Assurance Department) , It Revision 0, 

dated November 29, 1995, states, in Section 22.6, that ACCUTECH record 

packages shall contain CIPC documentation used to justify acceptance of 

the material. 

Standard Operating Procedure (SOP) 22.001 , "Transference of Material and 

Associated Documentation Between CIPC and ACCUTECH," Revision 2, dated 

November 15, 1995, states, in part, in Section 1.1 that the purpose of 

the procedure is to "ensure that ACCUTECH materials and documents are 

properly reviewed and approved prior to acceptance and use by ACCUTECH 

as Code and safety-re1 ated materi a1 s. I' Sect ion 2.2 requires evidence of 

document review to be recorded on a "Document Review Checklist" form for 

each ACCUTECH P.O. item reviewed. 

Contrary to the above, no documentation existed to verify that ACCUTECH 

had implemented the requirements of SOP 22.001 since July 1995 and had 

ensured that material and documents were properly reviewed and approved 

before materi a1 acquired from Cardinal Industrial Products warehouse 

stock was sold to utilities as ASME Code or safety-related 10 CFR 50, 

Appendix B. (Nonconformance 99901307/96-01-06) 

E. Criterion V, "Procedures," of 10 CFR Part 50, Appendix By requires in 

part that activities affecting quality shall be prescribed by documented 

instructions, procedures, or drawings, of a type appropriate to the 

circumstances and shall be accomplished in accordance with these 

instructions procedures, or drawings 

ACCUTECH Standard Operating Procedure (SOP) 17.002 "Reporting of 

Defects and Noncompliance," Revision 6, dated November 15, 1995, 

requires the documentation of 10 CFR Part 21 evaluations on SOP Form 

17.2. 

a 
Report No: 

Organization: 

Contact: 

Nuclear Industry 

Activity: 

Dates: 

Inspectors: 

Approved by: 

U. S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

99901 307/96-01 

ACCUTECH Division of B&G Mfg. Co., Inc. 

3873 W. Oquendo Road 

Las Vegas, Nevada 89118 

Roy P. Reindl, Branch Manager 

(702) 739 1966 

Manufacturer and supplier of threaded fasteners, 

fittings, flanges, and other items used primarily in 

nuclear applications. 

November 12-14, 1996 

January 28-30, 1997 

Uldis Potapovs, Senior Reactor Engineer 

Richard P. McIntyre, Senior Reactor Engineer 

Billy H. Rogers, Reactor Engineer 

Gregory C Cwalina, Chief 

Vendor Inspection Section 

Special Inspection Branch 

Division of Inspection and Support Programs 

9 
1 INSPECT ION SUMMARY 

During this inspection, the NRC inspectors reviewed the implementation of 

selected portions of B&G Division of ACCUTECH (ACCUTECH) qual i ty assurance 

(QA) program with emphasis on commercial grade material dedication practices 

and upgrading of unqualified source material for applications requiring 

certification to American Society of Mechanical Engineers (ASME) Boiler and 

Pressure Vessel Code (Code) requirements. The inspectors a1 so reviewed the 

implementation of ACCUTECH's program for reporting of defects under Part 21 of 

Title 10 of the Code of Federal Requlations (10 CFR Part 21). This inspection 

was a part of an NRC pilot program in the development of inspection procedures 

for assessing licensee oversight of supplier quality. 

inspection can be considered as vendor shop follow up of an NRC inspection of 

PECO Energy Company (PFCO). 

In that regard, this 

The inspection bases were: 

Appendix 6, "Qual i ty Assurance Criteria for Nuclear Power P1 ants and 

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

Requlations (10 CFR Part 50) 

10 CFR Part 21 , "Reporting of Defects and Noncompl i ance" 

ASME Code, Section 111, Subarticle NCA 3800. 

During this inspection, two violations of NRC requirements were identified and 

are discussed in Section 3.2.2.c of this report. 

During this inspection, five instances where ACCUTECH failed to conform to NRC 

requirements imposed upon them by NRC licensees were identified. 

nonconformances are discussed in Sections 3.2.2.~~ 3.3.2, 3.4.3, 3.6, and 3.8 

of this report. Also, Nonconformance 99901076/94-01-03 remains open. 

These 

2. STATUS OF PREVIOUS INSPECTION FINDINGS 

Although this was the first inspection under BAG'S ownership of this company 

(previously Cardinal Im*-trial Products, Limited Partnership [Cardinal J) , inspection findings from the NRC, December 1994 inspection of Cardinal that 

relate to material currently being supplied by ACCUTECH were reviewed. 

2.1 Nonconformance 99901076/94-01-03 (OPEN) 

Contrary to Criterion VI1 of 10 CFR Part 50, Appendix 6, Cardinal had 

not established a documented basis to substantiate that its destructive 

testing sampling plan for verifying critical characteristics provided 

reasonable assurance that dedicated commercial grade items (CGIs) met 

the applicable procurement document requirements. 

2 

10 
a. Bac kclround 

Several exchanges of correspondence took place between the NRC staff , Cardinal, and B&G in 1995. The last letter, dated August 30, 1995, was 

written after B&G’s acquisition of the company and included a rationale 

for selecting sampling plans for commercial grade item dedication. 

NRC’s acknowledgement of this letter, dated December 5, 1995, stated 

that, while the rationale presented in the August 30, 1995, letter could 

be expected to improve the overall assurance of product integrity, it 

placed heavy reliance on visual and dimensional inspection to assure lot 

homogeneity. 

destructive sampling plan, the NRC December 5, 1995, letter stated that, 

for sampling nonhomogeneous product lots, the NRC staff has generally 

accepted a confidence level of 90-35 % that no more than 5-10 % of the 

sampled items are ncnconforming. It stated further that the destructivc 

sample size and rationale described in the August 30, 1995, letter does 

not appear to provide this level of confidence for random lots 

(unverified traceability) of material, especially for the verification 

of critical characteristics related to the p,iysical properties of 

materi a1 s. 

With respect to the acceptability of the proposed 

The NRC December 5, 1995, letter also stated that a supplier who 

certifies his product as complying with 10 CFR Part 50, Appendix B can 

either supply a product that has been designed and manufactured under 

the applicable provisions of Appendix By or supply a dedicated 

commercial grade item after verifying all of the item’s critical 

characteristics. 

confidence level of the sampling plan should meet the criteria discussed 

above. The letter further stated that suppliers with dedication 

programs that do not demonstrate this confidence level for verifying 

critical characteristics have the option of supplying such material 

under their licensee approved quality programs, but without certifying 

compliance with Appendix B. 

If verification is based on product sampling, 

b. Observations and Findinqs 

The inspectors att,-pted to verify the sampling plan rationale as 

described in Cardinal’s August 30, 1995, letter. This was done by 

reviewing the documentation that would support Cardinal’s statements 

concerning visual and dimensional inspection and trending data and the 

use of this information as a basis for sampling decisions. Cardinal 

contended that for finished fasteners procured from non-approved/nonaudited suppliers, Cardinal performs 100% visual inspection to identify 

visual discontinuities per the applicable material specification. It 

further stated that visual inspection can identify mixed fastener lots 

since different types or size of product markings; product color 

difference; and different forging indications all could be an indication 

of a mixed fastener lot. It went on to say that after a product has 

received a dimensional and visual inspection, the probability of the 

product complying with the chemical and mechanical requirements are 

extremely high, based on nonconformance report (NCR) trending data and 

years of manufacturing and testing of fasteners. 

3 

11 
The inspectors reviewed ACCUTECH Standard Operat i ng Procedures (SOP) 

4.001, Purchase Order Processing, Revision 8, dated November 15, 1995, 

and SOP 9,001, Material Receiving Inspection, Revision 12, dated 

November 15, 1995, to verify that it included appropriate receipt 

inspection guidance to effectively identify mixed fastener lots that 

could provide an indication of lot homogeneity. 

determined that SOPS 4.001 and 9.001 did not include any procedural 

guidance for &-'--mining lot homogeneity beyond normal visual inspection 

and review for manufacturing process defects and shipping damage. Also, 

several fastener receipt inspection records reviewed did not include any 

documentation to indicate that visual inspection addressed lot 

homogeneity and mixed fastener lots. The inspectors also reviewed 

several Qual i ty Trending Reports for internal and external (vendor) 

NCRs, but these roports also did not include any specific trending data 

on fastener testing f:ilures related to sample plans. 

The inspectors 

In response to inspector questions concerning documentation and control 

of indicators cf lot homogeneity, ACCllTECH Ft.ated that they did not have 

a documented basis to support the information that was previously 

described to the NRC in Cardinal's August 30, 1995, letter. 

c. Conclusions 

In conclusion, the inspectors determined that ACCUTECH continues to 

place heavy reliance on visual and dimensional inspection to support the 

verification of lot homogeneity. 

verification of lot homogeneity, ACCUTECH then utilizes the ASTM A-325 

shipping lot sampling plan for destructive testing (material chemistry 

and mechanical properties) and the EPRI guidelines for nondestructive 

testing (dimensional). 

inspection for shipping damage and manufacturing defects can not assure 

that all items in the same product lot were manufactured from the same 

heat of material or were heat treated under the same conditions. 

Additionally, the inspectors noted that, as discussed in Inspection Report 99901076/94-01, the use of ASTM A-325 shipping lot sampling plan 

is inappropriate for this application. 

Based upon this method for lot 

The NRC inspectors determined that visual 

Based on the above information, Nonconformance 99901076/94-01-03 i s 

considered Open, and requires additional response from ACCUTECH to 

address the original concerns documented in Nonconformance 

99901076/94-01-03 and discussed further with the NRC inspectors during 

the November 1996 inspection. 

3. INSPECTION FINDINGS AND OTHER COMMENTS 

3.1 Descriotion of Facilities and Activities 

B&G Manufacturing Co., Inc. (B&G) purchased the name and certain assets 

of Cardinal on July 10, 1995, and operated the acquired company as 

Cardinal Industrial Products Division of B&G (B&G-Cardinal) until 

4 

12 
December 1995 , when they renamed it ACCUTECH. The transition i ncl uded 

acquisition of some new production equipment, upgrading of in-house heat 

treatment capabi 1 i ty, and changes in company management. 

According to ACCUTECH management, standard fasteners (ASTM/ASME) 

constitute about 80% of the cornpanj's product. 1 ine. Spccialty fasteners 

account for 10% of the product volume and standard material forms 

(plate, flanges) make up the remaining 10%. ACCUTECH has been 

accredited by the ASME as a Materials Organization (MO). Their ASME 

Quality Systems Certificate (QSC) expires on January 9, 1999. 

3.2 10 CFR Part 21 Prowam 

3.2.1 ImDl ement i na Procedure 

The inspectors reviewed ACCUTECH SOP 17.002 , "Reporting of Defects and 

Noncompliance," Revision 6, dated November 15, 1995, and representative 

documentation to verify implementation. Procedure 17.002 established 

the responsibilities and actions for the reporting of defects and 

informing of deviations in accordance with the requirements of 10 CFR 

Part 21. The inspectors reviewed procedure 17.002 to determine whether 

the weaknesses identified during the previous inspection and documented 

in Inspection Report 99903076/94-01, had been adequately addressed. 

Inspection Report 99901076/94-01 identified that procedure 17.002 did 

not contain provisions to inform customers of deviations that Cardinal 

could not evaluate. This weakness was addressed in Revision 6 of 

procedure 17.002, paragraph 2.4.3, which stated that, in the cases when 

ACCUTECH could not determine if a significant safety hazard exists, the 

Quality Assurance Manager should notify all affected customers within 

five working days. 

Inspection Report 99901076/94-01 identified that procedure 17.002 did 

not contain provisions for documentation of evaluations to determine 

whether a deviation was a defect. This weakness was addressed in 

Revision 6 of procedure 17.002, paragraphs 2.2, 2.3, 2.4.1, 2.4.2, and 

2.7, which required that deviations be documented on foltli CF 17.2 and 

reviewed by the Quality Assurance Manager to determine whether a 

substantial safety hazard existed, or whether the deviation information 

should be provided to the customer for evaluation. 

Inspection Report 99901076/94-01 identified that procedure 17.002 did 

not contain provisions to notify the NRC when a defect was determined to 

exist. 

paragraph 2.4.2, which required that the,NRC be notified within two days 

by telefax and thirty days in writing when ACCUTECh had evaluated a 

deviation and determined that a substantial safety hazard existed. 

This weakness was addressed in Revision 6 of procedure 17.002, 

The inspectors concluded, for the weaknesses discussed above, that the 

applicable revisions to procedure 17.002 had addressed the weaknesses 

and had adequately proceduralized the requirements of 10 CFR Part 21. 

5 

13 
In addition, Inspection Report 99901076/94-01 identified that procedure 

17.002 interchanged the terms "defect" and "deviation" in several 

places. 

completely addressed in Revision 6 of procedure 17.002. There were two 

instances were the term defect (or defective) was used in place of the 

appropriate term, deviation. Pa-agrapl, 2.3 stated "when an employee is 

aware of defective material which has been shipped to a customer, he 

shall complete form CF 17.2 and submit one copy to the following: 

Branch Manager and Quality Assurance Manager." As stated in 10 CFR Part 

21, a defect results from an evaluation of a deviation. An employee 

could become aware of material containing a deviation which would 

require evaluation by ACCUTECH management, or the customer, to determine 

if a defect exists. Additionally, paragraph 2.4.1 stated, "In cases 

where the defect does not create a substantial safety hazard ....I' 

the 10 CFR Part 21 definition, a defect can create a substantial safety 

hazard, and the correct term in the context of paragraph 2.4.1 is 

deviation. 

address the weakness identified in Inspection Report 99901076/94-01 and 

a weakness still existed concerning the correct use of the terms 

deviation and defect. 

indicated that this issue would be addressed in a subsequent revision of 

procedure 17.002. 

The inspectors determined that this weakness had not been 

By 

The inspectors concluded that this revision had partially 

This was discussed with ACCUTECH management who 

3.2.2 Part 21 Prosram ImDlementation 

a. Backqround 

In the following discussion Cardinal is used to refer to the pre-July 

10, 1995, business and B&G-Cardinal or ACCUTECH are used to refer to the 

B&G owned business as applicable to the particular activities, 

correspondence, or documentation reviewed. 

ACCUTECH had performed several 10 CFR Part 21 evaluations, all related 

to a single topic, during the period since the previous NRC inspection 

of Cardinal in 1994. The subject of the 10 CFR Part 21 evaluations was 

the failure of certain ASME SA-193, Grade 87 (B7) fasteners, of various 

sizes, to meet the required mechanical specifications. The failures to 

meet specifications were determined to be due to inadequate heat 

treatment, performed during the production of the fasteners, which had 

occurred during the time period prior to B&G's purchase of Cardinal on 

July 10, 1995. 

ACCUTECH used Heat Numbers, Trace Codes, and Heat Codes to provide 

traceability of the fasteners it manufactured, as follows: When 

ACCUTECH purchased raw material from the manufacturing mill, the 

material was received with a mill assigned Heat Number which identifies 

a specific melt of the material with uniform chemical properties. 

ACCUTECH assigned a Trace Code to the first heat lot of a particular 

sized fastener manufactured from this material. ACCUTECH used the same 

Trace Code, combined with a Heat Code, for additional heat lots of the 

fastener manufactured from the same material (each additional heat lot 

was assigned a unique Heat Code). As a result, the fasteners size, 

6 

14 
Trace Code and Heat Code uniquely identified the material that a 

fastener was produced from and during what furnace operation a fastener 

was heat treated. 

b. Identification of Defective and SusDect Heat Lots of Cardinal 

Yanufactured B7 Fasteners 

The inspectors reviewed the ACCUTECH 10 CFR Part 21 evaluations 

performed in accordance with ACCUTECH Standard Operating Procedure (SOP) 

17.002, "Reporting of Defects and Noncompliance," Revision 6, dated 

November 15, 1995, including representative documentation, to verify 

implementation of the requirements of 10 CFR Part 21. 

Shortly after the July 1995, B&G acquisition of Cardinal, the Duquense 

Light Company (Duquense) informed B&G-Cardinal that one of two 87 

capscrews, purchased from Cardinal, had failed during the performance of 

acceptance testing and, when an additional three specimens were tested, 

one failed the tensile test. In July 1995, Duquense witnessed 

additional testing at B&G-Cardinal ot four fasteners during which one 

failed the tensile test. In addition, an August 15, 1995, letter from 

Metallurgical Testing Corporation to B&G-Cardinal, which documented a 

metallurgical examination of the failed fasteners, discussed possible 

causes of the fastener failures and identified inadequate heat treatment 

as a likely cause. 

B&G-Cardinal generated an internal document on August 29, 1995, form 

CF 17.2, "Reporting Defects and Noncompliance 10 CFR Part 21," which 

noted the improper heat treating of heat lot TU1 where the center of the 

heat charge had apparently not reached heat treatment temperature. This 

was evaluated, determined to be a defect, and to be reportable to the 

NRC. Supporting documentation identified the customers, which included 

eight licensees. B&G-Cardinal followed with an August 29, 1995, letter 

to the NRC which identified three heat lats S4, S5, and TU1 as being 

defective, indicated that all customers had been notified, and indicated 

that a complete report would be provided to the NRC within 60 days. 

B&G-Cardinal issued letters on August 30, 1995, to all affected 

customers informing them that they had been supplied fast&:,,a-s from a 

heat lot which contained defective product (heat lots S4, S5, and TU1). 

In addition, on September 5, 1995, B&G issued letters to the affected 

customers, reiterating that defective heat lots existed and that B&G 

intended to investigate and appraise the customers of the results. 

September 11, 1995, B&G-Cardinal issued a letter to the NRC which 

summarized previous events concerning the three defective heat lots 

(heat lots S4, S5, and TUl), stated that all affected customers had been 

notified, and that all B7 inventory was on hold and being tested. 

On 

On September 11, 1995, B&G-Cardinal issued letters to all affected 

customers addressing a fourth heat lot, TS1, that contained defective 

product. On September 12, 1995, B&G issued a letter to all affected 

customers of heat lot TS1, reiterating that the defective heat lot 

existed and ihat B&G intended to investigate and appraise the customers 

of the results. On September 12, 1995, B&G-Cardinal issued a letter to 

7 

15 
the NRC which identified the additional affected heat lot TSl, for a 

total of four affected heat lots and identified the customers which had 

received material from the four affected heat lots. 

On September 26, 1995, B&G-Cardinal issued a letter to the NRC which 

identified five additional suspect heat lots of B7 fasteners (Q5, CXl, 

TU2, L6, and TU1 (this is an additional heat lot, of different sized 

fasteners, thacl Lne TU1 listed in the August 29, 1995, ACCUTECH 

letter)). The letter indicated that all customers had been notified, 

and indicated that a complete report would be provided to the NRC within 

60 days. On September 29, 1995, B&G issued a letter to all affected 

customers notifying that they had been supplied fasteners from a heat 

lot suspected of containing Pnfn-tive product (for the fifth through 

ninth heat lots 45, CXI, TU2, L6, and TUl). 

On November 21, 1995, B&G issued a letter to NRC, containing the 

ACCUTECH Final Report, which discussed the heat treatment concerns, 

immediate action, investigation, firldi?gs, conclusions, and summary. 

The Final Report documented the extensive testing and analysis that 

ACCUTECH had pwformed to determine the cause of the defective B7 

fasteners and to bound the concern. 

of tests in which twenty-two fastener heat lots, representative of 

earlier Cardinal productions heat lots, were heat treated and then 

tested for applicable mechanical properties. Six of the twenty-two test 

heat lots produced fasteners with unacceptable properties. The main 

process variables affecting the heat treatment were determined to be the 

total weight of the furnace charge, packing density of the material 

being heat treated, and the time at the required temperature. The 

larger the weight and density of the fastener heat lot and the shorter 

the length the of heating time, the more likely the fastener heat lot 

would contain fasteners which exhibited unacceptable mechanical 

properties. ACCUTECH then took the weight/heat-time combinations of the 

failed test heat lots, compared the information with Cardinal production 

heat lots, and determined that an additional five Cardinal production 

heat lots of 87 fasteners were potentially defective (suspect). The 

five heat lots identified as suspect (TUl, 45, CX1, TU2, and L6) had 

been addressed earlier in the September 26, 1995, B&G-Cardinal ietter to 

the NRC and the September 29, 1995, letter to the applicable customers. 

In addition, the report indicated that B&G-Cardinal believed that the 

problem of substandard fasteners did not extend beyond the heat lots of 

B7 fasteners identified as defective or suspect, including any other 

materials heat treated by Cardinal. The report also indicated that this 

conclusion could not be guaranteed and that the customer should consider 

replacing fasteners from the identified heat lots and consider testing 

other heat lots as necessary. On February 5, 1996, B&G issued a letter, 

which provided the B&G final report to all affected customers of the 

identified heat lots (through November 1995). 

ACCUTECH had performed a sequence 

On March 22, 1996, ACCUTECH received information from Washington Public 

Power Supply System (WPPSSj which indicated that WPPSS had determined 

that defective material had been provided to them from a heat lot which 

had not been included in the final report as defective or suspect (heat 

8 

16 
lot TS7). ACCUTECH had previously removed heat lot TS7 from nventory, 

but had some of the removed material still on site. In addition, 

material from another heat lot, K7, which had also been removed from 

inventory was still on site. ACCUTECH tested both heat lots, TS7 and 

K7, and determined that some items, in both heat lots, did not meet 

specification. 

and K7 by letter dated May 8, 1996. 

ACCUTECH notified the NRC of the defective heat lots TS7 

In addition, during the January 1997, NRC inspection, ACCUTECH 

identified an additional heat lot, M2, which had been heat treated as a 

part of the same furnace charge as heat lot K7. 

completion of the inspection, ACCUTECH filed a 10 CFR Part 21 report to 

the NRC which provided notification of the suspect heat lot M2. 

Subsequent to the 

During the NRC inspection, ACCUTECH indicated that, as part of the 

corrective action process, all 87 fasteners heat treated by Cardinal, 

had been removed from ACCUTECH inventory shortly after Duquense had 

informed ACCUTECH of the defective B7 fasteners. 

c. Control of the Heat Treatment Process 

The inspectors reviewed the available furnace logs, temperature recorder 

charts, and Cardinal heat treating procedure, SOP 18.001 Rev. 6, dated 

December 6, 1994, "Heat Treating," as a part of an assessment of 

ACCUTECH's final report, "Investigation and Analysis of Suspect 

Fasteners, I' dated November , 1995. Furnace records were avai 1 ab1 e for 

the last five years of operation of the heat treating furnace. 

that time 1,418 separate furnace charges were processed. 

included quenching and tempering, stress re1 ieving, age hardening and 

solution annealing. 

steels, precipitation hardenable materials, and austenitic and 

martensitic stain1 ess steel s. 

During 

Materials included carbon and medium carbon alloy 

The operations 

Review of SOP 18.001 indicated that this procedure provided very little 

guidance on the operation of the heat treating oven. 

there were no instructions to control the material charge size and 

distribution in the furtiace, or thermocouple location with respect to 

the furnace charge. As demonstrated by ACCUTECH's investigation, these 

variables can have a significant effect on the heat treatment results. 

Inadequate or inconsistant control of these variables would tend to 

significantly expand the bounding load density and furnace time limits 

established by the investigation (conducted under controlled conditions) 

to include heat lots not previously reported. The probability that 

inadequate process control s exi sted is supported by the situations 

discussed above where defective fasteners were identified from heat lots 

that were considered acceptable based on ACCUTECH's investigation. 

The inspectors also noted that a significant amount of austenitic 

stainless steel was solution annealed in the furnace during this time 

period. 

ACCUTECH's investigation, a review of furnace logs and heat treatment 

charts showed some instances where the charge characteristics and 

Specifically, 

While this material was not specifically addressed in 

9 

17 
furnace times similar to those considered suspect in ACCUTECH’s 

assessment of alloy steel fasteners were also recorded for solution 

anneal ing treatment of stainless steel fasteners. Under these 

conditions, if some fasteners did not reach the solution annealing 

temperature, they could become severely sensitized as a result of the 

heat treatment cycle and, therefore, be highly susceptible to 

intergranular corrosion. There was no physical evidence to indicate 

that improper solution annealing had actually occured. 

d. Conclusions 

Failure to Inform the NRC of a Defect In the Rewired Time 

Heat lot TS7 was evaluated and determined to contain a defect as 

documented on the ACCUTECH Form 17.2, which was approved by the Branch 

Manager on March 28, 1996. However, the NRC was not notified by 

facsimile or telephone within two days as required by ACCUTtCH SOP 

17.002 and 10 CFR Part 21.21(~)(3). In addition, the written 

notification was not sent by ACCUTECH to the NRC until May 8, 1996 (41 

days after the ACCUTECH Form 17.2 evaluation) which does not meet the 

requirements of ACCUTECH SOP 17.002 or 10 CFR 21.21(~)(3) which require 

the written notification be provided to the NRC within 30 days. 

inspectors concluded that the ACCUTECH had failed to notify the NRC 

within two days of the responsible officer obtaining the information of 

a defect affecting a basic component and had failed to provide the 

written notification within the 30 days of the responsible officer 

obtaining information of a defect affecting a basic component, as 

required by 10 CFR 21.21. 

written notification within 30 days was identified as a violation of 10 

CFR Part 21. (Violation 99901307/96-01-01) 

The 

Failure to notify within two days and provide 

Failure to Adeauatelv Evaluate Heat Lot TS7 

During discussion with the NRC inspectors, ACCUTECH indicated that the 

failure:; of heat lot TS7 were possiblv due to commingling of non-heat 

treated fasteners with heat treated fasteners. This conclusion was 

based on the weight and length of heat treatment of the fasteners being 

well outside the bounds of suspect fasteners established by the test 

document4 in thc November 1995 Final Report. However, the April i, 

1996, WPPSS letter to ACCUTECH reporting the defective fasteners of heat 

lot TS7 indicated that, subsequent to the mechanical test performed on 

the defective fastener, a metallurgical examination of a defective 

fastener had confirmed that the fastener had been inadequately heat 

treated. 

adequately evaluated whether the identified defects in heat lot TS7 

could affect the validity of conclusions documented in the Final Report 

where the weight and length of heat treatment boundaries for suspect 

heat lots had been established, and to determine if any further review 

of the earlier research was required. The failure to perform an 

adequate review of heat lot TS7 was identified as a violation of 10 CFR 

Part 21. (Violation 99901307/96-01-02) 

Failure to Adeauatelv Evaluate Heat Lot K7 and Identify Heat Lot M2 

The production heat lot K7 had been modeled as one of the test heat lots 

In addition, the inspectors concluded that ACCUTECH had not 

10 

18 
0 

used to establish the weight and heat treatment boundaries documented in 

the Final Report. The test heat lot modeled after heat lot K7 had not 

produced any defective fasteners and, therefore, heat lot K7 had not 

been reported earlier as a suspect heat lot. ACCUTECH had not 

documented any evaluation activities for heat lot K7, as discussed 

above, and also indicated during the inspection that ACCUTECH had not 

reviewed the contradictory situation of the existence of a "good" test 

charge modeled after a production heat lot later deterrhined to contain 

defective material (heat lot K7). In addition, during the NRC 

inspection, the ACCUTECH review of the K7 furnace records indicated that 

an additional heat lot had been heat treated in the same furnace charge 

as heat lot K7. ACCUTECH indicated that the additional heat lot, M2, 

would also be considered suspect, since it had been heat treated 

together with the known defective heat lot, K7. On January 31, 1997, 

subsequent to the Loc;?letion of the NRC inspection, ACCUTECH notified 

the NRC of the suspect heat lot M2 by facsimile. The inspectors 

concluded that ACCUTECH had performed an inadequate review of the heat 

lot K7 defect as indicated by ACCUTECH failing to identify the 

additional suspect heat lot M2. The failure to perform an adequate 

review of heat lot K7 was identified as an additional example of 

Violation 99901307/96-01-02. 

Failure to Document the K7 Evaluation in Accordance With Procedure 

The evaluation of heat lot TS7 had been documented on SOP Form CF 17.2, 

dated March 28, 1996, however no form CF 17.2 was on file documenting 

the evaluation of heat lot K7. 

ACCUTECH had performed an evaluation of heat lot K7, (based on the 

notification of the defect), ACCUTECH had failed to document the 

evaluation in accordance with ACCUTECH SOP 17.002, which was identified 

as a Nonconformance to Criterion V, "Procedures," of 10 CFR Part 50, 

Appendix B. (Nonconformance 99901307/96-01-07) 

The inspectors concluded that although 

3.3 Material SuDDlied under ASME Code Rules 

3.3.1 Procedural Control s 

The controls for prucessing orders for material supplied under the rules 

of the ASME Code are described in ACCUTECH's Quality Systems Manual 

(QSM), which had been reviewed and accepted by the ASME. 

Edition, Revision 0 of this manual, dated 11/29/95, was in effect at the 

time of this inspection. 

manual. 

Second 

SOPS supplement various sections of the 

3.3.2 ImDlementation Review 

The inspectors selected a sampling of purchase orders from NRC licensees 

(emphasis on material supplied to PECO) for ASME Code material that 

ACCUTECH had processed within the last year. The documented bases for 

compliance with the applicable ASME Code requirements in the processing 

of these orders were reviewed with significant observations summarized 

bel ow: 

11 
PECO Purchase Order (PO) LS 605919, dated January 18, 1996, for 

forty four 1 1/4 inch by 8 inches long ASME SA-193, grade 87 bolts 

in accordance with the requirements of Section I11 of the ASME Code, 

1992 edition. 

ACCUTECH supplied this material under their ASME (QSC) and provided 

a Certified Material Test Report (CMTR) and a Certificate of 

Compl iancr :COC). The CMTR contained transcribed 1 adle analysis 

results for this material as well as mechanical property test 

resuits and heat treatment description. Review of supporting 

documentation indicated that ACCUTECH obtained the starting material 

in the form of cold finished SA-193 bars from Republic Steel Co., an 

approved vendor. ACCUTECH heat treated the material and performed 

two sets of chemical analyses, impact tests and tensile tests. 

Review of documentltion supplied to PECO indicated that ACCUTECH 

failed to include the mill heat analysis in documents provided to 

the customer. Paragraph NCA 3862.1 of Section I11 of the ASME Code 

requires that certifications for subcontracted chemical analyses, 

i ncl udi ng me1 ti ng mi 11 heat analyses, heat treatment , tests, and 

examinations to be furnished as an identified attachment to the 

CMTR. 

Section 6.3, “Control of Certifications” of the ACCUTECH QSM 

requires that CMTRs include the actual results of chemical and 

mechanical attributes, description of heat treatment, and 

nondestructive examination reports. It does not require, however, 

that mill heat analysis reports provided by approved suppliers be 

included with the CMTR. 

standard practice was to provide the original mill heat analysis 

reports only when specifically requested by the customer. Fai 1 ure 

to provide all of the documentation required by Paragraph NCA 3862 

of Section I11 of the ASME Code was identified as Nonconformance 

ACCUTECH staff acknowledged that their 

99901307/96-01-03. 

.1 

PECO Purchase Order LS 607801, dated March 28, 1996, for twenty 1/2- 

13 by 2 1/4 inches long ASME SA-193, grade 67 bolts in accordance 

with ASME Code, Section 111, Class 2. 

ACCUTECH supplied this material in accordance with their QSC and 

provided a CMTR and a COC. The CMTR contained identification of the 

item supplied, transcribed ladle analysis, check analysis, tensile 

properties, description of the heat treatment performed, QSC number, 

and statement certifying compliance with the applicable 

requirements. 

supplied, ACCUTECH’s QSC number and certified compliance with the PO 

requirements, including 10 CFR Part 21 and 10 CFRSO, Appendix B. 

The COC also described the heat treatment performed and certified 

that satisfactory macroetch test had been performed, visual 

inspection conducted per SOP 11.002, Rev. 9 (report attached) and 

that no welding had been performed on this material. 

The COC contained identification of the item 

12 

20 
ACCUTECH (then Cardinal) purchased the starting material, cold 

finished SA-193, grade 87 bar on September 25, 1990, from Republic 

Stee: Co., an approved supplier. On August 27, 1993, Cardinal 

issued a purchase order to Rooke Manufacturing, an approved 

supplier, to cut and hot-head this material and, on the same date, 

issued a purchase order to Phoenix Heat Treating, an approved 

supplier, to normalize, harden and temper the hot headed blanks to 

SA-193 requirements. 

this lot of material. 

Cardinal then performed 2 tensile tests on 

The inspectors noted that, contrary to the ASME Code requirements, 

neither the mill heat analysis report nor Phoenix Heat Treatment 

certification were furnished as attachments to ACCUTECH's CMTR that 

was provided to PECO. 

Nonconformance 39901307/96-01-03. 

This was identified as an example of 

Consumers Power Company Purchase Order G 0183280, dated December 7, 

1995, for 100, 1/2-13, SA-194, grade 2H nuts to the requirements of 

ASME Code, Section 111, Class 1, 1989 Edition with 1990 Addenda. 

and 

Entergy Purchase Order MP 96L093, dated November 8, 1995, for 76, 

1/2-13, SA-194, grade 2H nuts to the requirements of ASME Code, 

Section 111, NB. 

ACCUTECH supplied this material from existing stock. The nuts used 

to fill this order were from the same heat of material, and had the 

same manufacturing history, as discussed in Paragraph 3.4.3 of this 

report under Wisconsin Electric Power Co. Purchase Order 4500021861. 

ACCUTECH certified the material for the above procurements as being 

provided in accordance with their- ASME QSC. The certifications did 

not indicate that the material was supplied as "size excluded," 

therefore all of the NCA quality requirements were considered 

applicable for these procurements. The inspector noted that the 

mi17 heat analyses and mechanical test certifications by approved 

suppliers were ,rere not included with, or referenced on ACCUTECH's 

certification and that there was no evidence in the document package 

that a macroetch test was performed on the source material as 

required by ASME SA-194. 

insufficient documentation (audit reports) to demonstrate that the 

suppliers of this product were qualified to the requirements of NCA 

3800. 

Failure to provide adequate documentation to demonstrate compl iance 

with NCA 3800 and with the applicable requirements of the material 

specification was identified as an example of Nonconformance 

The inspector also noted that there was 

99901307/96-01-03. 

13 

21 
3.4 Non-Code Material SuDolied Under 10 CFR Part 50. ADDendix B. 

3.4.1 Procedural Control s 

ACCUTECH's controls for processing non-Code material that is to be 

supp:ied under the rules of 10 CFli Part 50, Appendix B are described in 

Addenda No. 1 to its ASME QSM described above. According to ACCUTECH, 

the intent of this Addenda is to modify sections of its QSM to assure 

compliance with 10 CFR 50, Appendix B, ANSI N 45.2, ANSI/ASME NQA-1 and 

other quality related standards as it applies to non-Code items. SOPs 

are used to supplement the controls described in Addendum 1 to the QSM. 

For example, ACCUTECH's commercial grade item (CGI) dedication program 

is described in SOPs 4.001, 9.001, 10.010, 11.001, and 11.002. 

3.4.2 Use of SamDlina in thL Dedication Process 

a. Observations and Findincis 

As part of the review of Open Nonconformance 99901076/94-01-03, 

(previously described in Section 2.1) , the inspectors reviewed the 

current sample plan methodology as part of the commercial grade 

dedication process in place at ACCUTECH. 

it is basically the same process that was reviewed during the NRC's 

December 1994 inspection at Cardinal, in that it places heavy reliance 

on visual and dimensional inspection to support the verification of lot 

homogeneity. No additional controls had been implemented in the 

ACCUTECH QA program to support the sampl i ng process rational e descri bed 

in Cardinal's August 30, 1995, letter. In fact, the nondestructive 

dimensional inspection sampling plan described in Section 3.13.1 of 

revision 8 of SOP 4.001, which references the EPRI guidelines, is 

actually less conservative than the sampling plan reviewed (Revision 6 

of Cardinal Standard Practice 4.001) during the December 1994 

inspection. 

The inspectors determined that 

ACCUTECH representatives stated that they did not have a documented 

basis to support the information that was previously described to the 

NRC in the August 3b, 2S95, letter concerning sampling. ACCUTECH had 

recently written a "white paper'' that described proposed changes to the 

ACCUTECH sample plan methodology for testing and examination. 

document was dated November 7, 1996, but had not as yet been implemented 

as part of the ACCUTECH QA program. The inspectors did not review this 

document during the inspection since it had not been approved and 

implemented as part of the QA program and did not reflect the process 

for the dedication activities that were reviewed. 

This 

b. Conclusions 

The inspectors determined that ACCUTECH places heavy reliance on visual 

and dimensional inspection to support the verification of lot 

homogeneity. 

homogeneity, ACCUTECH then utilizes the ASTM A-325 shipping lot sampling 

plan for destructive testing (material chemistry and mechanical 

Based upon this method for verification of lot 

14 

22 
properties) and the EPRI guide1 ines for nondestructive testing 

(dimensional). The inspectors determined that visual inspection for 

shipping damage and manufacturing defects can not assure that all items 

in the same product lot were manufactured from the same heat of material 

or were heat treated under the same conditions and that, as discussed in 

Inspect ion Report 99901076/94-01 and re1 ated correspondence, the use of 

ASTM A-325 shipping lot sampling plan is inappropriate for this 

appl i cat i on. 

Nonconformance 99901076/94-01-03 is considered Open, and requires 

additional response from ACCUTECH to address the original concerns 

documented in this nonconformance and discussed further with the NRC 

inspectors during the November 1996 inspection. 

3.4.3 ImDlementation Review 

The inspectors selected a sampling of recent POs from NRC licensees 

(emphasis on material supplied to PECO) for non-Code items ordered to 

the quality requirements of 10 CFR Pdr t 50, Appendix 6. 

basis for compliance with the applicable Appendix 6 and customer PO 

requirements were reviewed with significant observations summarized 

bel ow: 

The documented 

PECO Purchase Order LS 606821, dated February 15, 1996, for twenty 

four 7/8-9 by 2 1/4 inches long ASTM A-325 bolts. 

ACCUTECH's certification for this material included heat number 

identification, transcribed ladle analysis, description of the heat 

treatment, hardness test results, and certification that this 

material was manufactured in accordance with ACCUTECH's QSM, Second 

edition, Revision 0, Addenda 1, and that the material supplied 

complies with the PO requirements. 

ACCUTECH drew this material from existing stock of 7/8-9 by 7 1/4 

inch bolts, performed chemical analyses and hardness tests on two 

samples, and cut the required number of bolts to the 2 1/4 inch 

length specif'2d in the PO. 

the bolts from Mitsuboshi Sangyo Co. Ltd., who obtained the starting 

material from Azuma Steel Co., Chiba, Japan. The inspector noted 

that the Azuma CMTR was dated November 19, 1980, but contained a 

typed statement that the material was manufactured and supplied in 

accordance with Azuma Steel Co. QA program that had been surveyed 

and approved by Cardinal on April 1, 1982. ACCUTECH staff could not 

explain the apparently erroneous date or dates on the Azuma CMTR and 

could not produce an audit report to indicate if or when an audit of 

Azuma Steel Company was performed. The inspector noted that, if the 

steel was produced in November of 1980, Cardinal did not have 

adequate assurance that the steel manufacturing process complied 

with a quality assurance program that they approved in April, 1982. 

ACCllTECH (then Cardinal) F2d purchued 
The inspector further noted that the llmited testing performed by 

ACCUTECH (two hardness tests and two chemical analyses) did not 

provide an adequate basis for certifying this material as conforming 

to the requirements of 10 CFR 50, Appendix 6. 

Subsequent to the first phase of the NRC inspection, ACCUTECH issued 

Shop Order PC 327982 for additional inspections consisting of two 

hardness :--:s, two proof load and wedje tests, and two chemical 

analyses on samples from 325 bolts from the same heat lot which 

remained in stock. The results from these tests (satisfactory) were 

included in a supplemental certification issued to PECO on December 

26, 1996. Failure to provide documented basis that material 

certified as comforming to the requirements of 10 CFR Part 50, 

Appendix B actually conforms to these requirements was identified as 

Nonconformance 99901307/96-01-04. 

Wisconsin Electric Power Co. Purchase Order 4500021861, dated 

November 1, 1996, for 100 1/2-13 SA-194, Grade 2H heavy hex nuts. 

ACCUTECH supplied this material from ex;sting stock. ACCUTECH (then 

Cardinal) had purchased the nuts from Unytite Fastener Mfg. Co. 

Ltd., Akashi Factory, Japan, who apparently obtained the starting 

material f,-om Kobe Works, Kobe Steel Ltd. through a supplier, MituiUnitaito-Nishikobe. Kobe issued a "Mill Sheet" for the starting 

material on June 30, 1984. The only pertinent information appearing 

on the mill sheet was the heat number (KC8972), material 

specification (KCH45KT-W), and chemical analysis. A stamp on the 

mill sheet indicated that it had been accepted by Unytite on July 

10, 1984. Unytite issued an "Inspection Certificate" on the 

finished product (252,717 nuts) on June 18, 1984, apparently before 

they received the mill sheet from Kobe and apparently before the 

acceptance of this document by Unytite. 

the material by a heat number traceable to the Kobe mill sheet and 

a lut number. It contained the ;-esults of chemical analysis 

(transcribed from the Kobe mill sheet), and indicated that five 

proof load, cone stripping, and hardness tests had been performed 

with satisfactory results. The certificate contained a quality 

program qtatement stating that the material was produced in 

accordance with Unytite's QA program M-00 Rev. 4, dated April 14, 

1983, that had been reviewed and accepted by Cardinal on October 7, 

1983. 

stating that material heat KC8972 had been produced in accordance 

with Kobe QA program, Rev. 6, dated January 20, 1983, that had been 

surveyed and approved by Cardirial on October 6, 1983. 

ACCUTECH supplied this material as having been manufactured in 

accordance with the current edition of their their QA manual and 10 

CFR Part 50, Appendix B. 

This certificate identified 

The file also contained a letter from Kobe Steel, Ltd., 

16 

24 
The inspector identified the following concerns with the 

documentation described above: 

3.5 

a) 

the starting material produced by Kobe Steel Co. complied with the 

internil quality requirements of ’,he applicable material 

specification (SA-194). Specifically, SA-194 requires a macroetch 

test in accordance with ASTM E-381 on each bar lot produced by the 

supplying mill. The Kobe mill sheet contained no reference to 

indicate that the material met the requirements of SA-194, and the 

Unityte inspection certificate did not contain any reference to a 

macroetch test. 

None of the documentation reviewed provided verification that 

b) Unityte apparently received the starting material from Kobe 

Steel and completed €heir manufacturing and testing activl$ies 

before receiving Kobe’s mill sheet for this material. 

c) 

their bases for accepting the valiaity of such surveys of either 

Unytite or Kobe which would have approved these companies as 

qual i f i ed suppl iers. 

ACCUTECH could not produce re7orts of Cardinal’s surveys or 

This item was identified as an example of Nonconformance 

99901307/96-01-04 

Nuclear Utilities Procurement Issues Committee (NUPIC) Audit of ACCUTECH 

The inspectors reviewed the Comanche Peak Steam Electric Station TU Electric 

(TUE) QA Audit Report QAA-96-010 of ACCUTECH dated May 23, 1996. The audit, 

conducted April 29 through May 2, 1996, was led by TUE and performed in 

accordance with the requirements of the TUE QA program, under the auspices of 

NUPIC. The audit also included representatives of Iowa Electric Services, 

Houston Light and Power, and Northern States P2wer. 

The inspectors reviewed the applicable sections in the TUE audit report that 

addressed the issues from NRC Inspection Report No. 99901076/94-01, commercial 

grade dedication 2nd sampling relating to commercial grade dedicatiorl. It 

appears that the audit reviewed the applicable portions of the ACCUTECH 

program as it relates to sampling as part of commercial grade dedication, 

however, the audit describes the process just as Cardinal/ACCUTECH letters 

described it to the NRC after the 1994 inspection. 

the NUPIC audit verified or attempted to verify the ACCUTECH basis for the 

sampling plans uti1 ized for destructive and nondestructive testing. 

It does not appear that 

Also, when reviewing the documentation on the NUPIC audit checklist, the 

inspectors did not identify any evidence that the auditors reviewed ACCUTECH’s 

rationale for verifying lot homogeneity for finished fasteners that are 

purchased from nonapproved suppliers. This method, in turn formulates the 

basis for ACCUTECH’s selection of the CGI sampling plans. 

17 

25 
In conclusion, the inspectors determined that the NUPIC audit did not address 

the ACCUTECH basis for their rationale for verification of lot homogeneity for 

finished fasteners that are purchased from nonapproved suppliers. Also, the 

audit apparently did not review the adequacy of documentation included with 

ACCUTECH's CMTRs for safety-related material suppl ied to 1 icensees under the 

rules of the ASME Code (See section 3.3 of this report). 

3.6 Corrective Action Activities 

The inspectors reviewed QSM Section 17, "Corrective Action," Revision 0, dated 

November 29, 1995 and SOP 17.001, "Corrective Action Report," Revision 5, 

dated November 15, 1995. The purpose of the SOP is to establish a method of 

identifying and establishing the responsibilities for initiating, processing, 

and resolving internal or external Corrective Action Reports (CARs). 

Section 3.1 of SOP 1/.001 lists a failure of the QA program to comply with an 

approved regulation or commitment (ASME, ASTM, ANSI, MIL-STD etc.), as one of 

the conditions that defines conditions adverse to quality. Section 3.2 states 

that conditions adverse to quality may be identified during an Internal Audit, 

Management Audit, Supplier Audit or by other means. NRC Inspection Report No. 

99901076/94-01 identified two instances (as Nonconformances) where 

Cardinal/ACCUTECH failed to meet the requirements of 10 CFR Part 50, Appendix 

B. These NRC Nonconformances appear to meet the threshold as a "condition 

adverse to quality" as defined in SOP 17.001 but were not entered into the 

corrective action process through the initiation of a CAR. 

When the inspectors requested to review the CARs written to address the NRC 

Nonconformances, ACCUTECH stated that CARs were not written for the evaluation 

and disposition of the two NRC findings. 

documented by ACCUTECH to support the information described in the various 

letters submitted to the NRC in response to the two Nonconformances, beyond 

the letters themselves. This made it difficult to review the actions to 

prevent recurrence for Nonconformance 99901076/94-01-04 described in 

Cardinal's January 30, 1995 letter. 

Therefore, nothing had been 

ACCUTECH's failure to identify the two NRC Nonconformances as conditions 

adverse to quality i.e. failures to comply with a regulation (10 CFR 59, 

Appendix B), and enter them into the corrective action process through the 

initiation of CARs, as required by SOP 17.001, was identified as 

Nonconformance 99901307/96-01-05, 

3.7 Trendinq Activities at ACCUTECH 

The inspectors reviewed SOP 16.006, "Trending," Revision 1, dated November 15, 

1995. 

and recommending appropriate action for quality trending activities at 

ACCUTECH. This process produces a Quality Trend Report that provides details 

of the number and deposition of External Nonconformance Reports and percentage 

of rejection of items received by suppliers; Internal Nonconformance Reports, 

by source; customer returns; status of open Nonconformance Reports; and a QC 

Production Log. 

customer inquiries and an internal performance report breakdown that is used 

The purpose of this procedure is to establish a method of identifying 

The Quality Trend Report also includes information on 

26 

18 
to identify any problem areas. 

a minimum of every six months, but were closer to being issued Gn a monthly 

basis. 

These reports are issued by the QA manager as 

A review of Quality Trend Reports for the last three years indicated that the 

trending pr ic,ram for ACCUTECH suppl ied p.*oductj and material/items received 

from vendors is effective in providing valuable information to the quality, 

technical, and manufacturing departments concerning the status of the QA 

program and important trends in the overall quality process at ACCUTECH. 

However, when reviewing the corrective actions for Nonconformance 

99901076/94-01-03, the inspectors did not identify any objective evidence in 

the Quality Trend Reports to support the correlation of trending data of NCRs 

to the selection of testing and inspection sample plans as described in 

ACCUTECH's August 30, 1995, letter concerning the sample plan nonconformance. 

3.8 ACCUTECH Use of Cardinal Industri a1 Products Warehouse Inventory 

a. InsDection ScoDe 

The inspectors reviewed SOP 22.001, "Transference of Material and Associated 

Documentation Between CIPC and ACCUTECH," Revision 2, dated November 15, 1995, 

to determine the process used by ACCUTECH to verify that warehoused material 

inherited from the previous owner, Cardinal, would meet the current QA program 

material qualification requirements for supply as ASME Code and safety-related 

materials. 

documents are properly reviewed and approved prior to use by ACCUTECH as ASME 

Code and safety-related materials. 

The purpose of SOP 22.001 is to ensure that materials and 

b. Observations and Findinqs 

The inspectors were told by ACCUTECH that this procedure was originally 

implemented in 1985 to handle bankruptcy proceedings material sales and was 

not really intended for regular program iiplementation. However, during 

review of various material data packages during the inspection, the inspectors 

determined that the process described in SOP 22.001, including use of the 

Document Review Checklist, was still being implemented until approximatelv 

June 1995. B&G assumed ,ontrol of Cardinal in July 1995. 

The inspectors requested material documentation data packages for old Cardinal 

material from warehouse stock that had been recently sold as safety-related or 

as ASME Code to nuclear utilities. The inspectors were told that they could 

not identify the entire population of old Cardinal material that was currently 

being sold by ACCUTECH, but they were currently in the process, as time 

permitted, to identify and determine full material qualification for all the 

Cardinal inventory in the warehouse. The inspectors then requested the data 

packages for Cardinal material still in inventory. 

The inspectors began with a review of the issue described in Section 3.4.3 of 

this report, concerning PECO PO LS 606821, dated February 15, 1996. 

package was chosen since the items that were sold to PECO in 1996 were from 

existing warehouse stock that was acquired from Cardinal after the B&G 

purchase of Cardinal in 1995. The inspectors reviewed the data package for 

This 

19 

27 
PECO PO LS 606821 to determine compliance to SOP 22.001 as it relates to the 

material shipped to PECO. The inspectors determined that for the material 

shipped to PECO, ACCUTECH could provide no evidence of any document review as 

required by SOP 22.001 to be performed and recorded on the Document Review 

Checklist form. 

Section 3.4.3 of this report. 

The specifics of the material sold to PECO is described in 

It was also determiK-” +hat additional warehouse inventory from the same 

material heat lot sold to PECO had been sold to additional utilities such as 

Union Electric (PO No. 093879 dated April 15, 1996) as safety-related 10 CFR 

50, Appendix B material. The inspectors then inquired if ACCUTECH addressed 

the problems with the PECO order from a generic standpoint to determine if 

other material from the same heat lot number was sold to other utilities. The 

inspectors were told that subsequent to the first phase of the NRC inspection, 

ACCUTECH issued Shop Order OC 327982 for additional inspections consisting of 

two hardness tests, two proof load and wedge tests, and two chemical analyses 

on samples from 325 bolts of the same heat lot which remained in stock. 

results from these tests (satisfactory) were included in a supplemental 

certification issued to PECO on December 26, 1996. ACCUTECH ststed that they 

also conducted additional testing for other data packages where material was 

supplied from the above heat lot, based upon the sample plans identified in 

their SOPS. The NRC has stated, in previous correspondence, that these sample 

plans did not providc an adequate basis for certifying material as conforming 

to the requirements of 10 CFR 50, Appendix B. 

The 

ACCUTECH did not initiate a Corrective Action Report (CAR) and did not 

evaluate the finding for its generic applicability for other material sold 

from the same heat lot and for other types of old Cardinal material sold from 

warehouse stock. 

The inspectors reviewed a second package for items that were also supplied 

from warehouse stock inherited from Cardinal and sold to utilities in 1996. 

This package was for material that was supplied to Wisconsin Electric under 

PO 4500021861, dated November 1, 1996. The technical details of this PO and 

supply is described in Section 3.4.3 of this report. 

previous example, the inspectors determined that for the material shipped to 

Wisconsin Electric, ACCUTXH could provide no evidence of any document review 

as required by SOP 22.001 to be performed and recorded on the Document Review 

Checklist form. 

same warehouse stock, that represents the same heat of material, was sold to 

numerous utilities since mid-1995. 

Again, as in the 

The inspectors then determined that similar material from the 

C. Concl us i ons 

The inspectors concluded that ACCUTECH had not implemented the requirements of 

SOP 22.001 since July 1995 and failed to ensure that material and documents 

are properly reviewed and approved for old Cardinal material sold from 

warehouse stock to utilities as ASME Code or safety-related 10 CFR 50, 

Appendix B. At the time of the inspection, ACCUTECH could not identify the 

entire population of old Cardinal material that was currently in warehouse 

stock and being sold by ACCUTECH with questionable material certification. 

This issue is identified as Nonconformance 99901307/97-01-06. 

20 

28 
The inspectors also concluded that subsequent to the first phase of the 

inspection in November 1996, ACCUTECH failed to initiate a Corrective Action 

Report (CAR) and did not evaluate the finding concerning the material supplied 

under PECO PO LS 606821 for its generic applicability for other material sold 

from the same heat lot or for other types of old Cardinal material sold from 

warehouse stock. 

3.9 Source InsDection Activities 

The inspectors revi ewed QSM Section 8.0, "Control of Purchased Materi a1 and 

Services," Revision 0, dated November 29, 1995 and SOP 8.005, "Performance of 

Source Inspection," Revision 3, dated November 15, 1995. The purpose of the 

SOP is to establish the method for performing and documenting source 

inspections. 

The inspectors reviewed the 1996 Source Inspection Log and selected the source 

inspections performed at Nortec Specialty Steel (Nortec), Lubbock Texas and 

Atlas Specialty Steel (Atlas), Welland, Canada for review. Two inspection 

trips were made to both facilities in 1996 in support of two different 

ACCUTECH purchases, one to fill a PO received from ABB Combustion Engineering 

Nuclear Power (ABB CENP) and one to fill a PO from a Korean company. 

issued to ACCUTECH were for 3.600-0 A276-93 type 304 stainless steel round 

bars, but for different quantities and length. 

bars 310 inches long. 

210 inches long and stated on their PO that it was for use at commercial 

nuclear power plants. 

ACCUTECH placed the POs for the round bar with Nortec. 

to Atlas, who as the original steel mill supplied the bars to Nortec. Nortec 

performed the straightening and grinding prior to delivery to ACCUTECH. 

inspectors reviewed the Source Inspection Plans (SIPs) for Atlas (96-010, 

dated September 17, 1996, and 96-021, dated November 19, 1996) and for Nortec 

(96-013, dated October 9, 1996, and 96-018, dated Gecember 4, 1996). The SIPs 

documented ACCUTECH's verification and comments for the inspection attributes 

identified on the SIP. 

Both POs 

The Korean order was for 78 

The PO issued by ABB CENP to ACCUTECH was for 85 bars 

Nortec subcontracted 

The 

The inspectors discussed wi~;, ACCUTECH the fact that the SIP for both Atlas 

and Nortec did not include documentation for heat treatment as an inspection 

attribute. Considering the fact that the AB8 CENP Purchase Specification for 

Quality Requirements listed heat treatment as an item requiring certification 

by ACCUTECH, heat treatment should have been included as an inspection 

attribute. 

The inspectors concluded, with the exception of the failure to include the 

review of heat treatment activities and records at both facilities, the 

implementation of source inspection activities appeared to document the 

verification of the appropriate inspection attributes for the scope of supply. 

21 

29 
3.10 Lead Auditor Qualification 

The inspectors reviewed QSM 8.0, "Control of Purchased Material and Services, I' 

Revision 0, dated November 29, 1995, Section 8.5, and SOP 20.001, 

"Qualification of Auditors and Lead Auditors," Revision 2, dated 

November 15, 1995. The purpose of the SOP 20.001 is to ensure that personnel 

performing either internal or suppl ier audi tlsurveys are adequately qual i fied 

to perform those quality activities. 

During the review of source inspection activities described aoove, the 

inspectors noted that the Quality Director, B&G Manufacturing had performed 

source inspections activities on behalf of ACCUTECH in 1996. According to 

the ACCUTECH QA Manager, this individual had been qualified as an lead auditor 

per the requirements of the ACCUTECH QA program and performed all audit and 

source inspection activities under the ACCUTECH QA program. 

was verified through review of auditor qualification records. 

This information 

The inspectors discussed with ACCUTECH that this personnel interface between 

B&G and ACCUTECH is not described in QSM Section 2.0, "Organization," or 

identified on the ACCUTECH Organizational Chart. Based on the review of 

auditor qualification documentation, with the exception of the observation 

concerning the use of B&G personnel for ACCUTECH audits, audit activities 

performed by ACCUTECH were determined to be adequately implemented. 

3.11 Entrance and Exit Meetinqs 

Entrance and exit meetings were held for each segment of the two-part 

inspection. In the entrance meetings, held on November 12, 1996, and on 

January 28, 1997, the NRC inspectors discussed the scope of the inspection and 

outlined the areas to be inspected. In the exit meetings, held on November 

14, 1996, and on January 30, 1997, the inspectors discussed their findings and 

concerns with ACCUTECH management. 

22 

30 
PARTIAL LIST OF PERSONS CONTACTED 

William A. Edmonds, President, B&G Manufacturing Company 

Jim Sands, Quality Director, B&G Manufacturing Company 

Roy P. Reindl, Branch Manager 

Bill Sul han , Qual i ty Assurance Manager. 

Greg Kel I er, Marketing & Technical Services Manager 

ITEMS OPENED, CLOSED, AND DISCUSSED 

ODened 

9990 130 7/96-0 1 -0 1 

99901307/96-01-02 

99901307/96-01-03 

99901307/96-01-04 

99901307/96-01-05 

99901307/96-C1-06 

99901307/96-01-07 

VI0 

VI0 inadequate Part 21 evaluation 

NON 

NON inadequate verification of conformance 

NON inadequate corrective action 

NON inadequate :wiew sf material in Warehouse stock 

NON inadequate documentation of Part 21 evaluation 

exceeding Part 21 notification timeliness limit 

incomplete documentation supplied to customer 

Discussed 

Nonconformance 99901076/94-01-03 remains open and requires a response. 

23 

31 
UNITED STATES 

NUCLEAR REGULATtd3Y COMMIS',. 2') 

WASHINGTON, D.C. 2055WWl 

*+*** 

February 27, 1997 

Mr. Calvin R. Hastings 

President and Chief Executive Officer 

Caldon, Inc. 

2857 Banksville Road 

Pittsburgh, PA 15216 

SUBJECT: NRC INSPECTION REPORT 99901311/97-01 

Dear Mr. Hastings: 

On January 21, 1997, the U.S. Nuclear Regulatory Commission (NRC) completed an 

inspection at the Caldon, Inc. facility. The enclosed report presents the 

results of that inspection. 

The inspection was conducted to ascertain your corrective actions relating to 

the Point Beach Nuclear Plant, Unit 2, Licensee Event Report (LER) 94-001-01, 

"Potential Feedwater Flow Measurement Inaccuracies," dated July 29, 1994, 

regarding degradation in the signals from the leading edge flow meter (LEFM) 

Model 601, used to measure main feedwater flow in nuclear power plants. We 

also assessed whether you informed other licensees of this event, and your 

screening for Part 21 applicability. 

During this inspection, the inspector determined that Westinghouse Electric 

Corporation manufactured the Model 601 and that you purchased the LEFM 

technology from Westinghouse but did not sell the Model 601 to the industry. 

Your corrective actions pertinent to LER 94-001-01 involved upgrading the LEFM 

Model 601 to Model 8300 at Point Beach to improve detection of degraded 

signals from LEFM transducers. 

inaccurate readings because the LEFM Model 601 was sold as a commercial-grade 

item and because you do not have a Part 21 program at your facility. 

You did not alert licensees about Model 601 

The inspector observed that you did not have information on which licensees 

operated Model 601 in their facilities. Based on your sales of Model 8300 to 

facilities, you currently believe that the Model 601 is no longer used in the 

industry . 

During this inspection, the NRC inspector did not identify any instances in 

which your activities failed to meet NRC requirements for the areas inspected. 

Therefore, no response to this letter is required. 

32 

In accordance with io CFR 2.790, a copy of this letter and its enciosure will 

be placed in the NRC Public Document Room. 
C. Hastings -2- 

If YOU have any questions concerning this inspection, we will be pleased to 

discuss them with you. 

Sincerely, 

ORIGINAL SIGNED BY 

Robert M. Gallo, Chief 

Special Inspection Branch 

Division of Inspection and Support Prograrr: 

Office of Nuclear Reactor Regulation 

Docket No. 93901311 

Enclosure: Inspection Report 9990131 1/97-01 
Report No: 

Organization: 

Contact: 

Nuclear Industry 

Act i v i ty : 

Date: 

Inspector: 

Approved by: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

99901311/97-01 

Cal do'n, Inc. 

C.1vin R. Hastings 

President and Chief Executive Officer 

41 2 /34 1 - 992 0 

Flow monitoring systems 

January 21, 1997 

Ani1 S. Gautam, Senior Engineer 

Gregory C. Cwalina, Chief 

Vendor Inspection Section 

Special Inspection Bra-nch 

Division of Inspection and Support Programs 

Enclosure 
1 INSPECTION SUMMARY 

During this inspection, the NRC inspector reviewed activities associated with 

the adequacy of Caldon's corrective actions in response to the Point Beach 

Nuclear Plant, Unit 2, Licensee Event Report (LER) 94-001-01, "Potential 

Feedwater Flow Measurement Inaccuracies, I' dated July 29, 1994, regarding 

degradation of signals from the leading edge flow meter (LEFM) Model 601 

transducers when installed in the Point Beach Nuclear Plant, Unit 2, and 

whether Caldon informed other licensees of problems with the Model 601. 

inspector also assessed Caldon's screening and reporting of issues for Part 21 

appl icabi 1 i ty . 

The 

The inspection bases were as follows: 

Appendix By "Quality Assurance Criteria for Nuclear Power Plants and Fuel 

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

Reaulations (10 CFR Part 50). 

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

LER 94-301-01 , "Potential Feedwater Flow Measurement Inaccuracies, Point 

Beach Nuclear Plant, Unit 2," dated July 29, 1994. 

For the areas inspected, the inspector did not identify any instances in which 

Caldon's practices did not conform to NRC requirements. 

2 

3 

STATUS OF PREVIOUS INSPECTION FINDINGS 

This was the first NRC inspection of Caldon. 

INSPECTION FINDINGS AND OTHER COMMENTS 

3.1 ResDonse to Point Beach Nuclear Plant. Unit 2, LER 94-001-01 

a. InsDection ScoDe 

The inspector examined Caldon's corrective actions in response to LER 94- 

001-01, ascertained whether Caldon informed licensees of corrective 

actions, and evaluated Caldon's screening and reporting of issues for 

Part 21 appl i cabi 1 i ty . 

b. Observations and Findinas 

About 1980, Westinghouse Electric Corporation manufactured and sold the 

LEFM Model 601 to licensees for correction of the feedwater flow venturi 

readings. The correction for the venturi was needed because water 

corrosion and fouling in the throat of the venturi had resulted in 

reactor power being reduced. The feedwater flow readings are used to 

calibrate the power range instruments, calculate reactor coolant system 

flow, and as the basis for the over temperature delta temperature and 

over pressure delta temperature setpoint calculations. 

2 

35 
Hodel 601 was a digital ultrasonic flow measurement system consisting of 

a pressure-retaining spool piece (pipe section) installed in the main 

feedwater header. 

temperature transducers (see Figure 1) to measure feedwater flow 

velocities along four ultrasonic paths. The feedwater flow measurement 

was based on the difference in transit times of upstream and downstream 

ultrasonic pulses between the transducers. The LEFM main feedwater 

reading was considered by licensees to be more reliable than the venturi 

reading and was used to determine a correction factor applied to the 

feedwater venturi reading to account for fouling of the venturi. 

correction factor was normally calculated at the beginning of each 

operating cycle and verified approximately once a month to ensure no 

significant changes had occurred during the cycle. 

Each spool piece contained four pairs of highThe 

The LEFM Model 601 had +he potential to develop inaccurate flow readinas 

because of inadequate tietection of any degraded transducer signals. 

June 1989, Caldon purchased the LEFM technology from Westinghouse but did 

not sell the Model 601 to the industry. Caldon used the LEFM technology 

to develop an upgraded LEFM Model 8300 capable of improved detecting and 

compensating for degraded transducer signals. Between 1992 and 1996, 

Caldon upgraded the Model 601 to Model 8300 at several facilities, 

including Point Beach Units 1 and 2. 

In 

Wisconsin Electric Power Company (WEPC) identified conditions which 

indicated that feedwater flow in Unit 2 may have been underestimated 

since the beginning of Cycle 20 on October 30, 1993. WEPC established a 

team of engineers to evaluate the problem, including contracting Caldon 

to conduct an assessment of the Point Beach Nuclear Plant feedwater flow 

measurement system. The assessment centered around the accuracy of the 

LEFM Model 601. On December 15-17, 1993, Caldon temporarily installed 

the LEFM Model 8300 to provide an independent measure of the feedwater 

flow in both units. Caldon's assessment indicated that the Model 601 was 

accurately measuring Unit 1 feedwater f:ow but was underestimating Unit 2 

feedwater flow by approximately 1%. Other factors evaluated by WEPC 

increased this estimation to approximately 2%. 

issued LER 94-001-01 which identified underestimation of feedwater flow 

by approximately 2% at Point Beach Unit 2 and believed this problem to be 

caused by degradat'qn of signals from the LEFM Model 601 transducers. 

About April 1995, Caldon permanently upgraded the LEFM Model 601 to Model 

8300 at Point Beach. 

classified as non-safety-related at Point Beach and not included in its 

calibration program. 

On July 29, 1994, WEPC 

The inspector determined that the LEfMs were 

The inspector observed that Caldon did not notify other licensees about 

the Model 601 problems at Point Beach, and did not have information on 

which licensees operated Model 601 in their facilities. Caldon stated 

that it did not inform licensees of the event because the LEFM Model 601 

was sold to licensees as a commercial-grade item, and because Caldon did 

not have a Part 21 program. Caldon stated that Westinghouse did not tell 

Caldon the names of licensees who had purchased the Model 601 from 

Westinghouse, but that based on its sale of the upgraded LEFM Model 8300 

to licensees it currently believes that the Model 601 is no longer used 

in the industry. 

3 

36 
The inspector contacted a representative of Westinghouse who informed the 

inspector that Westinghouse sold the LEFM Model 601 commercial-grade to 

the following 1 icensees: Northern States Power Company (Prairie Island 2), 

Consolidated Edison Company (Indian Point 2), Wisconsin Electric Power 

Company (Point Beach 1 and 2), TU Electric Company (Comanche Peak 1 and 

2), Southern Nuclear Operating Company (Farley 1 and 2), Duquesne Light 

Power Company (Beaver Valley l), and Kansas Gas & Electric Company (Wolf 

Creek). 

Caldon’s sales of Model 8300 to licensees, and determined that the Model 

601 LEFMs purchased by the aforementioned 1 i censees’ were no 1 onger 

operable in those nuclear plants. 

The inspector contacted 1 icensees and reviewed evidence of 

The inspector observed that Caldon did not evaluate the LEFM 601 problem 

in accordance with Part 21. Caldon stated that it did not evaluate under 

Part 21 because both LEFM models were sold as commercial-grade items to 

licensees. 

c. Conclusions 

In general , Cal don’s corrective act ions were adequate. 

and discussions with Caldon and Westinghouse, Model 601 is no longer used 

in the industry. 

problems because it was sold commercial-grade. 

Based on records 

Caldon did not inform licensees of the Model 601 

3.2 Entrance and Exit Meetinqs 

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

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

established interactions with Caldon’s management. 

on January 21, 1997, the inspector discussed his observations. 

In the exit meeting 

4 

6 

39 
February 10, 1997 

UNITED STATES 

IkJCLEAR REGULATORY COMMISSION 

WASHINGTON, ... C. -1 

Mr. Andrew Pike 

President 

Ellis 81 Watts 

Division of Dynamics Corporation of America 

4400 Glen Willow Lake Lane 

Batavia, OH 45103 

SUBJECT: NRC INSPECTION REPORT 99901308/96-01 AND NOTICE OF VIOLATION AND 

NOTICE OF NONCONFORMANCE 

Dear Mr. Pike: 

On !Member 6, 1996, the U.S. Nuclear Regulatory Commission (NRC) completed an 

inspection at the Ellis & Watts (E&W) facility. The enclosed report presents 

the results of that inspection. 

The inspection was conducted to ascertain specific attributes and 

implementation of your qual i ty assurance (QA) program, and whether 1 icensees 

effectively monitored your control of quality for safety-related heating 

ventilation and air conditioning (HVAC) systems and associated spare parts 

purchased by 1 icensees for nuclear power plants. We assessed your commerci a1 - 

grade dedication activities, your screening of issues for Part 21 

applicability, and your monitoring of the control of quality by your subvendors. 

During this inspection, the NRC inspector determined that certain of your 

activities appeared to be in violation of NRC requirements. Specifically, YOU 

failed to adopt appropriate procedures to ensure the evaluation and proper 

reporting of deviations, as required by 10 CFZ Part 21. In your December 30, 

1996, letter to the NRC, you reported taking appropriate corrective actions to 

address evaluation of deviations, including requiring pertinent training for 

appropriate employees. No further response is required. 

In general, the QA manual and procedures were adequate, and your staff 

competent. However , the NRC inspector determined that the imp1 ementat ion of 

your quality assurance program did not meet certain NRC requirements imposed 

on you by your customers. Specifically, you did not verify the adequacy of 

design of certain HVAC components and spare parts supplied to licensees, and 

take effective corrective actions for deficiencies in your commercial-grade 

dedication program, as required by 10 CFR Part 50, Appendix B. 

These nonconformances are cited in the enclosed Notice of Nonconforwnce 

(NON), and the circumstances surrounding them are described in detail in the 

enclosed report. 

should follow the instructions specified in the enclosed NON when preparing 

your response. 

You are requested to respond to the nonconformances and 

40 
A. Pike c 

-cIn addition, the inspector determined that licensee monitoring of your qual ity 

assurance program and commercial-grade dedication activities was adequate and 

effective. Licensees indicated that they continue to have concerns regarding 

your Val idation of certificates of conformance and certified material test 

reports. Lack of validation during dedication could result in unqualified 

HVAC equipment or spare parts. Reassessment of dedication activities appears 

war ranted. 

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

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

Room. 

Sincerely, 

ORIGINAL SIGhTD BY 

Robert M. Gallo, Chief 

Special Izipectim Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Docket No. 99901308 

Encl osures : 1. Notice of Vi ol at i on 

2. Notice of Nonconformance 

3. Inspection Report 99901308/96-03 

41 
Ellis & Watts (E&W) 

Batavia, Ohio 

NOTICE OF VIOL82TION 

Docket No. : 99901308 

During a U.S. Nuclear Regulatory Commission (NRC) inspection from December 2 

through 6, 1996, a violation of NRC requirements was identified. 

accordance with the "General Statement of Policy and Procedure for NRC 

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

In 

Section 21.21 of Title 10 of the Code of Federal Requlations (10 CFR Part 

21.21), "Notification of failure to comply or existence of a defect and its 

evaluation," requires, in part, that each corporation subject to the 

regulations adopt appropriate procedures to ensure the evaluation acid proper 

reporting of deviations and failures to comply. 

Contrary to the above requirement, E&W's procedure QA-112, "Compliance to 

Regulation 10 CFR Part 21 Procedure," Revision 3, failed to address the 

identification or evaluation of deviations. 

a1 erted to the need to ident i fy devi at i ons from safety-re1 ated purchase order 

requirements, and to evaluate the deviations to determine if they could become 

defects. Based on E&W's December 30, 1996, letter to the NRC, the QA manager 

revised procedure QA-112 to address evaluation of deviations and require 

pertinent training for appropriate employees. No further response is required 

As a result, personnel were not 

(99901308/96-01-01), 

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

Dated at Rockville, Maryland 

this 10th day of February 1997 

42 

Enclosure 1 
NOT IC E OF NONCONFORMANCE 

Ellis & Watts (E&W) 

Batavia, Ohio 

Docket No.: 99901308 

On the basis of an inspection by the staff of the U.S. Nuclear Regulatol*y 

Comnission (NRC) from December 2 through 6, 1996, it appears that the 

following activities were not conducted in accordance with NRC requirements: 

Criterion III nf Appendix B to 10 CFR Part 50, "Design Control," 

requires, in part, that measures provide for verifying or checking the 

adequacy of design. The verification or checking process shall be 

performed by individuals or groups Dih2r than those who performed the 

original design. 

E&W Quality Assurance Manual QC-4000N, Section 3, "Design Control ,'I 

Revision 7, paragraphs 3.1, 3.1.4, and 3.1.6, require design 

requirements, as specified in the procurement documents, to be correctly 

translated into specifications, drawings, procedures, and instructions. 

Design control measures and product design shall be verified by E&W for 

adequacy of the design, and the verification or checking process shall 

be performed by individuals or groups other than those who performed the 

original design. 

A. Contrary to the above requirements, E&W did not verify the 

adequacy of design pressure ratings of four air conditioner units 

(ACUs) supplied by E&W to the Tennessee Valley Authority (TVA) for 

the Browns Ferry Nuclear Plant. lhe shell side pressure rating of 

the installed ACU condensers was less than the minimum pressure 

necessary. As a result, coolant leaked from the ACU condensers. 

(99901308/96-01-02}. 

6. Contrary Lo the above requirements, E&W's commercial-grade 

dedication plans did not address the verification of the following 

critical characteristics (equipment performance features 

identified in 1 icensee procurement specifications and E&W 

dedication documents): (i) that cooling coil fans would operate 

during tornado conditions at a reduced pressure, (ii) that an oil 

pump would develop a required minimum oil pressure, (iii) that a 

thermostatic expansion valve would provide the required flow of 

liquid refrigerant to the evaporator coil at the specified 

accident temperature, and (iv) that a gasket for an oil pump and 

bearing assembly would perform its safety function during the 

specified accident temperature (99901308/96-01-03). 

Enclosure 2 

43 
2 Criterion XVI of Appendix B to 10 CFR Part 50, "Corrective Action," 

requires, in part, measures to assure that the cause of the condition is 

determined and corrective action taken to preclude recurrence. 

E&W Quality Assurance Manual QC-4000N, Section 16, "Corrective Action," 

Revision 7, paragraphs 16.1 and 16.5 require, in part, assurance that 

conditions adverse to quality are promptly identified and corrected and 

that corrective actions taken are effective to preclude recurrence. 

Contrary to the above requirements, E&W failed to establish effective 

corrective action for deficiencies in its commercial-grade dedication 

activities, including deficiencies re1 ated to critical characteristics 

and certified material test reports of commercial-grade materials 

(99901308/96-01-04) . 

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

Commission, ATTN: Document Control Desk, Washington, D.C. 20555-0001, with a 

copy to the Chief, Special Inspection Branch, Division of Inspection and 

Support Programs, Office of Nuclear Reactor 9egula:ion, within 30 days of the 

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

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

contain for the nonconformances (1) a description of steps that have been or 

will be taken to correct these items, (2) a description of steps that have 

been or will be taken to prevent recurrence of these items, and (3) the dates 

your corrective actions and preventive measures were or will be completed. 

Dated at Rockvi 11 e, Mary1 and 

this 10th day of February 1997 

2 
Report No: 

3rgani zat i on: 

Contact: 

Nuclear Industry 

Act ivi ty: 

Dates: 

Inspector: 

Approved by: 

U. S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

9990 1308/96-0 1 

Ellis & Watts 

Division of Dynamics Corporation-'of America 

Craig Hunt, QA Manager 

5:3!752-9000 

Heating, ventilation, and air conditioning (HVAC) 

systems and associated spare and replacement parts 

December 2-6, 1996 

Ani1 S. Gautam, Senior Engineer 

Gregory C. Cwalina, Chief 

Vendor Inspection Section 

Special Inspection Branch 

Division of Inspection and Support Programs 

Enclosure 3 

45 
1 INSPECTIOW SUMMARY 

During this inspection, the NRC inspector reviewed activities associated with 

implementatiep of selected portions of Ellis & Watts' (E&W's) quality 

assurance (QA) program and licensee monitoring of E&W's control of quality. 

The inspection bases were as follows: 

e 

e 

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

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

Reaulations (10 CFR Part 50). 

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

NRC Regulatory Guidc 1.144, "Auditing of Qual i ty Assurance Programs for 

Nuclear Power P1 ants. 'I 

E&W's Quality Assurance Manual QC-4000N, Revision 7, dated September 7, 

1995, and associated implementing procedures. 

During this inspection, a violation of NRC requirements was identified. The 

violation is discussed in Section 3.1 of this report. 

inspector noted three instances in which E&W failed to conform to NRC 

requirements imposed upon it by NRC licensees. This nonconformance is 

discussed in Section 3.1 of this report. 

In addition, the 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

This was the first NRC inspection of E&W. 

3 INSPECTION FINDINGS AND OTHER COMRENTS 

3.1 Qual i tv Assurance Prosram 

a. InsDection Scope 

The inspector examined E&W's QA program, policy, implementing procediires, 

conformance to procurement documents, corrective actions in 

response to licensee audit findings, commercial-grade item dedication, 

Part 21 evaluations, and monitoring of subvendors. 

b. Observations and Findinss 

E&W's QA staff comprised the QA manager, two quality engineers, three QA 

shop inspectors, and six test technicians. The QA manager reported 

directly to the president of E&W. 

The inspector observed that E&W' s procedure QA-112, "Compl i ance to 

Regulation 10 CFR Part 21 Procedure," Revision 3, failed to address the 

identification or evaluation of "deviations" from procurement document 

requirements. 

identify deviations from safety-related purchase order requirements, and 

As a result, personnel were not alerted to tk need to 

2 
to evaluate the deviations to determine if they could become defects. 

Part 21 requires adopting appropriate procedures to evaluate deviations 

to identify defects and failures to comply associated with substantial 

safety hazards. Failure to adopt appropriate procedures, as required by 

10 CFR Part 21.21, constitutes Violation 99901308/96-01-01. Based on 

E&W's December 30, 1996, letter to the NRC, the QA manager revised 

proredure QA-112 to address eva1:ration of deviations and require 

pertinent training for appropriate employees. No further response is 

requ i red. 

The inspector observed that E&W's Nonconformance/Deviat ion Control 

Procedure Form QC-12, "Request for Material Disposition," for documenting 

"major nonconformances, 'I did not address screening of issues for Part 21 

applicability. Based on E&W's December 30, 1996, letter to the NRC, the 

QA manager revised Form 2C-12 to address Part 21 applicability. 

further response is required. 

The inspector observed that QA-112, Revision 3, page 5, stated, in part, 

that Part 21 regulations were on file for employee "off-shift review in 

the QA department during work breaks, and after working nours." The 

inspector was concerned that the procedure would discourage or prevent 

employees from adequately reviewing or implementing Part 21. Based on 

E&W's December 30, 1996, letter to the NRC, the QA manager revised QA-112 

to remove restrictive instructions so that employees would review Part 21 

regulations and procedures during work hours. 

required. 

No 

No further response is 

E&W issued Part 21 report No. 101, dated August 20, 1994, io address 

deficiencies in the design pressure ratings of four air conditioner units 

(ACUs) supplied by E&W to the Tennessee Valley Authority (TVA) for the 

Browns Ferry Nuclear Plant. The shell side pressure rating of the 

installed ACU condensers was less than the minimum pressure necessary. The deficietxies resulted in coolant leaking from the ACU condensers. In 

February 1995, TVA audited E&W's QA program and identified deficiencies 

in the design and configuration control of the ACUs. On the basis of 

E&W's April 28, 1995, letter in response to TVA's findings, the inspector 

observed that E&W believed the defects in the ACUs to be a result of 

errors in TVA's design documents. QC-4000N, Section 3, "Design Control ,'I 

Revision 7, paragraph 3.1.6, requires product design to be verified by 

E&W for adequacy of the design and the verification or checking process 

to be performed by individuals or groups other than those who performed 

the original design. 

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

50, constitutes Nonconformance 99901308/96-01-02. 

E&W's failure to verify the adequacy of design, as 

47 

The inspector assessed E&W's design control and commercial-grade 

dedication activities by evaluating purchase orders (POs) from Texas 

Uti1 ities Electric Company (TUEC) , TVA, and Duquesne Light Company (DLC) 

to E&W for the following items: 

3 
ltem Licensee Licensee PO No. P.O. Date 

Coil fan assembly TUEC S013040056Dl July 25, 1994 

Oil pump TVA P-94N2S-105874-000 March 25, 1994 

Therrna 1 expans i on Val ve TVA P-94N2S-115662-000 September 19, 1994 

Gasket TVA P-94N2S-105639-000 March 18, 1994 

Pressure switch DLC 0134982 November 3, 1994 

The inspector assessed E&Wy s commercial-grade dedication documents for 

the above items, including review and analysis of intended safety 

functions, identification and verification of critical characteristics 

(equipment performance features identified in 1 icensee procurement 

specifications and E&W’s dedication documents), tests conducted to verify 

characteristics, and any failures during testing. The inspector observed 

that dedication plans did not address verification of the following 

critical characteristics: 

e that cooling coil fans would operate during tornado conditions at a 

reduced pressure of 11.7 psia at the Comanche Peak Steam Electric 

St at i on. 

e that an oil pump would develop a minimum oil pressure of 11 psi at 

the Browns Ferry Nuclear Plant. 

e that a thermostatic expansion valve would provide the required flow 

of liquid refrigerant to the evaporator coil at the specified 

accident temperature at the Browns Ferry Nuclear Plant. 

0 that a gasket for oil pump and bearing assembly would perform its 

safety function without degradation during the specified accident 

temperature at the Browns Ferry Nuclear Plant. 

E&W stated that the items were qualified based on certificates of 

conformance (COCs) from subvendors but could not provide test 

documentation for the above CharacterisLics. The inspector contacted 

TUEC and TVA to determine whether the above items were qualified to 

perform their intended safety functions. 

analysis, the licensees reported no operability concerns associated with 

the installed items. EW’s failure to implement measures to assure 

review for suitability of application of materials, parts, equipment, and 

processes that are essential to the safety-related functions of systems 

and components, as required by Criterion 111, “Design Control,” of 

Appendix B to 10 CFR Part 50, constitutes Nonconformance 99901308/96-01-03. 

On the basis of a preliminary 

The inspector observed that there were continuing deficiencies in E&W’s 

dedication program. The Nuclear Utilities Procurement Issues Committee 

(NUPIC) audited E&W in August 1995, to determine, in part, whether E&W 

had established and effectively implemented a QA program in compliance 

with 10 CFR 50 Appendix B. 

design control, configuration control, and commercial-grade dedication. 

Deficiencies included E&W’s failure to reference materials as a critical 

attribute during dedication, and to verify the validity of certified 

materi a1 test reports (CMTRs) and COCs suppl ied by E&W’s commerci a1 -grade 

NUPIC identified deficiencies in the areas of 

4 

48 
subvendors. In April 1996, NUPIC wdited E&W's corrective actions and 

again identified deficiencies in E&A' i dedication program, including E&W 

not identifying materials as a critical attribute and not verifying the 

validity of CMTRs. E&W failed to establish effective corrective action 

for deficiencies in its commercial-grade dedication activities that were 

identified by NUPIC. 

regarding E&W's dedication activities did not address verification of 

certain critical characteristics. T9e inspector conclude4 that E&W's 

continuing failure to assure that conditions adverse to quality are 

promptly corrected and that corrective actions are taken to preclude 

repetition, as required by Criterion XVI, "Corrective Action," of 

Appendix B to 10 CFR Part 50, constitutes Nonconformance 99901308/96-01- 

04. 

Similarly, documents reviewed by the inspector 

E&W implemented measures to identify nonconforming items by applying 

appropriate tags. However, the inspector observed a lack of physical 

segregation betwew safety-related items and non-safety-related items in 

storage and staging areas. Criterion XV, "Nonconforming materials, prts 

or components," of Appendix B to 10 CFR Part 50, requires segregation of 

nonconforming materials, parts or components to prevent their inadvertent 

use or inntallation. QC-4000N, Sectim 15, "Nonconforming Materials, 

Parts or Components," Revision 7, paragraph 15.2.2, states, in part, that 

storage areas for nonconforming items shall be segregated. Based on 

E&W's December 30, 1996, letter to the NRC, E&W fabricated and installed 

a bonded locked cage to segregate nuclear non-conforming materials during 

"receiving inspection." E&W also reported that it was in the process of 

establishing a locked segregated and controlled area for storage of 

nuclear material. No further response is required. 

The inspector observed that TUEC's PO S013040056Dl to E&W, dated July 7, 

1994, included an agreement between E&W and TUEC to inform the licensee 

immediately of any allegations of discrimination (Section 211 of the 

Energy Reorganization Act of 1974) filed by an E&W employee or former 

employee with a Federal, State, or local administrative agency, in 

connection with E&W's activity related to the Comanche Peak Steam 

Electric Station. 

agreements in its POs to increase TUEC's efforts in responding to 

allegations of discrimination filed by vendor employees. These efforts 

included monitoring the adequacy of the vendor's investigation znd 

actions to achieve d remedy (as indicated by the Commission in its 

Statement of Policy [61 FR 243361, "Freedom of Employees in the Nuclear 

Industry to Raise Safety Concerns Without Fear of Retaliation," dated May 

14, 1996). At the exit meeting, the inspector expressed concern that the 

agreement, as delineated in the PO, could inadvertently discourage an E&W 

employee from engaging in protected activity (as described in 10 CFR Part 

50.7 (a)(l)(i)), for fear of retaliation from TUEC. However, after 

further review of the above policy statement, the NRC determined that the 

PO agreement was appropriate, and was consistent with the Commission 

Pol icy Statement referenced above. 

the vendor that the PO agreement was appropriate, and was not in 

violation of NRC requirements. 

inform employees that the PO agreement was meant to protect, rather than 

intimidate, employees who raised safety concerns. 

were i dent i f i ed . 

The inspector determined that TUEC included\such 

The inspector subsequently informed 

E&W stated that it would take measures to 

No further concerns 

5 

49 
c . Concl u-sjpns 

The QA manual and procedures were adequate, except for the violation and 

nonconformances described herein. 

reviewed indicated a weakness in the areas of design control and 

comerc i a1 -grade dedi cat i on. 

Problems with dedication plans 

3.2 Review of-Moritorino of E&W bv Licecsees 

a. InsDection ScoDe 

The inspector evaluated 1 icensee monitoring of E&W’s control of quality 

for safety-related items purchased by 1 icensees, including audits and 

surveillances of E&W’s commercial-grade dedication, Part 21 reports, and 

monitoring of subvendors. 

b. Observations a&d Findinas 

The inspector contacted TVA, TUEC, DLC, and Carolina,Power & Light (CP&L) 

to discuss the scope and findings of their audits and surveillance of 

E&W. 

TVA audited E&W in February 1995, to evaluate, in part, the defects TVA 

identified in the ACUs suppl ied by E&W for the Browns Ferry Nuclear 

Station, Units 1 and 3. TVA determined that E&W had deficiencies in the 

areas of design control and configuration control. TVA made two 

findings: a lack of formalized design analysis review affecting equipment 

installed in Units 1 and 3, and inadequate control of changes to design 

documents. In February 1996, TVA accepted E&W’s corrective actions but 

imposed procurement restrictions, one of which was that TVA review E&W’s 

design for any new purchases. 

NUPIC audited E&W in August 1995. 

audit team included CP&L (team leader), TVA, and Niagara Mohawk Power 

Corporaiion. 

had established and effectively implemented a QA program in compliance 

with the requirements of 10 CFR Part 50 Appendix B and other industry 

standards. 

ineffective” a,.d mde eight findings concerning E&W’s QA program: (l) weak design control measures not assuring documented independent 

evaluations of materials and engineering changes, (2) computer software 

not controlled in accordance with E&W procedures, (3) various 

deficiencies in the monitoring of subvendors, (4) unqualified welder, (5) 

inadequate 1994 and 1995 internal audits and related corrective actions, 

(6) inadequate measures for traceability of weld filler material, (7) 

inadequate storage facilities for quality documents, and (8) lack of 

implementation of procurement requirements of one PO. NUPIC recomnended 

that uti1 ities explicitly communicate to E&W their expectations in the 

area of commercial-grade dedication through their purchase orders, and 

require E&W to provide objective evidence of dedication for review by the 

utility at receipt. 

In March 1996, NUPIC (represented by CP&L) examined E&W’s corrective 

actions in response to NUPIC’s August 1995 findings. NUPIC found 

continuing deficiencies in E&W’s dedication documents for comercial6 

50 

The utilities represented on the NUPIC 

The scope of the audit, it: part, was to verify whether E&W 

NUPIC concluded that E&W’s QA program was “weak and partially 
grade ibems. For example, certain aedication plat13 did ndL identify 

materials as critical attributes and did not verify the validity of the 

certified material test reports. NUPIC concluded that E&W had not 

developed a comprehensive program to control purchased materi a1 and 

equipment, and ensure that all critical attributes were adequately 

verified during comnerci a1 grade surveys. NUPIC recommended that 

licensees perform detailed verificatim of E&W’s design reviews and 

surveillance before and during perforrnancc of work for future orders. 

CP&L retained E&W on its list of approved suppliers with these 

procurement restrictions: that CP&L verify the design of any new 

purchases, that CP&L approve E&W’s dedication plan for commercial-grade 

spare parts prior to E&W’s performance of work, and that E&W verify the 

validity of COCs or CMTRs for material purchased through subvendors who 

did not have provisions for material verification. 

TUEC and DLC’s monitoring of E&W was based on the results of the NUPIC 

audits. 

NUPIC’s audit applied to DLC’s purchases. No concerns were identified. 

DLC provided the inspector a copy of their evaluation of how 

c. Conclusions 

In general, licensees effectively audited E&W in accordance with proper 

criteria, procedures, and check1 ists. Licensees continue to have 

concerns regarding E&W’s design control and commercial-grade dedication 

activities despite E&W’s corrective actions. 

3.3 Entrance and Exit Meetinas 

In the entrance meeting on December 2, 1996, the NRC inspector discussed 

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

established interactions with E&W management. 

December 6, 1996, the inspector disciissed his findings and observations. 

In the exit meeting on 

7 
4 PERSONNEL CONTACTED 

Craig Hunt, QA Manager 

Roger Schertler, QA Engineer 

Jim York, Manager, Special Projects 

Richard Porco, Manager, Engineering 

Licensees (co ntacted by teleDhone1 

Danny Leigh, Plant Overview Supervisor, TUEC 

John Taylor, Procurement Engineer, TUEC 

Hrach Minassi an , Procurement Engineer , TUEC 

William Sidberry, Senior Anallst, CP&L 

R.G. Newsome, NUPIC Team Leader, CP&L 

James Johns, Supervi sor Qual i ty Services , DLC 

Larry Spiers, Procurement Engineer, TVA 

ITEMS OPENED, CLOSED, AND DISCUSSED 

ODened 

99901308/96-0 1-01 NOV inadequate Part 21 procedure 

99901308/96-01-02 NON inadequate design control 

99901308/96-01-03 NON inadequate design control 

99901308/96-01-04 NON inadequate corrective action 

Closed 

99901308/96-01-01 NOV inadequate Part 21 procedure 

8 

52 
- ****4 

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTOI., C ' ;35650001 

March 11, 1997 

Dr. Ivan E. Wilkinson, P. E. 

Director, Ergi neeri ng 

Limitorquz Corporation 

5114 Woodall Road 

P.0.Box 11318 

Lynchburg, VA 24506 

Dear Dr. Wilkinson: 

SUBJECT : NRC INSPECTION REPORT 99900100/97-01 

On February 26, 1997, the 'J.S. Nuclear Regulatory Comnission (NRC) c::pleted 

an inspection at your Lynchburg facility. 

results of that inspection. 

The enclosed report presents the 

During this inspection, the NRC team reviewed the root cause analysis you 

performed on a motor-operated-valve actuator from Oyster Creek Nuclear 

Generating Plant to determine why it had failed to electrically operate in 

1996, and reviewed the corrective actions taken on a violation identified in 

Inspect ion Report 99900100/93-01 and an unresolved i tem ident i f ied in 

Inspection Report 99900100/91-01. 

The inspectors did not identify any instances where your quality assurance 

program failed to meet NRC requirements for the areas inspected. Therefore, 

no response to this letter is required. 

viol at ion 99900100/93-01-01 and unresolved i tem 9900100/91-01-07 were 

determined to be adequate and are closed. 

The corrective actions taken on 

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

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

Room (PDR) . 

Si ncerelfl- - 

Robert M. Gallo 

Speci a1 Inspection Branch 

Division of Inspection and Support Programs 

Office of Nucl ear Reactor Regul at i on 

Docket No.: 99900100/97-01 

Enclosure: Inspection Report 99900100/97-01 
Reoort No. : 

Organization: 

Contact : 

Nuclear Industry: 

Activity 

Date: 

Inspectors: 

Approved by: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

99900 100/97-01 

Limi torque Corporation 

5114 Woodall Road 

Lynchburg, Virginia 24506-1318 

q. Segen, Director, Qual i ty Assurance 

Limitorque Corporation 

5114 Woodall Road 

Lync hburg , Vi rg i n i a 24506- 13 18 

(804) 528-4400 

Manufactures, services and provides rep1 acement 

components and sub-assemblies for its motor-operated 

valve actuators. 

February 26, 1997 

Kamalakar R. Naidu, Senior Reactor Engineer 

Joseph 3. Petrosino, Q.A. Specialist 

G. C. Cwalina, Chief 

Vendor Inspection Section (VIS) 

Speci a1 Inspect ion Branch 

uivision of Inspection and Support Programs 

Office of Nuclear Reactor Regul at ion 
1. INSPECT ION SUMMARY 

The team reviewed the root use analysi th t Limi torque Corporation 

(Limi torque) performed on a SMB Size '1 motor-operated-valve (MOV) actuator 

from General Public Uti1 ities (GPU) Nuclear Corporation to determine the cause 

of failure to electrically operate. 

nonsafety-rvlatecl application at Oyster Creek Nuclear Generating Station 

(Oyster Creek). 

The actuator had been installed in a 

Inspection bases were: 

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

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

Federal Reaulations (10 CFR Part 50, Appendix 8) 

e 10 CFR rL. t 21, "Reporting Defects and Noncompliance" 

2. STATUS OF PREVIOUS INSPECTION FINOINGS 

One violation identified in Inspection Report 99900100/93-01 and one 

unresolved item identified in Inspection Report 99900100/91-01-07 were 

reviewed during this inspection. The inspectors did not review information on 

unresolved items (9990100/91-01-03, 04, and 05) during this inspection. 

Violation 93-01-01 (Closed). A May 10 through 14, 1993, inspection identified 

that contrary to 10 CFR Part 21.21, Limitorque failed to complete its 

evaluation and did not report a condition associated with the relaxation of 

motor actuator spring packs. 

Limitorque revised Quality Assurance Procedure (QAP) 13.2, "Reporting of 

Defects for Safety Related Equipment," on May 14, 1993, to require the use of 

Form L-345 to document future evaluations. 

Limitorque Log No. 08 dated June 8, 1993, documented that the effect of spring 

evaluation was insignificant and so advised the industry with its Technical 

Update 93.02 of the same date. 

Unresolved Item 994~013-91-01-07 (Closedl. It was identified that Lint; torque 

had committed to, but had not notified NRC or informed its customers by way of 

Maintenance Update as of May 14, 1993, of a possible defect concerning 

improper machining of actuator limit stop housings for HBC-1 actuators. 

Limi torque issued "Limi torque Maintenance Update 93-1 , " on August 31 , 1993 , to 

address this subject and inform its customers of the deviation. 

3. INSPECTION FINDINGS AND OTHER COMMENTS 

3.1 Introduction 

In a letter dated January 23, 1997, Limitorque Corporation, pursuant to the 

reporting requirements of 10 CFR Part 21, informed the NRC that it had 

discovered the existence of a "counterfeit" component installed in a 

Limi torque actuator at a nuclear power plant. 

discrepant worm shaft clutch gear in a SMB Size 1 actuator which failed to 

operate in a nonsafety-related application at GPU's Oyster Creek facility. 

Limi torque discovered the 

2 

55 
During this inspection, the NRC inspectors reviewed Limitorque's root cause 

analysis, inspected the worm shaft clutch gear and collected additional 

information to evaluate further action. 

According to Oyster Creek's Deviation Report (DVR 96-870), the failed worm 

shaft clutch gear in question was supplied by Sigma, Inc. The gear was 

manufactured by C. I. Supply Company. 

tc, replace a failed gear in a nonsafety-related actuator. Oyster Creek states 

that only payts and operators provided directly from Limitorque are used in 

safety-re1 ated appl i cat i ons however several nonsafety-re1 ated actuators 

could contain parts that were not provided by the Original Equipment 

Manufacturer (OEM). 

The gear was purchased by Oyster Creek 

DVR 96-870 indicates that Oyster Creek experienced two failures of nonsafetyrelated Limitorque operators due to worm shaft clutch gear problems. 

operators had been procurd from Power Equipment Supply Company. 

failure resulted in a replacement gear being installed with an original 

Limitorque gear from the Oyster Creek warehouse. 

Oyster Creek to purchase the gear in questiQr from Sigma. 

The 

T;,e first 

The second failure caused 

3.2 Limitoraue Root Cause Analvsis 

a. InsDection ScoDe 

Evaluate the root cause analysis performed by Limitorque on the worm shaft 

clutch gear that failed to operate at Oyster Creek. 

for other similar nonconforming parts to be installed in safety-related 

applications in the nuclear industry. 

b. Observations and Findinqs 

Determine the potential 

The inspectors reviewed the root cause analysis performed by Limitorque on tlle 

worm shaft clutch gear that failed at Oyster Creek (see figures 1 and 2). 

Limitorque subjected the gear to various examinations and determined that: 

The motor clutch gear cam pin staking operation had been omitted. 

The worm shaft clutch gear lugs were not heat treated. 

The gear teeth were inappropriately shaped and not shaved. 

There were a variety of other non-serious disparities in the geometry of 

the part. 

Based on the above results, Limitorque determined that it had not manufactured 

the gear and that it was a "counterfeit." 

During a plant tour, Limitorque personnel demonstrated the machinery used to 

hob and shave gear teeth and to precisely measure tire resulting tooth profile. 

Limitorque personnel informed the inspectors that they do not use any 

subcontractors to manufacture the worm shaft clutch gear and stated that they 

had no information on other potentially "counterfeit" or nonconforming parts. 

3 

56 
The inspectors observed that Limi torque uti1 ized Procedure QAP 13.2, 

"Reporting of Defects for Safety Related Equipment," to document its 

evaluation on Form L-345; Limitorque concluded that the matter was not 

reportable to the NRC pursuant to Part 21. 

matter was not reportable, Limitorque concluded that the worm shaft clutch 

gear assembly was inferior in design, was not suitable for the intended 

application, and that the non-OEM part would have a short service life 

.terminating in failure with associated consequences depending upon the 

criticality of the application. Since the suspect worm shaft clutch gear can 

be used in Limitorque's SMB, SB, SBD, or HBC type actuators installed in 

safety-re1 ated appl i cat i ons, a substant i a1 safety hazard could be created. 

Therefore, in a letter dated January 27, 1997, Limitorque informed the NRC. 

Even though it considered the 

c. Conclusion 

The root cause eval uat i on Was we1 1 documented, compl ete and thorough. 

Although not required, in a letter dated January 27, 1997, Limitorque informed 

all nuclear power plants of the discovery of a potential condition concerning 

the suspect qear so that the plants could evaluate the issue for applicability 

to their plants. 

4. PERSONS CONTACTED 

I. E. Wilkinson, Director, Engineering 

*+R.D. Segen, Director, Quality Assurance 

*+P.G. McQuillan, Manager, Special Projects OtWrn. 3. Miluszusky, Quality Assurance 

Denotes attendance at the entrance meeting on February 26, 1997. 

t Denotes attendance at the exit meeting on February 26, 1997. 
Oyster Creek Ncn-OEM .Sear 

Figure 1 - Worm Shaft Clutch Gear (Plan View) 

. . . . . . . . ._ - . . ... .. . . . 

5 

58 
Ident i f i er 

Informat 

Informat 

Selected Generic Correspondence on the Adequacy of 

Vendor Audits and the Quality of Vendor Products 

on Not 

on Not 

Title 

ce 97-8 Potential failures of General Electric MagneBlast Circuit Breaker Subcomponents 

ce 97-12 Potential Armature Binding In General Electric 

Type HGA Relays 

59 

U.S. NUCLEAR REGULATORY COMMISSION IRC FORM 335 

!-El91 

RCM 1102, 

201,3202 BIBLIOGRAPHIC DATA SHEET 

(see instructions on the reversel 

.TITLE AND SUBTITLE 

Licensee Contractor and Vendor Inspecti on 

Status Report 

Quarterl y Report 

January - March 1997 

. AUTHOR(S) 

1. REPORT NUMBER 

(Assignad by NRC. Add Vol., Supp., Rev., and Addendum Numbers. if any.) 

NUREG-0040 

Volume 21, No. 1 

3. DATE REPORTPUBLISHED 

MONTH YEAR 

July ' 1997 

4. FIN OR GRANT NUMBER 

6. TYPE OF REPORT 

Quarterl y 

7. PERIOD COVERED llnclusive Dates1 

January-March 1997 

. PERFORMING ORGANIZATION - NAME AND ADDRESS lIfMRC,provi& Division. OfficeorRegion, U.S. Nuclear Regulatory Commission, andmailingaddress;ifcontractor,provide 

name and mailing addred 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

US. Nuclear Regulatory Commission 

Washi ngton , D. C. 20555-0001 . SPONSORING ORGANIZATION - NAME AND ADDRESS /If MRC, type 'Same as above'>if conrracmr, provide NRC Division, Office or Region, U.S. Nuclear Rwulatow Commission, 

and mailing address) 

Same as above 

0. SUPPLEMENTARY NOTES 

1. ABSTRACT (200 words or led 

This periodical covers the results of inspections performed by the NRC's Special 

Inspection Branch, Vendor Inspection Section, that have been distributed to the 

inspected organizations during the period from January through March 1997. 

2. KEY WORDSjDESCRIPTORS (List words orphrases rhar willassistresearchenin locatingrhereport.l 13. AVAlLABl LlTY STATEMENT 

Unl imi ted 

Vendor Inspection 14. SECURITY CLASSIFICATION 

(This Page1 

Uncl assi f i ed 

Uncl assi f i ed 

(This Report1 

15. NUMBER OF PAGES 

16. PRICE 

RC FORM 335 (2-891 


on recycled 

Federal Recycling Program