NUREG-0040 

Vol. 19, No. 1 

Licensee Contractor 

and Vendor Inspection 

Status Report 

Quarterly Report 

January - March 1995 

U.S. Nuclear Regulatory Commission 

OfEce of Nuclear Reactor Regulation 
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NUREG-0040 

Vol. 19, No. 1 

Licensee Contractor 

and Vendor Inspection 

Status Report 

Quarterly Report 

January - March 1995 

Manuscript Completed: April 1995 

Date Published May 1995 

Division of Technical Support 

Office of Nuclear Reactor Regulation 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

DISTRIBUTION OF THIS DOCUMENT IS UNLIMITED 

ABSTRACT 

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

Special Inspection Branch, Vendor Inspection Section, that have been 

distributed to the inspected organizations during the period from January 1995 

through March 1995. 

iii 

CONTENTS 

PAGE 

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

Introduction ............................................................. vi i 

Inspection Reports ....................................................... 

Gi rard Development , Inc. 

Rancho Cordova, Cal i forni a 

Mackson Inc. 

Rock Hill, South Carolina 

(99901283/95-01) ............... 2 

(99901079/95-01) ............... 14 

Sandvik Special Metals Corporation (99900764/94-01) ............... 30 

Kennewick, Washington 

Tel edyne Wah Chang 

A1 bany, Oregon 

(99901229/94-01) ............... 46 

West i nghouse Electric Corporation (99900404/94-01) ............... 62 

Pittsburgh, Pennsylvania 

Wyl e Laboratories 

Huntsvi 1 1 e, A1 abama 

(99900902/95-02) ............... 94 

Select Generic Correspondence on the Adequacy of Vendor .................. 104 

Audits and the Quality of Vendor Products 

Correspondence Re1 ated To Vendor Issues .................................. 105 

V 

INTRODUCTION 

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

licensing and inspection program is that licensees are responsible for the 

proper construction and safe and efficient operation of their nuclear power 

plants. 

for the inspection of commercial nuclear facilities to provide for multiple 

levels of inspection and verification. Each 1 icensee, contractor, and vendor 

part i ci pates i n a qual i ty veri fi cation process in compl i ance wi t h requi rement s 

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

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

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

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

industry within the framework of quality verification programs. 

The Federal government and nuclear industry have established a system 

The licensee is responsible for developing and maintaining a detailed quality 

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

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

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

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

codes, and standards. 

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

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

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

laboratories performing equipment qualification tests, and holders of NRC 

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

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

problems are determined and appropri ate corrective acti ons are devel oped. The 

inspections also review vendors to verify conformance with applicable NRC and 

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

coordination between 1 icensees and vendors. 

These 

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

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

and review of equipment problems found during operation and their corrective 

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

the inspected organization is required to take appropriate corrective action 

and to institute preventive measures to preclude recurrence. 

implications are found, NRC ensures that affected 1 icensees are informed 

through vendor reporting or by NRC generic correspondence such as information 

notices and bull et i ns . 

When generic 

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

during the calendar quarter for which it is published. 

report lists the nuclear facilities inspected. This information will also 

alert affected regional offices to any significant problem areas that may 

require special attention. Appendices list selected bulletins, generic 

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

correspondence invol ving vendor issues . 

Each vendor inspection 

viii 
INSPECTION REPORTS 

1 
+**it* '- 

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 205SSOOO1 

March 1, 1995 

Mr. Don Girard, President 

Gi rard Devel opment , Inc. 

P.O. Box 338 

Rancho Cordova, CA 95741 

SUBJECT: NRC INSPECTION NO. 99901283/95-01 

Dear Mr. Girard: 

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

of your facility at Rancho Cordova, California, conducted by Messrs. 

U. Potapovs and B.H. Rogers of this office on January 24 and 25, 1995, and the 

discussions of their findings with you at the conclusion of the inspection. 

The inspection was Conducted to evaluate your quality assurance program and 

its implementation in selected areas such as (1) control of purchased material 

and services, (2) material and traceability control, (3) supplier audits, 

(4) design control, and (5) a review of your program for implementing Part 21, 

"Reporting Defects and Noncompliance," of Title 10 of the Code of Federal 

Requl at ions (10 CFR) . 

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

enclosed inspection report. 

procedures and representative records, discussion, and observations by the 

inspectors. 

This inspection consisted of an examination of 

A1 though your qual i ty assurance program imp1 ementation was generally 

satisfactory, the inspection identified that it did not meet applicable NRC 

requirements in the areas of corrective action, qualification of suppliers of 

safety re1 ated servi ces, and certi f i cation of materi a1 . and references to the pertinent requirements are identified in the enclosures 

to this letter. 

The speci f i c findings 

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

statement in accordance with the instructions specified in the enclosed Notice 

of Nonconformance. 

show good cause for us to do so. 

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

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

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

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

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

We will consider extending the response time if you can 
D. Girard -2- 

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

discuss them with you. 

Sincerely, 

Speci a1 Inspection Branch 

Division of Technical Support 

Office of Nuclear Reactor Regulation 

Docket No. 99901283 

Encl osures : 

1. Notice of Nonconformance 

2. Inspection Report 99901283/95-01 

3 
NOTICE OF NONCONFORMANCE 

Gi rard Development , Inc. 

Rancho Cordova, Cal i forni a 

Docket No. : 99901283/95-01 

Based on the results of an NRC inspection conducted on January 24 and 25, 

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

accordance with NRC requirements. 

A. Criterion XVI, "Corrective Action," of Appendix B to Title 10 of the 

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

measures shall be established to assure that conditions adverse to 

quality such as failures, malfunctions, deficiencies, deviations, 

defective material and equipment, and nonconformances are promptly 

i dent i f i ed and corrected. 

Contrary to the above, Girard Development, Inc. (GDI) did not establish 

measures to assure that conditions adverse to quality were promptly 

corrected. Specifically, a 1989 customer audit finding concerning the 

lack of design review for all Girard tube clamps was not corrected until 

1993, when another external audit identified the same deficiency. 

(Nonconformance 99901283/95-01-01) 

B. Criterion VI1 "Control of Purchased Material , Equipment and Services," 

of Appendix B to Title 10 of the Code of Federal Requlations (10 CFR) 

Part 50, requires, in part, that measures shall be established to assure 

that purchased material , equipment, and services conform to the 

procurement documents and that the effectiveness of control of quality 

by contractors and subcontractors shall be assessed at interval s 

consistent with the importance, complexity, and quantity of the product 

or service. 

Contrary to the above, GDI did not establish measures to assure that 

purchased services conformed to the procurement documents and to assess 

the effectiveness of control of quality by subcontractors at intervals 

consistent with the importance, complexity, and quantity of the service. 

Specifically, GDI had contracted Pacific Consulting Engineers (PCE) to 

verify the adequacy of GDI's tube clamp design and had not taken action 

to assess PCE to assure the effectiveness of control of quality and to 

assure that the services conformed to procurement documents. 

(Nonconformance 99901283/95-01-02) 

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

Title 10 of the Code of Federal Requlations (10 CFR) Part 50, requires, 

in part, that activities affecting quality shall be prescribed by 

documented instructions or procedures appropriate to the circumstances 

and shall be accomplished in accordance with these instructions or 

procedures. 

4 

Enclosure 1 
GDI's Quality Assurance Manual commits to compliance with the procedures 

and instructions contained in Subarticle NCA 3800 of Section I11 of the 

ASME Boiler and Pressure Vessel Code (Code). 

Paragraph NCA 3820(a) of Section I11 of the Code states that a Material 

Manufacturer shall hold a Quality System Certificate (Materials) issued 

by the Society or have his Quality System Program surveyed, qualified, 

and audited by the Material Supplier holding a Quality System 

Certificate or by the Certificate Holder for the items. 

Contrary to the above, GDI did not accomplish activities affecting 

quality in accordance with established procedures. Specifically, GDI 

did not accomplish the procurement activities associated with the supply 

of tube clamps under Georgia Power Company's purchase order 60191890000, 

which required compliance with Section I11 of the ASME Code, in 

accordance with the instructions and procedures contained in Subarticle 

NCA 3800 of Section I11 of the ASME Code, as required by GDI's Quality 

AsstmTze Manual. GDI was neither a Quality Systeni Certificate holder 

nor htJ been qualified by a Certificate holder to supply material in 

accordance with NCA 3800, yet issued certification to Georgia Power 

Company stating that the tube clamps which they supplied met the 

requirements of Section I11 of the ASME Code. 

(Nonconformance 99901238/95-01-03) 

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

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

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

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

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

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

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

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

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

and preventive measures were or will be completed. 

Dated at Rockville, Maryland 

this 1st day of March, 1995 

5 
REPORT NO.: 

ORGANIZATION: 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY : 

INSPECTION DATES: 

LEAD INSPECTOR: 

OTHER INSPECTORS: 

REVIEWED BY: 

APPROVED BY: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF TECHNICAL SUPPORT 

99901283/95-01 

Girard Development, Inc. 

P.O. Box 338 

Rancho Cordova, California 95741 

Don Girard, President 

Girard Development, Inc. provides split block tubing 

cl amps 

January 24 and 25, 1995 

n 

Uldis Potapovs \ Date 

Vendor Inspection Section 

Special Inspection Branch 

Billy H. Rogers 

Vendor Inspection Section 

Special Inspection Branch 

Special Inspection Branch 

-Robert M. Gallo, Chief, Datb 

Special Inspection Branch 
1 SUHHARY OF INSPECTION FINDINGS 

During this inspection, the NRC inspectors evaluated Girard development, 

Inc. 's (GDI) implementation of qual i ty assurance measures for those 

activities which directly affected the quality and performance capability of 

their product. These activities included the control of purchased material 

and services, audits, material traceability, and design control. The team 

also reviewed GDI's program for implementing Part 21, "Reporting Defects and 

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

The inspection basis consisted of the following: 

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

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

Federal Requlations (10 CFR Part 50) 

10 CFR Part 21 

Section I11 of the ASME Boiler and Pressure Vessel Code (Code) 

Girard's Qual i ty Assurance Manual, Revision 22 

1.1 Violations 

No violations were identified during this inspection. 

1.2 Nonconformances 

1.2.1 Nonconformance 95-01-01 

This nonconformance, described in Section 3.4.7 of the report identifies an 

instance where GDI failed to establish measures for corrective action as 

required by Criterion XVI of Appendix B to 10 CFR 50 and did not take prompt 

corrective action when a condition adverse to quality was identified during a 

customer audit. 

1.2.2 Nonconformance 95-01-02 

This nonconformance, described in Section 3.4.2 of the report identifies an 

instance where GDI did not assess the effectiveness of the control of quality 

by a subcontractor as required by Criterion VI1 of Appendix B to 10 CFR 50. 

This subcontractor was hired to perform an independent design review of GDI's 

1 oad capacity cal cul at i ons . 

1.2.3 Nonconformance 95-01-03 

This nonconformance, described in Section 3.4.8 of this report, relates to 

GDI's failure to implement the commitments of their quality assurance manual 

in the procurement and certification of material supplied under the rules of 

the ASME Code. 

2 

I 7 
2 STATUS OF PREVIOUS INSPECTION FINDINGS 

This was the first NRC inspection of GDI 

3 INSPECTION FINDINGS AND OBSERVATIONS 

3.1 Entrance and Exit Meetincrs 

During the entrance meeting on January 24, 1995, the NRC inspectors discussed 

the inspection scope and developed general information about GDI’s products 

and activities. During the exit meeting on January 25, 1995, the NRC 

inspectors discussed their findings and observations with GDI management. 

3.2 Description of Facilities 

GDI supplies pipe and tube split block clamps of their own design in sizes 

ranging from 3/16 to 1 1/2 inches. These clamps are advertised as nuclear 

grade seismic tubing clamps, produced to ASME Code, Section I11 NF and NCA. 

The clamps are typically ordered and supplied to a catalogue part number which 

identifies a standard size and configuration. Although the clamps are 

designed to be used with standard Unistrut supports, neither the supports nor 

the bolts required to assemble the clamp are provided by GDI. 

capacity data sheets provided with these clamps are based on using them with 

ASTM A-354 Grade BD or equivalent bolts. 

The load 

The scope of safety related work activity at GDI is presently limited to 

accepting customer orders and filling them, including packaging and shipping, 

from existing stock as well as maintaining an inventory of clamps in the 

warehouse. 

supplied by the manufacturer. All other activities, including manufacturing, 

identification stamping, as well as any testing and dimensional inspections of 

the product are subcontracted. 

GDI also performs a limited receiving inspection on clamps 

GDI has an approved subcontractor, Southern Tool , Inc. , Anniston, Alabama 

(STI), which produces the clamps using GDI owned tooling and performs all 

inspection/verification activities. 

require no machining operations. 

The clamps are investment cast and 

3.3 10 CFR Part 21 PROGRAM 

The inspectors verified that GDI had posted a current copy of 10 CFR Part 21 

as required by 10 CFR Part 21.6. 

10 CFR Part 21 implementing procedure indicated a weakness in that it 

contained several inaccuracies and did not appear to be based on the current 

version of 10 CFR Part 21 which GDI had posted. The inspectors reviewed the 

current version of 10 CFR Part 21 with GDI and discussed the important 

concepts re1 ated to evaluation of deviations and the reporting of defects. 

However the inspector’s review of GDI’s 

CFR Part 21 implementing procedure would be updated 

GDI also stated that it had not encountered any 

a 10 CFR Part 21 evaluation. 

GDI indicated that the 10 

based on this discussion. 

deviations which required 

3 

8 
3.4 Qual i tv Assurance Proqram Imp1 ementati on 

GDI's Quality Assurance Program is described in a manual which committed to 

meet the applicable requirements of Appendix B to 10 CFR 50 and Subarticle NCA 

3800 of Section I11 of the ASME Code. The manual was issued in 1980 and has 

been periodically revised with the current revision being Revision 22 dated 

March 15, 1992. Selected areas of the manual and its implementation were 

examined during this inspection as indicated below. 

3.4.1 Design Control 

The inspectors identified weaknesses in GDI's quality assurance manual in 

this area. The manual states that design control is achieved through 

compliance with ASME Section I11 requirements and the generation of a load 

capacity data sheet based on allowable stresses for an approved material after 

the product is designed. The manual also states that it is GDI's intent not 

to change the design, once it is established. The manual does not address 

design verifjration as required by Criterion I11 of Appendix B to 10 CFR 50. 

After this issue was identified during several customer audits, GDI contracted 

with an engineering company in 1993 to perform an independent review of their 

load capacity data sheets. However, the quality assurance manual was not 

revised to include this requirement. 

The inspectors also noted that Section I11 of the ASME Code contains 

requirements for design review which are not addressed in GDI's quality 

assurance manual. Specifically, Paragraph NCA 3551.2 requires load capacity 

data sheets to be certified by a Registered Professional Engineer who is 

qualified in accordance with ANSI/ASME N626.3. 

the product can not be considered as complying with the requirements of 

Section I11 of the Code. 

Without such certification, 

The inspectors also questioned GDI's classification of the product as "Seismic 

Tube Clamps" and whether GDI had performed any seismic analyses of their 

product for a specific application. GDI stated that the use of the word 

"seismic" in the product's name was not to imply that the product had been 

seismically qualified to a specific set of seismic parameters but that a 

licensee could use the product's specified load capacities, along with data on 

the fastener and mounting surface in the licensee's seismic calculations to 

determine if the system (tube clamp, fastener, and mounting surface) would 

meet the site seismic requirements. Based on a review of the GDI product 

1 iterature, customer purchase orders, and documentation provided to the 

customers, the inspectors concluded that, although the word "seismic" is a 

part of the product name, GDI did not represent their product as seismically 

qualified in their product literature or in documentation provided to the 

users at del ivery. 

4 

9 
3.4.2 Control of Purchased Materi a1 and Servi ces 

As noted in paragraph 3.2, GDI has an approved subcontractor, STI, who 

manufactures the cl amps using an investment casting process. STI a1 so 

performs all examinations and tests and heat treatment required by the 

material specification and issues a material test 'report and material 

certification to GDI . GDI has qual if i ed and approved STI 's qual i ty assurance 

program as meeting the requirements of Appendix B to 10 CFR 50 and Subarticle 

NCA 3800 of Section I11 of the ASME Code. GDI contracted with an outside 

auditor to perform the audits of STI. The same auditor was used for the last 

three annual audits and adequate documentation was available to establish the 

auditor's qualifications to perform these audits. 

The inspectors noted that, as corrective action for a 1993 customer audit, GDI 

purchased the services of Pacific Consulting Engineers (PCE) to review GDI's 

design calculations supporting the load capacity data sheets for seven types 

of tube clamps. PCE had documented this review in a letter to GDI dated 

June 17, 1993, which concluded that GDI's calculations were complete and 

correct, the sketches were labeled correctly with the calculations, the 

formulas used were per the applicable portions of the 1989 ASME code, Section 

111, Subsection NF and its addendum, allowable stresses used were per the ASME 

code and cut sheets for materials were per the ASME code. 

certified that the calculation process and math was done correctly and 

completely. 

for this service solely on the basis of the PCE engineer performing the review 

being a registered professional engineer and that GDI had not taken any 

actions to qual i fy PCE, or the performing engineer, as an approved suppl ier of 

a safety related service. 

Criterion VII, "Control of Purchased Material , Equipment, and Services," of 10 

CFR Part 50 Appendix B. (Nonconformance 99901283/95-01-02) 

The letter also 

However, discussion with GDI indicated that GDI had selected PCE 

This was identified as a nonconformance to 

3.4.3 Material Receiving, Identification, and Warehousing 

GDI's quality assurance manual contains a receiving checklist to document 

verification of material to purchase order requirements. The clamps are 

received in boxes marked with the material heat code (one heat of material in 

a box). Additionally, each clamp halve is individually stamped with the heat 

code. As a part of the receiving process, the heat code identification is 

verified on a sample basis and the chemical analyses/mechanical properties 

reported on suppliers certification is checked against the applicable material 

specification. A table containing the material specification requirements was 

posted in the receiving area. No dimensional inspections are performed at 

this time. GDI relies on the clamp manufacturer to perform this function. 

Material is stored on warehouse shelves (full containers only) or on a large 

table in the warehouse area (partial containers). 

markings by the inspectors indicated adequate traceability control in this 

ar.ea. 

Spot checking of heat code 

5 

10 
3.4.4 Training 

GDI has only one full time employee (the owner and president) and uses part 

time employees for several activities. GDI had a letter on file dated 

December 1, 1983, which stated that the listed employees had been trained in 

the Quality Assurance requirements of the company relative to their respective 

jobs. 

time employees and the work they performed. The last update had been in 1988 

which GDI stated had added the personnel presently used by GDI. 

activities included packaging, shipping, and secretarial duties. GDI stated 

that the president prepares packing lists and certification and the part time 

employees use the packing lists to assemble and box the material. 

This letter had been updated on several occasion to reflect new part 

These 

3.4.5 Internal Audits 

GDI's quality assurance manual requires internal audits to be performed every 

odd year. Due to the size of GDI (one full time employee), GDI contracted 

with an outside auditor to perform the internal audits of GDI. GDI had used 

the same auditor for the previous three audits and had documentation onsite to 

establish his qualification for performing the audits. 

reviewed the three most recent audits which were performed July 18, 1989, 

May 23, 1991, and May 25, 1993. The audit reports appeared to document well 

performed internal audits, however, the auditor failed to identify the lack of 

design review for the GDI tube clamps or the material certification issues 

discussed in Sections 3.4.6 and 3.4.8 of the report, respectively. 

The inspectors 

3.4.6 Audits by Customers 

GDI has supplied numerous licensees with safety related tube clamps and, 

consequently, has been audited by many 1 icensees and other organizations. 

inspectors reviewed audit reports and re1 ated correspondence for a number of 

these audits including audits performed by Tennessee Val ley Authority on 

February 2, 1988, Wisconsin Electric Power Company on February 16, 1989, 

Florida Power and Light Company on April 27, 1989, Louisiana Power and Light 

on June 27, 1989, Washington Public Power Supply System on August, 29, 1990, 

Flour Constructors International on February 4, 1992, a NUPIC Joint Utility 

Audit led by Southern California Edison Company on April 20, 1993, and Bechtel 

on September 8, 1993. 

The 

The inspectors noted that Wisconsin Electric Power Company (WEPC) documented 

the February 16, 1989, supplier audit of GDI in a letter dated March 16, 1989. 

This letter contained a comment stating that "Drawing and load capacity sheets 

have no independent review. This does not meet the requirements of 10 CFR 

Part 50 Appendix B nor standard engineering practices." Although the WEPC 

letter requested a written GDI response to all items, including comments, GDI 

had no documentation on file to indicate that a response to this comment had 

been provided. 

findings, it would appear that GDI had not provided WEPC with a written 

response to the comment. 

Since there was documentation on file related to other WEPC 

6 

11 
Washington Public Power Supply System (WPPSS) had performed a supplier audit 

of GDI on August 29, 1990. The WPPSS letter to GDI dated October 2, 1990, 

documenting the audit, indicated that there was no objective evidence of an 

independent design review. The letter continued on to say "as GDI is a one 

person organization, the Supply System has agreed to perform this independent 

review." 

NUPIC had performed a Joint Utility audit of GDI on April 20 through 22, 1993. 

The audit led by Southern California Edison, was documented in Audit Report 

No. GIRD-1-93, and transmitted with a letter dated May 20, 1993. The NUPIC 

audit covered areas such as purchase documents, design, procurement , material 

control, handling, storage and shipping, document control, and program 

compliance. 

Corrective Action Request (CAR) S-1428 which stated, in part, that no 

objective evidence was available to support that an independent verification 

of design adequacy was performed. Subsequently, Girard had its design 

calculations reviewed by Pacific Consulting Engineers (PCE) as a corrective 

action to the CAR issued by NUPIC. The inspectors also noted that the NUPIC 

audit report, in the section titled "Quality History," stated that GDI had 

been audited by WPPSS in August 1990, and that no deficiencies were 

identified. The inspectors observed that, although the 1990 WPPSS audit 

report did not characterize the lack of design verification by GDI as a 

deficiency, it did clearly point this out as a weakness in the GDI quality 

assurance system. 

of GDI's quality history. 

The NUPIC audit identified one deficiency documented on 

The inspectors did not agree with NUPIC's characterization 

3.4.7 Nonconformances and Corrective Action 

Although GDI's quality assurance manual discusses corrective action for 

defective hardware discovered before or after delivery to the customer, it 

does not address corrective actions for other identified nonconforming 

conditions such as component design. 

The inspectors observed that a nonconforming condition in the tube clamp 

design had been identified to GDI in 1989 by WEPC and in 1990 by WPPSS. On 

both occasions the licensees decided to perform their own review of the GDI 

design to verify adequacy for their application. In each case, this review 

was applicable only to the utility performing the review, and did not suffice 

as a design verification supporting any additional safety related sales by 

GDI. GDI did not take any corrective action for the lack of design 

verification until the item was presented as a finding in the 1993 NUPIC audit 

which required corrective action prior to the purchase of the GDI product. 

The inspectors concluded that a condition adverse to quality had been 

identified to GDI by a licensee in 1989 and 1990 but GDI had not taken any 

corrective action until 1993. This was identified as a nonconformance to 

Criterion XVI, "Corrective Action," of 10 CFR Part 50 Appendix B. 

(Nonconformance 99901283/95-01-01) 

7 

12 
3.4.8 Certification of Material 

Selected customer purchase orders (PO) and re1 ated documents were reviewed to 

determine whether adequate controls were provided by GDI to assure compliance 

with the customer specified technical and quality requirements. The 

inspectors observed that the clamps were generally ordered and supplied to 

quality assurance requirements of Appendix B to 10 CFR 50 and that numerous 

orders also required compliance with the applicable requirements of the ASME 

Code. 

One of the POs reviewed was Georgia Power Company’s (GPC) order 60191890000, 

dated November 7, 1994, for 12 part number 3/8 T 2D clamps. The material was 

specified as SA 351 Grade CF8, to be supplied as meeting ASME Code, Section 

111, Class 2 requirements. Additionally, the purchase documents required the 

material to be provided in accordance with GDI’s quality assurance program, 

revision 22 conforming to ASME Code, Section 111, Subsection NCA 3800. The 

inspectors noted that GDI’s files did not contain any documentation indicating 

that GDI had aver been audited and qual ified by GPC. 

GDI supplied this material along with their Certificate of Conformance (CoC) 

dated November 7, 1994, attesting compliance with the PO requirements and 

containing a statement that this material meets the requirements of ASME Code, 

Section 111, Class 2. GDI also provided a material certification issued by 

STI, dated June 11, 1985, which certified that the material meets ASTM 

Specification A 351 and was produced in accordance with the ASME Code, Section 

111, Subsection NCA, Paragraph 3800, as applicable. 

ST1 also certified that the material had been heat treated to 1900 degrees F 

and water quenched per ASTM A 743, and that the material was produced in 

accordance with STI ’s qual i ty assurance manual , dated July 17, 1983, Revision 

6. STI’s certification included a test report of the chemical and physical 

properties of the material , provided to STI by Spectrochemical Laboratories, 

Pittsburgh, PA. 

The GPC PO was typical of several other 1 icensee orders which invoked ASME 

Code requirements. 

orders, the inspectors concluded that several of these procurements may not 

fully comply with the ASME Code. 

After a review of the documentation related to these 

In general, unless the licensee is an ASME certificate holder, the licensee 

can not purchase material from suppliers that have not been accredited by ASME 

(GDI is in this category) and be in compliance with the ASME Code 

requirements. Material can be certified as meeting Code requirements only by 

organizations that possess Quality System Certificates (QSC), or N-Type 

Certificates i ssued by ASME. 

certificates can, however, certify that the material has been produced in 

accordance with a quality assurance program which meets the requirements of 

Subsection NCA, subarticle 3800 of the Code. Under these circumstances, an 

ASME certificate holder can accept (and sell) this material and certify that 

the material meets the Code requirements, providing that this certificate 

Material manufacturers not holding such 

8 

13 
holder has qualified (by audit) the quality assurance program referenced in 

the manufacturer's certification as complying with NCA 3800. 

procurement chain in several of the document packages reviewed apparently did 

not include any organizations accredited by the ASME, such material should not 

have been certified as meeting ASME Code requirements. Certification of 

material to the requirements of the ASME Code without fully complying with the 

applicable Code provisions was identified as a nonconformance to Criterion V, 

"Instructions, Procedures and Drawings" of 10 CFR Part 50 Appendix B. 

(Nonconformance 99901 283/95-01-03) 

Since the 

4 PERSONS CONTACTED 

The NRC staff participating in the inspection and GDI personnel contacted 

during the inspection are listed below. All individuals listed attended the 

entrance and exit meetings . 

6i rard Development , Inc. 

Girard, D. President 

U.S. Nuclear Regulatory Conmission 

Potapovs, U. 

Rogers, B. H. 

Team Leader, VIS/TSIB 

Reactor Engineer, VIS/TSIB 

9 

14 
UNITED STATES 

WASHINGTON, D.C. 20565-oOoi 

NUCLEAR REGULATORY COMMISSION 

fiZarch 22, 1995 

Mr. William Blackwell, President 

Mackson Inc. 

2346 Southway Drive 

Rock Hill, SC 29731 

SUBJECT: NRC INSPECTION NO. 99901179/95-01 

Dear Mr. Blackwell: 

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

of your facility at Rock Hill, South Carolina, conducted by Messrs. 

U. Potapovs and R.P. McIntyre of this office on February 15 through 17, 1995, 

and the discussions of their findings with you and members of your staff at 

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

your quality assurance program and its implementation in selected areas such 

as (1) control of purchased material and services, (2) material and 

traceability control, (3) supplier audits, and (4) a review of your program 

for implementing Part 21, "Reporting Defects and Noncompliance," of Title 10 

of the Code of Federal Requlations (10 CFR) . 

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

enclosed inspection report. 

procedures and representative records, discussion, and observations by the 

inspectors. 

This inspection consisted of an examination of 

Our review of your commercial grade item dedication activities indicated that, 

a1 though your qual i ty assurance program had establ i shed good controls in 

several areas, your material sampling practices did not fully meet NRC 

requirements. Specifically, your prograin did not contain a basis F objective 

evidence to show that your destructive sampling plan for verifying critical 

characteristics of material provides reasonable assurance that the material 

meets all of the applicable procurement document requirements. The inspection 

also identified certain weaknesses in the areas of ASME Code material 

upgrading and in the documentation of corrective actions. 

findings and references to the pertinent requirements are identified in the 

enclosures to this letter. 

The specific 

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

statement in accordance with the instructions specified in the enclosed Notice 

of Nonconformance. 

show good cause for us to do so. 

We will consider extending the response time if you can 

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

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

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

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

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

15 
W. Blackwell -2- 

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

discuss them with you. 

Sincerely, 

Robert M. Gallo, Chief 

Special Inspection Branch 

Division of Technical Support 

Office of Nuclear Reactor Regul ati on 

Docket No. 99901179 

Enclosures : I. Not i ce of Nonconformance 

2. Inspection Report 99901179/95-01 

16 
NOTICE OF NONCONFORMANCE 

Mackson , Incorporated Docket No.: 99901179/95-01 

Rock Hill, South Carolina 

Based on the results of an NRC inspection conducted on February 15 through 17, 

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

accordance with NRC requirements. 

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

of Appendix B to Title 10 of the Code of Federal Reaulations (10 CFR) 

Part 50, requires, in part, that measures shall be established to assure 

that purchased material conforms to procurement documents. 

Contrary to the above, Mackson had not established a documented basis to 

substantiate that its sampling plan for verifying critical 

characteristics such as dimensions and chemical and mechanical 

properties provided reasonable assurance that dedicated commerci a1 grade 

items supplied met applicable procurement document requirements. (99901179/95-01-01) 

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

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

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

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

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

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

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

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

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

and preventive measures were or will be completed. 

Dated at Rockville, Maryland 

this day of , 1995 

17 

Enclosure 1 
REPORT NO. : 

ORGANIZATION: 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY: 

INSPECTION DATES: 

LEAD INSPECTOR: 

OTHER INSPECTORS: 

REVIEWED BY: 

APPROVED BY: 

U. S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF TECHNICAL SUPPORT 

99901179/95-01 

Mackson, Inc. 

2346 Southway Drive 

Rock Hill, South Carolina 29730 

Thomas Sharp, Manager 

Quality Assurance 

Mackson, Inc. is a supplier of threaded fasteners and 

metal products. Fastener products comprise more than 

90% of their business. 

February 15 through 17, 1995 

Vendor Inspection Section (VIS) 

Special Inspection Branch (TSIB) 

Richard P. McIntyre, VIS/TSIB 

Veniior Inspection Section 

Special Inspection Branch 

Date 

Enclosure 2 
1 SUMHARY OF INSPECTION FINDINGS 

During this inspection, the NRC inspectors evaluated Mackson Inc.’s 

(Mackson’s) impJementation of quality assurance measures for those activities 

which directly affected the quality and performance capability of their 

product. 

services, audits, material traceability, and commercial grade item dedication. 

The team also reviewed Mackson’s program for implementing Part 21, “Reporting 

Defects and Noncompliance,” of Title 10 of the Code of Federal Resulations 

(10 CFR). 

These activities included the control of purchased material and 

The inspection basis consisted of the following: 

Appendix 8, “Quality Assurance Criteria for Nuclear Power Plants and 

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

Federal Resulations (10 CFR Part 50) 

10 CFR Part 21 

Section I11 of the ASME Boiler and Pressure Vessel Code (Code) 

Mackson’s Qual 

1.1 Violations 

No violations were ident 

ty Systems Manual, Revision 4, dated October 11, 1994 

fied during this inspection. 

1.2 Nonconformances 

1.2.1 Nonconformance 95-01-01 

This nonconformance, described in Section 3.4.2.2 of the report identifies the 

failure of Mackson’s quality assurance program to establish a documented basis 

to substantiate that its sampling plan for verifying critical characteristics 

provides reasonable assurance that the dedicated items meet all of the 

applicable procurement specification requirements. 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

This was the first NRC inspection of Mackson 

3 INSPECTION FINDINGS AND OBSERVATIONS 

3.1 Entrance and Exit Meetinss 

During the entrance meeting on February 15, 1995, the NRC inspectors discussed 

the inspection scope and developed general information about Mackson’s 

products and activities. 

NRC inspectors discussed their findings and observations with Mackson 

management. 

During the exit meeting on February 17, 1995, the 

2 

19 
3.2 Description of Facilities 

Mackson has been accredited by the American Society of Mechanical Engineers 

(ASME) to supply ferrous and nonferrous bars , threaded fasteners , castings, 

forgings, plate, seamless fittings, flanges, NPT stamped tubular products, and 

similar items. The scope of their Quality Systems Certificate (QSC) also 

includes the qualification of material manufacturers and suppliers and 

suppliers of subcontracted services as well as upgrading of stock material. 

Their QSC expires on March 3, 1997. 

According to Mackson management , threaded fastener products currently comprise 

more than 90% of their nuclear sales. Most of their stock material larger 

than 1/4 inch diameter is reportedly purchased from qualified suppliers/manufacturers although some upgrading is performed. 

upgrading and dedication activities including chemical analyses, mechanical 

testing, and nondestructive examination (NDE) are subcontracted to qualified 

vendors. In-house activities are limited to receiving inspection (visual), 

dimensional inspection, and warehousing. No machining operations are 

performed. 

Typical 

3.3 10 CFR Part. 21 Proqram 

3.3.1 Procedures and Implementation 

The latest issue of 10 CFR Part 21 along with Section 206 of the Energy 

Reorganization Act of 1974 and Mackson's implementing Quality Procedure V-A, 

"Reporting lOCFR21 Discrepancies," Revision 2, dated May 31, 1994, were 

observed to be posted as required by the regulation. 

imp1 ement ing procedure determined general compl i ance with the regul ati on. The 

inspectors noted that in several instances the term I' defect" 

pl ace of "devi at i on. 'I These terms are speci f i call y defined in the regul ati on 

and can not be interchanged. 

Review of the 

was used in 

This weakness was discussed with the QA Manager who indicated that the 

implementing procedure would be revised to address the concern. 

3.4 Quality Assurance Proqram 

Mackson's Quality Assurance Program is described in their Quality Systems 

Manual (QSM), Revision 4, dated October 11, 1994, which is committed to 

meeting the applicable requirements of Appendix B to 10 CFR Part 50, 

Subarticle NCA 3800 of Section I11 of the ASME Code, and the American National 

Standards Institute (ANSI) Standard N45.2. The quality assurance program is 

implemented through the procedures identified in the Mackson Quality Procedure 

Manual (QPM), Revision 6, dated July 8, 1994. Selected areas of the manual 

and its implementation were examined during this inspection as indicated 

bel ow. 

3.4.1 Material Supplied to ASME Code Requirements 

Mackson's QSM and its implementing procedures require that purchase orders 

(PO) for Code material and services be placed only with vendors appearing on 

3 

20 
Mackson’s Approved Vendor List (AVL). Vendors are placed on the AVL based on 

holding a valid ASME QSC or following the completion of a satisfactory survey 

by Mackson in accordance with the requirements of Subarticle NCA 3800 of ASME 

Code, Section 111. 

Mackson’s procedures for upgrading of stock material were reviewed and 

determined to be in compliance with the applicable requirements of the ASME 

Code. 

applicable material specification on each piece of the stock material or, when 

traceability has been established, the performance of the required tests on 

one sample of each heat of the material. 

each piece of the material in either case. 

These procedures require the performance of all tests specified in the 

Chemical analysis is required on 

Small products as defined in paragraph NX-2610 of ASME Code, Section I1 are 

supplied under the provisions of that paragraph only when authorized by the 

customer. Mackson’s procedures require that materi a1 suppl ied under NX 2610 

comply with the requirements of NCA-3867.4 and NCA-3866.6 with respect to 

certification and identification/marking, respectively. Mackson’s procedures 

permit such material to be supplied under their commercial grade dedication 

program. Specifically, the materi a1 manufacturer’s commerci a1 grade qual i ty 

program is reviewed by Mackson and the material is sampled for chemical 

composition and mechanical properties initially and during every fifth 

shipment thereafter. 

Mackson’s QSC number. 

Material supplied in this manner is not certified under 

3.4.1.1 Control of Subtier Vendor; and Documentation of Basis for Material 

Certi f icati on 

Mackson purchases most of the material to be supplied under their QSC from 

suppliers that have been qualified by survey as having quality assurance 

programs meeting the requirements of NCA-3800. Mackson may perform additional 

testing on this material before issuing a certified material test report 

(CMTR). To evaluate Mackson’s quality assurance program implementation, the 

inspectors reviewed several recently shipped customer orders as summarized in 

the following paragraphs. 

Duke Power Co. PO F45588-K3, dated October 26, 1994, for 1 1/8-7-2A 

threaded rod conforming to SA 193-GR7, in accordance with ASME Sec. 111. 

Mackson’s certification of this material was accompanied by a 

Certificate of Conformance (CoC) issued by Plainville Mfg. Co. (PM) who 

is on Mackson’s AVL as a material manufacturer qualified in accordance 

with ASME NCA-3800. PM’s CoC included chemical analysis, tensile test 

results, impact test results, hardness test results, magnetic particle 

test results, and contained a statement that the raw material was 

supplied by North Star Steel Co. 

mill heat analysis from North Star Steel Co. were provided with PM’s 

COC. 

The corresponding test reports and the 

PM’s CoC did not describe the material heat treatment, however, the file 

contained a heat treatment certificate issued by Copperweld Steel Co. 

4 

2 1. 
This certificate was issued to Hy Alloy Steel Co., while the material 

was shipped to A.M. Castle & Co., who apparently supplied the steel to 

PM . 

None of the manufacturers/suppl iers besides PM were 1 isted on Mackson’s 

AVL and it could not be determined how some of the lower tier vendors 

were qualified to provide material or services. 

The inspectors noted that under the provisions of NCA-3800, material suppliers 

who do not possess QSCs may not qualify other material suppliers. 

unless all of the subtier vendors were on PM’s AVL, the material obtained from 

A.M. Castle would need to be treated as stock material (each piece tested) and 

the heat treatment certification by Copperweld could not be considered as 

validated. 

Mackson’s management for potential followup with their suppliers. 

Thus, 

The subtier vendor qualification issue was discussed with 

Carolina Power and Light PO 7N3894ASY dated April 01, 1994, for 1 1/4-8 

threaded rod conforming to SA 193-GRB7, in accordance with ASME Sec. 

111, Class 2, 1974 Edition, 1976 winter Addenda. 

Mackson’s certification of this material was based on a CMTR issued by 

A&G Engineering, dated January 29, 1993, under their ASME QSC. (Note: 

A&G’s QSC was terminated on March 18, 1993, for nonpayment of fees). 

A&G’s CMTR reported the material’s chemical analyses, tensile 

properties, impact test results, heat treated condition, and attested 

that visual and magnetic particle testing had been performed. Test 

reports supporting the information included in the CMTR were not 

available. The material producing mill was not identified on the CMTR 

and the original mill heat analysis report was not provided. 

The inspectors noted that, while Mackson’s procurement of this material based 

only on a CMTR issued under A&G’s QSC appeared to have complied with the ASME 

Code requirements at the time of this procurement, the material traceability 

to the producing mill was not demonstrated by the available documentation. 

The inspectors also advised Mackson management that the NRC was preparing to 

issued an Information Notice advising licensees that A&G may have ipplied 

potentially nonconforming fasteners to the nuclear industry. The inspectors 

further noted that Mackson did not appear to implement a consistent policy in 

determining what quality records need to be provided by their suppliers. This 

concern is demonstrated in the two examples discussed above and was identified 

as a weakness in Mackson’s quality program. (Also, see discussion in 

paragraph 3.4.4) 

3.4.2 Non-Code Material Supplied to 10 CFR 50, Appendix B Requirements 

(Commercial Grade Item Dedication Program) 

3.4.2.1 Methodol ogy 

The requirements for Mackson’s commercial grade item (CGI) dedication process 

are prescribed in Section 111-1, “Dedication of Commercial Grade Material ,‘I 

Revision 2, dated July 8, 1994, of the Mackson QPM. The following sections of 

5 

22 
the QPM provide additional requirements for performing and documenting various 

dedication activities and were reviewed during the inspection: 

IV-A, "Visual Exam of Bolts, Screws, and Studs," Revision 1, dated 

January 7, 1992 

IV-By "Visual Exam of Nuts," Revision 1, dated January 7, 1992 

IV-C, "Visual Exam of Threaded Rod," Revision 1, dated January 7, 1992 

IV-D, "Visual Exam of Washers," Revision 1, dated January 7, 1992 

IV-E, "Visual Exam of Piping Material," Revision 1, dated 

January 7, 1992 

IV-M, "Verification of Material to P.O.," Revision 1, dated 

January 7, 1992 

IV-N, "Material Shipped to Service Vendors," Revision 1, dated 

Januarv 7, 1992 

IV-0, "Material For Chemical and Physical ,'I Revision 1, dated 

January 7, 1992 

XII-By "Proper Use of Thread Ring Gages," Revision 0, dated 

February 26, 1993 

XII-Cy "Proper Use of Thread Plug Gages," Revision 0, dated 

February 26, 1993 

XII-D, "Proper Use of Calipers, Revision 0, dated February 26, 1993 

XI I-E, "Qual i fication Requirements for Dimensional Inspector, I' 

Revision 0, dated February 26, 1993 

Mackson selects critical characteristics for CGIs based on the testing 

requirements specified by the applicable ASME Boiler and Pressure Vessel Code, 

Section 11, "Materials," or American Society for Testing and Materials (ASTM) 

material specifications and any other testing requirements specified by the 

customer. For purposes of testing, the critical characteristics of the 

material include the chemical, mechanical, and dimensional requirements of 

the applicable material specification. Testing is performed by testing 

laboratories who are lis 

verification of critical 

discussed in Section 3.4 

Material to be dedicated 

traceable to the materia 

from the Mackson list of 

ed on the Mackson Approved' Vendor List. 

characteristics is performed using sampling plans as 

2.2 of this inspection report. 

by Mackson, is purchased, when possible, with CMTRs 

manufacturer. Material suppliers of CGIs are chosen 

"SuDDliers ADDroved Based on Performance Record." 

The - 

Suppliers are placed on this'iist aftel the receipt and testing of six 

consecutive purchases identifies zero deficiencies with the critical 

6 

23 
characteristics of the material. 

sample size required for visual and dimensional inspections is decreased by 50 

percent. However, the accept/reject rate remains the same. 

Once a supplier is placed on this list, the 

The total verification of critical characteristics for CGIs consists of a 

receipt inspection, visual and dimensional inspections, and a review of the 

results from chemical and mechanical tests performed on a sample of the 

material, and serve as the basis for Mackson to issue a certification 

statement that the dedicated CGIs meet the applicable material and customer PO 

requirements. 

program addresses the essential elements of the dedication process and that 

sufficient guidance for performing activities such as inspection and testing 

are provided. 

The NRC inspectors concluded that Mackson’s CGI dedication 

The NRC inspectors considered the following to be strengths in Mackson’s 

dedication program: 

Mackson selects critical characteristics based on the testing and 

material requirements of the applicable material specification and its 

customers. 

Mackson attempts to obtain material certification for each item or lot 

of i tems suppl i ed from the manufacturer (qual i f i ed and non-qual i f i ed 

manufacturers) and reviews the certification, when received, for 

conformance with the applicable material specification. 

The NRC inspectors considered the following to be a weakness in Mackson’s CGI 

dedication program: 

Mackson has no documented bases, (qualitative or quantitative) for its 

dimensional and destructive sample plan used in verifying critical 

characteristics (see Section 3.4.2.2 of this report). 

3.4.2.2 Use of Sampling in the CGI Dedication Process 

Materi a1 purchased as commerci a1 grade from nonqual if ied suppl iers or from 

suppliers included on the list of Suppliers Approved Based on Performance 

Record and dedicated as basic components, are inspected and tested in 

accordance with the foll owing sampl e pl ans unl ess a customer specifies another 

sampl i ng pl an. 

1. Visual inspection is performed as follows for nonqualified suppliers: 

LOT SIZE SAMPLE SIZE ACCEPT/REJECT* 

0-50 

51-90 

91-150 

151-280 

281-500 

501-1200 

16 

26 

40 

64 

100 

160 

2 j3 

314 

7 

24 
2. 

3: 

4. 

1201-3200 

3201-10000 

250 

400 

516 

718 

Vi sua1 inspecti on is performed as foll ows for suppl i ers approved based 

on performance record: 

LOT SIZE 

0-50 

51-90 

91-150 

151-280 

281-500 

50 1- 1200 

1201-3200 

3201-10000 

SAMPLE SIZE 

8 

13 

20 

32 

50 

80 

125 

200 

ACCEPT/REJECT* 

Dimensional inspection is performed as follows for nonqualified 

suppliers: 

LOT SIZE 

2-8 

9-15 

16-25 

26-50 

5 1-90 

91-150 

15 1-280 

281-500 

501-1200 

1201-3200 

3201-10000 

SAMPLE SIZE 

4 

4 

4 

4 

6 

6 

6 

6 

10 

10 

10 

ACCEPT/ RE J ECT* 

Dimensi onal inspect i on i s performed as fol 1 ows for suppl i ers approved 

based on performance record: 

LOT SIZE 

2-8 

9-15 

16-25 

26-50 

51-90 

91-150 

151-280 

281-500 

501-1200 

1201-3200 

3201-10000 

SAMPLE SIZE 

2 

2 

2 

2 

3 

3 

3 

3 

5 

5 

5 

8’ 

25 
5. According to Mackson, destructive testing to verify critical 

characteristics such as chemical and mechanical properties are performed 

on fasteners (bolts, screws and studs) and other CGIs being dedicated in 

accordance with the fol 1 owing sample pl an : 

LOT SIZE SAMPLE SIZE ACCEPT/ REJECT" 

0-50 

51-500 

501-35000 

2 

3 

5 

* Accept or Reject lot on this number of defects per sample size. 

It is also noted that Mackson states a lot shall be defined as a single 

material type, size and heat. 

The NRC discussed the use of Mackson's sampling methodology in detail during 

the inspection and asked Mackson to provide the bases for its destructive 

sampling plan as providing reasonable assurance that the fasteners not tested 

had the required material chemistry and strength and other properties required 

by the applicable material specification. According to Mackson, the sampling 

plans discussed above are not based on statistical methods but are considered 

adequate for the required testing. Mackson uses its destructive sampling plan 

for dedicating CGIs regardless of the size or type of product, whether the 

CGIs were purchased directly from the manufacturer or through a distributor, 

the use of single or multiple production lots, or the historical performance 

of the manufacturer or distributor. 

Mackson that such a sample size could include more than one production lot or 

heat of material. 

The NRC inspectors expressed a concern to 

Mackson informed the NRC inspectors that it had not established a quantitative 

basis for their sampling plan nor an associated confidence level. 

dimensional and destructive sample plans appear to be based on undocumented 

qualitative factors such as engineering judgement and experience of its 

personnel in inspecting and testing fasteners, and not on using qudntitative 

statistics. 

quality requirements of Appendix B to 10 CFR Part 50. 

Mackson's 

The inspectors stated that this sampling method does not meet the 

(Nonconformance 

99901179/95-01). 

3.4 

The 

the 

ver 

2.3 Review of Mackson CGI Dedication Program Implementation 

NRC inspectors reviewed several completed CGI dedications to determine if 

critical characteristics for the items had been properly identified and 

fied, and if adequate procedural controls were in place. The NRC 

inspectors reviewed the following completed CGI dedications. 

Mackson Purchase Order (PO), C000387, dated January 17, 1995, was for 

the purchase of 1800 heavy hex nuts, 5/8-11, ASTM A563, Grade C, in 

accordance with Duke Power PO F28721-C4. 

from Decker Manufacturing Corporation, Albion, Michigan, a nonaudited 

supplier that was on the list of Suppliers Approved Based on Performance 

Mackson purchased these nuts 

9 

26 
e 

e 

Record. 

identification and overall workmanship) on 250 of the 1800 items and 

performed a dimensional exam of selected dimensions on 10 of 1800 items. 

Mackson also reviewed the nonqual ified manufacturer’s material test 

report for conformance with the material specification requirements and 

subcontracted to an approved non-code vendor for performance of chemical 

analyses on 52 of the 1800 nuts, a proof load test and a hardness test 

on 52 of the 1800 nuts. 

Mackson performed a receipt inspection, visual exam (markings, 

The test results met applicable material specification requirements and 

were, in general , consistent with the nonqual ified manufacturer’s 

material certification. Based on the above dedication activities, 

Mackson certified that the nuts were supplied in accordance with 

customer PO requirements and the Mackson Quality System Manual. 

Mackson PO, C002829, dated July 22, 1994, was for the purchase of 50 

heavy hex bolts, 1-8 by 6 inches long, ASTM A325, Type 1, and 25 14-7 by 64 inches long in accordance with Duke Power POs F23053-K3 and F30399- 

K3. Mackson purchased these bolts from a distributor, House of Threads, 

Charlotte, North Carolina, a nonaudited supplier that was on the Mackson 

list 01 Suppliers Approved Based on Performance Record. NUCOR 

Fasteners, St. Joe, Indiana, the manufacturer of the bolts, is listed on 

the Mackson Non-Code Approved Vendors List for traceabil ity only. 

However, during the inspectors review of the 1993 Mackson survey of 

NUCOR, it was identified that NUCOR does not audit subtier vendors and 

therefore would question the ability to maintain traceability. 

Mackson supplied a total of 17 (6 1-8s and 11 14-7s) House of Threads 

bolts to Duke as part of these POs. 

Duke were purchased by Mackson from an approved Part 50 Appendix B 

supplier. Mackson performed a receipt inspection, a visual exam 

(markings, identification and overall workmanship) on 8 of each order 

items and a dimensional exam of selected dimensions on 2 of each order 

i tems. Mackson a1 so reviewed the nonqual i f i ed manufacturer’s materi a1 

test report for conformance with the material specification 

requirements, and subcontracted chemical analyses, and tests of 

mechanical properties on 2 of each order of bolts. 

The remaining bolts supplied to 

The test results met applicable material specification requirements and 

were, in general , consistent with the nonqual ified manufacturer’s 

material certification. Based on the above dedication activities, 

Mackson certified that the bolts were supplied in accordance with 

customer PO requirements and the Mackson Qual i ty System Manual . 

Mackson PO, C001732, dated April 29, 1993, was for the purchase of 200 

socket head cap screws of 5 different sizes ranging from 5/16-18 by 24 

to 3/4-10 by 24, A574 in accordance with 5 Duke Power POs. Mackson 

purchased these bo1 ts from Georgia Fasteners, Augusta and Savannah, 

Georgia, another nonaudited supplier that was on the Mackson list of 

Suppliers Approved Based on Performance Record. Mackson performed a 

receipt inspection, a visual exam (markings, identification and overall 

workmanship) on 32 of 200 items for each of the 5 different sizes and a 

10 

27 
dimensional exam of selected dimensions on 3 of 200 items for each of 

the 5 different size cap screws. Mackson also reviewed the nonqualified 

manufacturer's material test report for conformance with the material 

specification requirements. In this case Georgia fasteners did have the 

CMTRs from the steel mill. A Mackson approved non-code vendor performed 

chemical analyses and mechanical properties tests on 3 of 200 for each 

of the 5 different sizes bolts. 

The test results met applicable material specification requirements and 

were, in general , consistent with the nonqual ified manufacturer's 

material certification. Based on the above dedication activities, 

Mackson certified that the bolts were supplied in accordance with 

customer PO requirements and the Mackson Quality System Manual. 

3.4.3 Audits 

3.4.3.1 Internal Audits 

The inspectors reviewed Section 11, "Audits," of Mackson's Quality System 

Manual, and Section 111-A, "Vendor Audit" of the Quality Procedure Manual. 

The inspectors reviewed the lead auditor qualification and certification 

records for the Mackson QA Manager and also for two outside subcontractors who 

have performed internal audits of Mackson. All documentation appeared 

current. The team a1 so reviewed re1 ated internal audit documentation such as 

annual audit plans, audit checklists, audit reports and responses to audit 

report findings. The team verified that appropriate corrective action was 

performed by the Mackson QA Manager for internal audit report findings. 

3.4.3.2 External Audits 

External vendor audits are predominantly performed by the QA Manager, but the 

President of Mackson is also an certified lead auditor and performs some 

external audits along with the QA Manager. Mackson maintains a separate 

Approved Vendor List for ASME code vendors and non-code vendors. The 

inspectors reviewed the most recent audits of Plainville Manufacturing Company 

and Texas Bolt Company (ASME code vendors) and JLK Industries, Inc. and NUCOR 

Fasteners (non-code vendors). A1 1 audits reviewed i ncl uded appropriate 

documentation of objective evidence for the areas reviewed. However, the 

inspectors did identify that the JLK and NUCOR audits only verified 

traceability control at the audited vendor, and did not verify sub-tier vendor 

traceability. 

The team concl uded that Mackson was imp1 ement ing an appropriate internal and 

external audit program per their QSM requirements. 

3.4.3.3 Audits by Customers 

To evaluate the effectiveness of audits by Mackson's nuclear customers, the 

inspectors reviewed the latest audit performed by the Nuclear Utilities 

Procurement Issues Committee (NUPIC). The audit was conducted on March 16-19, 

1993, with South Carolina Electric & Gass Company providing the lead auditor. 

The audit resulted in two findings: failure to verify thread dimensions as a 

11 

28 
part of the dedication process, and failure to present a Nuclear Quote Sheet 

to the Quality Assurance Manager for Approval. 

satisfactorily corrected. 

procedure (111-1, Revision 2) was reviewed and determined to be satisfactory. 

The report did not address sampling plans used in the commercial grade 

materi a1 dedi cation process. 

Both findings were 

Mackson’s commercial grade material dedication 

3.4.4 Nonconforming Material and Corrective Actions 

Section V-B of Mackson’s Quality Procedures Manual provides guidance for the 

control and disposition of items and activities that are found to be in 

nonconformance with specified requirements. Additionally, the Quality 

Assurance Manager maintains a nonconformance log which provides the status of 

nonconforming conditions. The inspectors reviewed the nonconformance log and 

noted that the majority of the entries had been properly dispositioned in a 

timely manner. A review of selected nonconformance reports indicated that, in 

general , adequate documentation was provided to describe the nonconformance, 

identify the root cause and support the method of disposition. However, the 

review identified several concerns with the evaluation and disposition process 

as summarized bel ow. 

NCR 2161, dated July 27, 1994 was issued to document apparently 

incorrect material (1/4 inch hexagonal cap screws) supplied by Non 

Ferrous Bolt & Manufacturing Co. 

Grade B8M. The material (heat number 701522) was supplied with a CMTR 

from Carpenter Technology certifying the Carbon content less than .03%. 

Verification tests performed by Mackson’s contractor showed the Carbon 

content of this material to be .064%. 

The specified material was ASTM-A 193, 

Replacement material, certified to the same heat number, was tested and 

found to be within specification suggesting that cap screws from a 

different heat of material than identified in the accompanying CMTR had 

been provided in the initial shipment. 

The inspectors noted that, a1 though the replacement material had been 

accepted in August, 1994, the NCR had not been closed and the root cause 

of this occurrence had not been identified on the NCR form. 

NCR 2167, dated November 15, 1994, was issued to document material with 

impact values below specified requirements received from Texas Bolt Co. 

Texas Bolt had supplied 20 SA-194 GR7 nuts with a CMTR showing impact 

values below the specified minimum. Mackson identified the discrepancy 

during receiving inspection. 

The nuts were returned to the vendor with a request for corrective 

action/root cause identification. The vendor responded by stating that 

the impact values were low because the test samples had been taken from 

finished nuts rather than from a test coupon. 

included reheat treating the nuts with a test coupon from the same heat 

of material and retesting. The results were acceptable and the NCR was 

closed on November 18, 1994. 

Corrective action 

12 

29 
e 

e 

The inspectors noted that the vendor’s root cause analysis appeared to 

be inadequate in that it did not address the question of why the vendor 

shipped material with a CMTR showing impact test values below the 

speci f i ed requ i rements . 

NCR 2114, dated July 11, 1992, was issued to document visually 

identified cracking of SA-193 GR8 hexagonal cap screws supplied to 

Mackson by Texas Bo1 t Co. 

and Light Co. during receiving inspection and the material was returned 

to Mackson. 

The defects were identified by Carol ina Power 

Mackson’s root cause analysis determined that the material should not 

have passed vendor’s in-process examination and should not have been 

shipped in this condition. Corrective action was requested from the 

vendor. 

address the failure to identify these defects during Mackson’s receiving 

inspection. 

The inspectors noted that the root cause analysis did not 

NCR 2170, dated December 20,1994, was issued to document an instance 

where the material manufacturer (US Bolt Co.) was unable to provide a 

liquid penetrant test report for SA 193 GRB8 threaded rod after 

Mackson’s customer requested a copy of the test report (the CMTR 

certified that this testing had been performed) 

As a part of the corrective action, Mackson performed liquid penetrant 

testing on this material utilizing the services of a qualified 

contractor. 

The examples of inadequate root cause analyses and timeliness of NCR 

disposition were discussed with Mackson’s management as potential weaknesses 

in the implementation of the corrective action program. 

that the last example appeared to be directly related to a previously 

identified weakness (paragraph 3.4.1.1) concerning the lack of consistent 

policy in defining what quality records must be provided by Mackson’s 

suppliers. 

The inspectors noted 

4 PERSONS CONTACTED 

The NRC staff participating in the inspection and Mackson personnel contacted 

during the inspection are listed below. 

entrance and exit meetings. 

All individuals listed attended the 

Mackson Inc. 

B1 ackwell , W. 

Sharp, T. 

Muse, H. 

President 

Quality Assurance Manager 

Vice President, Operations 

U. S. Nuclear Regul atory Conmi ssion 

Potapovs, U. 

McIntyre, R. 

Team Leader, VIS/TSIB 

Sr. Reactor Engineer, VIS/TSIB 

13 

30 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-0001 

January 31, 1995 

Mr. Kaydell C. Bowles 

Manager, Quality Assurance 

Sandvik Special Metals Corporation 

P.O. Box 6027 

Kennewick, Washington 99336-0027 

SUBJECT : NRC INSPECTION NO. 99900764/94-01 

Dear Mr. Bowles: 

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

(NRC) inspection of Sandvik Special Metals Corporation (SSM), Kennewick, 

Washington, conducted on December 5 through 7, 1994. The NRC inspection team, 

led by Steven M. Matthews and comprising the other inspectors named in the 

report, conducted a performance-based evaluation of the SSM management, staff, 

and quality programs and the implementation of those programs related to the 

manufacture of zirconium alloy nuclear fuel clad tubing. The inspection was 

conducted to provide a basis for confidence that SSM zirconium alloy fuel clad 

tubing supplied to the U.S. nuclear industry in 'fuel assemblies would perform 

its safety function. 

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

representative records, (b) conducted interviews, (c) held discussions, 

(d) listened to presentations, and (e) made various observations. On the 

basis of this inspection, the NRC determined that, in one instance observed, 

the implementation of the SSM quality system, documented in the SSM Quality 

Assurance Manual (QAM), Revision 1, dated April 7, 1994, did not meet the 

requirements of Appendix B to Part 50 o f T i t l e 10 of the Code of Federal 

Requlations (Appendix B to 10 CFR Part 50). Therefore, we are issuing a 

Notice of Nonconformance (Enclosure 1) to SSM. The report (Enclosure 2) 

contains a detailed discussion of the areas examined and our findings. 

The NRC determined that SSM had not established adequate measures to control 

the identification of tube specimens after burst testing used to characterize 

the biaxial mechanical properties of zircaloy tubes. Specifically, the 

identification of tube specimens was not maintained in a manner that ensured 

SSM the ability to perform further analysis of the failed tube specimens, 

where the results of further analysis of the failed tube specimens may affect 

the quality of the zircaloy fuel clad tubing. In the instance observed, posttesting failure analysis was both requested by the fuel vendor and necessary 

to assign root cause for the reduced ductility of the tube specimens and to 

ensure that the corrective actions taken by SSM were adequate to preclude 

other failed tube specimens. 

31 
K. Bowles -2- 

The NRC identified strengths in the SSM problem analysis and corrective 

actions planned and taken to prevent further instances of the "whitish" 

discolorations observed on the outside diameter tube surfaces, documented in 

SSM Technical Report No. 94-009, "Evaluation of Paper Divider Trays," dated 

September 26, 1994, which was thorough, competent, and conclusive as to the 

root cause of the discolorations. 

For the nonconformances cited in the NRC report of its previous inspection of 

SSM (documented in NRC Report No. 99900764/87-Ol), the NRC determined that SSM 

corrective actions were adequate and had been effectively implemented. 

enclosed report documents closure of those nonconformances. 

The 

Please respond to this letter following the instructions specified in the 

enclosed notice. 

any additional actions you plan in order to prevent recurrence. Please send 

us written response within 30 days from the date of this letter. 

Your response should document the specific actions taken and 

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

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

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

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

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

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

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

Sincerely, 

Robert M. Gallo, Chief 

Special Inspection Branch 

Division of Technical Support 

Off i ce of Nucl ear Reactor Regul at i on 

Docket No. 99900764 

Enclosures: 

1. Notice of Nonconformance 

2. Report No. 99900764/94-01 

32 
NOTICE OF NONCONFORMANCE 

Sandvi k Speci a1 Metal s Corporation 

Kennewick, Washington 

Docket No. 99900764 

Report No. 94-01 

On the basis of the results of the U.S. Nuclear Regulatory Commission (NRC) 

inspection of Sandvik Special Metals Corporation (SSM), Kennewick, Washington, 

conducted on December 5 through 7, 1994, it appears that certain of your 

activities were not performed in accordance with NRC requirements. 

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

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

Part 50), requires, in part, that activities affecting quality shall be 

prescribed by documented instructions and shall be accomplished in accordance 

with these instructions. 

Section QAM-2-5, "Instructions, Procedures, and Drawings," of the SSM Quality 

Assurance Manual (QAM), Revision 0, dated December 1, 1993, required, in part, 

that activities and operations affecting quality of tubing shall be controlled 

through the use of formally approved instructions and procedures. 

Contrary to the above, SSM did not control the identification of tube 

specimens that had failed the burst test (used to characterize the biaxial 

mechanical properties of zircaloy tubes) through the use of formally approved 

instructions or procedures, such that SSM could not perform further analysis 

of the failed tube specimens necessary to assign root cause for the reduced 

ductility of the failed specimens. Specifically, Laboratory Procedure 

1300.21, "Burst Testing Zirconium Base Alloy Tubing (Closed and Open End 

Testing)," Revision 4, dated August 21, 1985, did not prescribe documented 

instructions to accomplish the identification of failed tube specimens where 

further analysis of the specimens may affect quality (94-01-01). 

Section 3.5.1) 

(Report 

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

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

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

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

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

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

the following: (a) a description of steps that have been or will be taken to 

correct this item; (b) a description of steps that have been or will be taken 

to prevent recurrence; and (c) the dates your corrective actions and 

preventive measures were or will be completed. 

Dated at Rockvi 11 e, Mary1 and 

this 31st day of January, 1995 

33 

Enclosure 1 
REPORT NO.: 

ORGANIZATION: 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY: 

INSPECTION DATES: 

LEAD INSPECTOR: 

OTHER INSPECTORS: 

REVIEWED BY: 

APPROVED BY: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF TECHNICAL SUPPORT 

99900764/94-01 

Sandvik Special Metals Corporation (SSM) 

P.O. Box 6027 

Kennewick, Washington 99336-0027 

Kaydell C. Bowles 

Manager, Qual i ty Assurance 

SSM provides zirconium alloy fuel clad tubing used in 

nuclear fuel assemblies supplied to the nuclear power 

industry. 

December 5 through 7, 1994 

Steven M. Matthews 

Vendor Inspection Section (VIS) 

Special Inspection Branch (TSIB) 

David H. Brewer, VIS/TSIB 

Carl J. Czajkowski, Brookhaven National Laboratory 

GregoryK.&wal ina, Chief, VIS/TSIB 

Robert M. Gallo, Chief, TSIB Dat'e 

34 

Enclosure 2 
1 SUMMARY OF INSPECTION FINDINGS: 

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

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

programs related to the manufacture of zirconium alloy nuclear fuel clad 

tubing. 

SSM zirconium alloy fuel clad tubing supplied to the U.S. nuclear industry in 

fuel assemblies would perform their safety function. 

The inspection was conducted to provide a basis for confidence that 

The inspection basis consisted of the following: 

e 

e 

e 

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

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

of Federal Requlations (10 CFR Part 50) 

Part 21, "Notification of Failure to Comply or Existence of a 

Defect," of 10 CFR 

SSM Quality Assurance Manual (QAM), Revision 1, dated April 7, 

1994, which described a quality system that, according to SSM, was 

intended to meet the requirements of Appendix B to 10 CFR Part 50 

1.1 Violations 

No violations were identified during this inspection. 

1.2 Nonconformances 

1.2.1 Nonconformance 94-01-01 

This nonconformance, described in Section 3.5.1 of this report, identifies an 

instance where the documented instructions for certain activities affecting 

quality and the performance of those activities failed to comply with the 

requirements of Criterion V of Appendix B to 10 CFR Part 50 and Section 

QAM-2-5 of the SSM QAM. 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

During an inspection of SSM conducted on March 16 through 19, 1987, the NRC 

inspection team determined that certain SSM activities were not conducted in 

accordance with NRC requirements contractually imposed on SSM by purchase 

orders from nuclear fuel manufacturers. The fol lowing nonconformances, issued 

by the staff on April 27, 1987, and the associated SSM corrective actions, 

were evaluated by the team during this inspection. Based on the evaluation of 

SSM corrective actions, the team determined that each of the nonconformances 

was closed. 

-2- 

35 
2.1 Nonconformance 87-01-01 (CLOSED) 

Contrary to Criterion V, "Instructions, Procedures, and Drawings," of 

Appendix B to 10 CFR Part 50 and Sections 3.4.1 and 3.5.2 of QA-GA-5, 

Revision 14, dated February 7, 1986, the inside diameter (ID) of 19 tube 

hollows from ingot number 2241904 were inspected with the wrong go-no-go 

gauge. 

During its evaluation of the SSM tube reduction process, the team found that, 

for the inspections observed, SSM had appropriately implemented its inspection 

requirements with the correct equipment. Therefore, the team concluded that 

SSM corrective actions taken appeared adequate to ensure that the appropriate 

inspection equipment was used to determine the acceptability of tube hollows. 

2.2 Nonconformance 87-01-02 (CLOSED) 

Contrary to Criterion V, "Instructions, Procedures, and Drawings," of 

Appendix B to 10 CFR Part 50 and Section 2.1.1.8 of 2121, "Alkaline Cleaning," 

Revision 10, dated June 3, 1985, the west end of Tank No. 3 did not exhibit 

the required identification concerning the filling of the tank with fresh 

cleaning sol uti on. 

During its evaluation of the SSM tube reduction process, the team found that, 

for the process observed, SSM had appropriately identified and labeled all 

production process parameters. Therefore, the team concluded that SSM 

corrective actions taken appeared adequate to ensure that the appropriate 

identification for filling the tank with fresh cleaning solution. 

3 INSPECTION FINDINGS AND OTHER COMMENTS 

3.1 Backsround 

According to SSM, the company was founded in 1966 as a result of the Atomic 

Energy Commission diversification program for the Hanford Atomic Products 

Facility in Richland, Washington. 

AB (formerly Sandvik Jernverks) of Sweden. 

inspection, zircaloy, or Zr2 and Zr4) fuel clad tubing for the nuclear 

industry was first produced by SSM in 1968. 

SSM is a wholly-owned subsidiary of Sandvik 

Zirconium alloy (for this 

SSM supplies nuclear fuel manufacturers *zircaloy fuel clad tubing used in fuel 

assemblies for commercial power reactors in the U.S., Japan, Korea, and 

Europe. SSM's domestic clients for fuel clad tubing include ABB Combustion 

Engineering Nuclear Operations (ABB CENO) for its pressurized-water reactor 

(PWR) fuel designs, Siemens Power Corporation - Nuclear Division (SPC-ND) for 

its boi 1 ing-water reactor (BWR) fuel designs. Zi rcal oy sleeve materi a1 was 

supplied to Babcock and Wilcox Company (B&W). In addition to fuel clad 

tubing, SSM also supplied the nuclear power industry zircoloy material for the 

following fuel assembly or fuel rod applications: 

-3- 

36 
instrument tubes 

guide tubes 

water rods 0 plenum spacers 

3.2 Entrance and Exit Meetinqs 

During the entrance meeting in Kennewick, Washington, on December 5, 1994, the 

team met with members of SSM management and staff, discussed the scope of the 

inspection, and established contact persons for the team within the management 

and staff of the SSM organization. 

The team conducted a performance-based inspection of the SSM organizations 

through technically directed observations and evaluations of processes, 

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

procedures, and representative records, (b) conducted interviews, (c) held 

discussions, (d) listened to presentations, and (e) made various observations. 

Inspection participants and contacts are listed in the Appendix. 

During its exit meeting on December 7, 1994, with SSM management and staff, 

the team summarized the inspection findings. 

3.3 Tubinq Production 

3.3.1 Tube Reduction 

To manufacture fuel clad tubing, SSM received from approved zircaloy suppliers 

(e.g., Teledyne Wah Chang (TWC), Albany, Oregon) either (a) extruded tube 

shells, 3.2 inch outside diameter (OD) x 0.790 inch wall thickness, or (b) 

tube reduced extrusions (TREXs) which are extruded tube shells that have been 

pilgered, or reduced, once by the zircaloy supplier to a 2.5 or 2.0 inch OD. 

The tubing is produced by reducing the tube shells or TREXs through the basic 

reduction steps to form a final tube hollow. The final tube hollow is then 

reduced to meet the final dimensional requirements for the fuel clad tubing. 

3.3.1.1 Basic Reduction Steps 

The basic reduction steps used by SSM, as observed by the team, to produce 

fuel clad tubing from either extruded tube shells or TREXs were as described 

bclow. 

thickness for the final tube hollow were achieved. 

(1) 

(2) 

These steps would normally be repeated until the required OD and wall 

Vacuum annealing - tube shells and TREXs were annealed at 1300 - 1400°F 

(704 - 760°C) for at least 1-1/2 hours. 

Pilgering - the process of cold reduction of the extruded tube shells or 

TREXs to produce tube hollows. 

several pilgering passes, SSM produces fuel clad tubing to the specified 

dimensions. 

"cold pilgering," "tube reducing" or "rocking." This process relies on 

tube elongation by rolling back and forth (hence the term "rocking") 

By working the tube hollows through 

The pilgering process was referred to interchangeably as 

-4- 
between two grooved dies. During this process, the tube is rotated and 

advanced by small amounts and pulled over a tapered stationary mandrel. 

Both the tube diameter and wall thickness are reduced by this process. 

(3) Alkaline cleaning - where the zircaloy tube hollow was cleaned. 

(4) Vacuum annealing - the tube hollow annealing process heats the tube 

hollow to a specified temperature to promote both recrystallization and 

softening. 

(5) Inspection and Conditioning - an inspection of the surfaces of the tube 

hollow (both inner and outer surfaces) for small cracks or tears. The 

conditioning process removed these discontinuities to prevent 

propagation during subsequent pilgering passes, or reductions. 

(6) 

3.3.1.2 Final Tube Hollow Reduction 

Straightening - allowed the tube hollow to flow more easily during 

subsequent pi 1 geri ng process. 

At this stage of the tube reduction process, the final tube hollow would be 

ready to be rocked (pilgered) for the last time; also known as final hollow 

reduction (FHR). The following steps, as observed by the team, would normally 

be performed during the FHR process: 

Final pilger - the last cold reduction pilgering pass performed on the 

tubing which produces the required OD and wall thickness specified for 

the final fuel clad tubing. 

Final cleaning - the zircaloy tubing grit blast cleaned on the inner 

surface, followed by inner surface etching performed in a flowing acid 

etch bath with a proprietary rinsing system. This combination reduces 

residual fluoride and produces a very smooth ID surface. SSM stated 

that the produced finish provides iodine stress corrosion cracking 

resistance superior to other finishes. 

Pre-anneal inspection - the ID surfaces of the tubing were inspected for 

flaws prior to final anneal. 

Final thermal treatment - on the basis of the intended service 

conditions of the tubing, as determined by the fuel manufacturer, the 

final thermal treatment performed by SSM can be one of the following: 

(a) a full recrystal 1 ization anneal, or (b) a stress re1 ieving 

operation. According to SSM, the majority of tubing lots were 

recrystallized to produce specified mechanical properties and ductility 

requirements. 

Belt grinding and polishing - the tubing was belt polished on the OD 

surface to remove all evidence of the old metal. At this stage, tube 

specimens were sent to the SSM laboratory for testing. 

-5- 

38 
(6) Final cleaning - as described above. After this step, the lot of tubing 

was sent to final inspection. 

3.3.2 Beta Quench 

The team observed that SSM procured material to produce fuel clad tubing from 

TWC. All of this material had been "early" beta quenched by TWC as forged 

logs. SSM stated that approximately 2% of its fuel clad tubing production 

received "late" beta quenching and all of this material was supplied to SPC-ND 

for use in BWR appl ications. 

The team evaluated the qualification reports on the development of two methods 

of "late" beta quenching process performed by SSM; one used induction heating 

of 10 kilo hertz (kHz) which heated the tubing through the wall thickness to 

the required temperature; the second used induction heating of 450 kHz which 

heated the tubing from the outside circumference partially through the wall. 

SSM personnel stated that somewhat more than 1% of their fuel clad tubing 

production used the first method while somewhat less than 1% was produced by 

the second method. 

The "late" beta quench was performed on the tubing immediately prior to the 

last two pilgering passes. 

two-color infrared optical pyrometer. This type of pyrometer takes 

information from two different portions of the infrared spectrum and thereby 

factors out the emissivity variations that have made single color infrared 

pyrometers inaccurate unless emissivity was carefully measured and controlled. 

The temperature of the tubing was controlled by a 

As the tubing passed through induction heating, the system that provided the 

through-wall beta quench maintained the temperature in the beta range for 10 

seconds while the partial-wall beta quench maintained the beta range 

temperature for 5 seconds. 

during both processes; nitrogen gas was used as the coolant for the throughwall beta quench process and water for the partial-wall process. 

Coolant was passed through the ID of the tubing 

3.4 Final Inspection 

The team observed the following final inspection operations: 

(1) OD surface inspection - performed in accordance with SSM procedure ETA 

641, "Post Grind OD Surface Inspection," Revision 0, dated August 30, 

1991, the automated electro-optical evaluation system responded to any 

instantaneous reflectivity changes occurring on tubing OD surfaces 

caused by pits, rocked-in metal, and grinder fatspots. 

(2) Dimensional inspection - the OD and wall thickness dimensions of the 

tubing were verified. 

-6- 

39 
(3) 

(8) 

3.5 

Ultrasonic examination - a tube standard was utilized with four notches 

(on the ID and OD surfaces, oriented in both the axial and 

circumferential directinis) in order to calibrate the ultrasonic 

equipment. 

discontinuities in the tubing. 

Tube Cutting - the tubing was cut to the length specified by the PO. 

Straightness verification - the tubing straightness was verified. 

Final Visual inspection - final visual inspection was performed. 

Tube marking - final identification marking for each unique lot number 

was electro etched near one end of each tube. 

Boxing for shipment - various spacers were inserted to prevent tube-totube fretting of the tubes in transit. 

Testing 

Four transducers were utilized to detect structural 

During its review of the testing lab practices, the team observed the 

following testing operations: (a) tensile and elevated tc.n:ile testing and 

(b) burst testing and ci rcumferent i a1 el ongat i on measurement. A random 

sampling of calibration stickers was performed on the following equipment: 

tensile tester, micrometer, digital calipers, and balance; all were within 

cal i brat i on. 

3.5.1 Burst Test 

The burst test (used to characterize the biaxial mechanical properties of 

zircaloy tubes) required that a tube sample be pressurized to failure. 

begin the test, the wall thickness of each end of the tube sample was measured 

at 6 places, 60" apart, and all measurements were recorded. A rod was 

inserted into the tube to prevent bending of the specimen during deformation. 

The tube sample was installec' on a test fixture and gradually hydraulically 

pressurized to failure (rupture). After failure, a piece of brass shim stock 

was wrapped around the tube sample at the rupture location (from one fracture 

surface circumferentially to the other fracture surface), in accordance with 

ASTM B353, "Standard Specification for Wrought Zirconium and Zirconium Alloy 

Seam1 ess and We1 ded Tubes for Nucl ear Service, I' and Annex A. 1, "Recommended 

Close-End Burst Testing Procedure for Zirconium Alloy Tubing." 

was creased at each fracture surface and the length of the shim stock between 

the creases was measured using calibrated dial calipers and the measurements 

were recorded. 

To 

The shim stock 

These measurements were then entered into a computer program that calculated 

the ultimate hoop stress, total circumferential elongation (TCE %), and 

eccentricity. The burst test practice was prescribed in SSM Laboratory 

Procedure 1300.21, "Burst Testing Zirconium Base Alloy Tubing (Closed and Open 

End Testing)," Revision 4, dated August 21, 1985. The burst test is performed 

on a sample of two specimens per tubing lot. The burst test characterized the 

biaxial mechanical properties of the Zr4 fuel clad tubing. 

-7- 

40 
The team evaluated the burst test practices for tubing lot DHS12, supplied to 

ABB CENO. The initial burst test TCE for this tubing lot failed on the basis 

that, in part, of the two specimens tested, one failed to meet TCE minimum 

requirements (TCE 2 12%), as specified in ABB CENO Specification No. 00000-PD301, "Specification for Zircaloy 4 Fuel Rod Cladding Tubes," Revision 4, dated 

July 17, 1991. The burst hoop strength was acceptable. SSM performed a 

retest of four tube sample specimens from the lot (twice the original number 

tested). 

requirements, and, therefore, the entire tube lot failed the TCE specification 

requirements. 

Of the retest tube sample specimens, one specimen failed the TCE 

According to the SSM request for approval or review 9491063-002, dated 

June 15, 1994, before reworking lot DHS12R (R standing for reworked), the 

tubing contained surface imperfections on the OD of the Zr4 fuel clad tubing 

that had caused the reduced ductility during burst testing. 

imperfections as shallow axial OD microcracks attributed to parallel platelet 

microstructure and pilgering conditions. In its approval of the SSM rework 

procedure, ABB CEMO stated that SSM woulci provide ABB CENO the root cause of 

the defects and the steps taken by SSM to avoid these defects in the future. 

SSM described the 

ABB CENO also requested (a) photographs of the visual appearance of the 

surface defects and (b) a fracture surface examination of low and high 

ductility burst test specimens to aid in characterization of the defects prior 

to its approval of the rework procedure. 

After receiving approval from ABB CENO to rework the tubing lot, SSM reworked 

the tubing to remove the surface imperfections and additional tube specimens 

were tested. The rework consisted of belt-polishing the tube lot to remove 

approximately 0.0006 inch of the OD surface, considered by SSM to be 

sufficient to remove the discontinuities which caused the tube lot's failure 

to meet the ICE specification requirements. A31 tube samples of lot DHS12R 

tested met the TCE specification requirements with values that ranged from 

15.1 to 62.6% TCE. 

Note: 

From its previous inspection of ABB CENO (documented in NRC 

Report No. 99900002/94-01; 99900102/94-Ol) , the team noted 

that as of October 7, 1994, (a) the SSM root cause analysis 

requested by ABB CENO had not been performed, (b) ABB CENO 

had approved the lot of fuel clad tubing for use, and 

(c) SSM shipped the lot of tubing to AB6 CENO without 

determining the cause of the microcrack condition and 

submitting to ABB CENO a root cause analysis report. The 

team had identified the ABB CENO evaluation of the failed 

burst test specimens as an open item and requested that NRC 

be notified by ABB CENO when it completed its problem 

analysis of the failed burst test specimens, identified 

possible causes, evaluated the possible causes, confirmed 

the true cause, and identified corrective actions to be 

taken. On October 20, 1994, ABB CENO gave the team a root 

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

final report of the SSM analysis. 

-8- 

41 
However, on December 7, 1994, the team was advised by SSM that its evaluation 

of the failed tube specimens was not complete and that it would be submitting 

to ABB CENO a final problem analysis with assigned root cause. 

confusion, therefore, appeared to exist regarding the status of the analysis 

report reviewed by the team on October 20, 1994. 

Some 

The team was told by SSM that the identity of the failed test specimens was 

lost after the testing process and, therefore, the failed tube specimens were 

not available for further analysis. The tube specimens were originally 

identified with "indelible" ink. However, the oil used in the burst test 

process dissolved the ink and washed away the tube specimen identification. 

Because the test operator only performs the test and records the results 

without knowing whether tube specimens met the TCE requirements, after the 

test was completed and the test results recorded, all tube specimens were 

discarded into a bucket with no apparent method of retrieval (because 

identification on each tube specimen was inadvertently washed off) if 

required. 

Without the failed tube specimens to evaluate, SSM was not able to 

(a) photograph the surface defects and perform a visual examination or 

(b) examine the fracture surfaces of the failed tube specimens or the low and 

high ductility tube specimens. Because the SSM test lab practices for 

conducting the burst test permitted lost identification of the tube specimens, 

the SSM final problem analysis was not able to evaluate the failed tube 

specimens. 

The team concluded that the assigned root cause of the failed tube specimens 

was not conclusive. The SSM failure to control the identification of failed 

tube specimens in a manner that ensured the ability to perform further 

analysis and assign true root cause for failed tube specimens was an activity 

that affected quality and the adequacy of the corrective actions taken to 

preclude reduced ductility in other tube specimens. 

Laboratory Procedure 1300.21 and determined that the procedure a1 so failed to 

address the continued identification of tube specimens necessary to provide 

for post-testing failure analysis. By not establishing measures in Laboratory 

Procedure 1300.21 to control the identification of failed tube specimens that 

may affect quality and by not utilizing test practices to identify and retain 

failed tube specimens, where further analysis may affect quality, SSM failed 

to comply with Section QAM-2-5, "Instructions, Procedures, and Drawings," of 

the SSM QAM that required, in part, that activities and operations affecting 

quality of tubing shall be controlled through the use of approved instructions 

and procedures and was, therefore, a nonconformance. As a result, 

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

The team reviewed SSM 

3.6 Corrective Actions 

During its inspection of ABB CENO (documented in NRC Report No. 

99900002/94-01; 99900102/94-Ol), the team reviewed two concerns that ABB CENO 

had developed regarding surface indications on fuel clad tubing supplied by 

SSM. During this inspection of SSM, the team evaluated the SSM problem 

analysis of the issues identified on ABB CENO Corrective Action Requests 

(CARS) for the concerns described below. 

-9- 

42 
3.6.1 CAR V-20-94 

Discussions with ABB CENO indicated that since July 1994, ABB CENO h:d 

experienced problems with SSM tubing involving blue, pink, black, grhy, and 

brown stains and white deposits on various surfaces. There was evidence that 

the problems were being addressed as required by the ABB CENO QA programs, but 

the corrective action had not been successful in preventing the recurrence of 

new but similar indications of nonconforming tubing for fuel cladding. On 

November 8, 1994, ABB CENO issued SSM CAR V-20-94 that identified cleanliness 

and surface condition problems on the following SSM lots of tubing: 

DHU51 

DHU2 1 

DHT82 

DHT14 

DHW42 

DHY62 

Although the SSM problem analysis and corrective action was not complete as of 

December 7, 1994, during this inspection the NRC team reviewed the SSM 

evaluation atld corrective actions taken for the "whitish" discolorltions 

observed on the tube OD surface. The results of the SSM evaluation were 

documented in SSM Technical Report No. 94-009, "Evaluation of Paper Divider 

Trays," dated September 26, 1994. For its evaluation of the "whitish" 

deposits, SSM utilized both scanning electron microscopy (SEM) and energy 

dispersive spectroscopy (EDS). 

EDS is an analytical technique, capable of performing elemental analysis of 

microvolumes, typically on the order of a few cubic microns in bulk samples 

and considerably less in thinner sections. 

sample is accomplished by crystal spectrometers which use energy dispersive 

techniques permitting analysis by discriminating among x-ray energies. The 

feature of electron beam microanalysis that best describes this technique is 

its mass sensitivity. For example, it is often possible to detect less than 

The 

minimum detectable quantity of a given element or its detectability limit 

varies with many factors, and in most cases is less than 

grams/microvolume. 

Analysis of x-rays emitted from a 

grams of an element present in a specific microvolume of a sample. 

SSM determined that the gly;sy coating on the back of its vendor supplied 

paper divider trays (Rondo trzys), used in shipping fuel clad tubing, was 

responsible for the "whitish" deposits found on the OD surfaces of some tubes. 

SSM corrective action was to extend the paper separating the layers of tubing 

beyond the tube ends. 

specify a type of paper backing which will not produce stains. 

SSM also altered their purchase specification to 

For the various color stains that were observed by ABB CENO, the SSM 

preliminary evaluation of the problem indicated that, in part, the 

illumination in the packaging and final visual inspection area would be 

increased. 

- 10 - 

43 
Additionally, the ink identification marking practices and the cleaning 

practices would be revised, as they are prescribed respectively in SSM 

procedure QA-GA-21, "Additional Lot Identification by Color Coding of Tubes," 

Revision 5, dated June 8, 1994, and Process Specification No. 2449, "Belt 

Grinding Zircaloy - Niederberger Grinders," Revision 17, dated April 17, 1994. 

On the basis of its evaluation of the SSM problem analysis and corrective 

actions, the team concluded that the SSM evaluation appeared to be thorough, 

competent, and conclusive as to the root cause of the discoloration. 

3.6.2 CAR V-21-94 

During the inspection of ABB CENO referenced in Section 3.5.1 above, the team 

observed apparent indications of pitting on the OD surface of two Zr4 fuel 

clad tubes from SSM lot DHU51. As of November 2, 1994, ABB CENO had neither 

completed its problem analysis of the surface indications to identify possible 

causes, evaluate the possible causes, and confirm the true cause, nor had it 

identified corrective actions to be taken. The team identified this to ABB 

CENO as an open item and requested that NRC be notified by ABB CENO when the 

corrective action had been completed. 

On November 8, 1994, ABB CENO issued SSM CAR V-21-94 that identified two tubes 

from lot DHU51 as exhibiting a spiraljhelix condition on the OD surface of the 

tube; ABB CENO added that the marks were indicative of pitting or areas that 

did not clean up at belt grinding. 

Although the SSM problem analysis and corrective action was not complete as of 

December 7, 1994, during this inspection the team reviewed the SSM evaluation 

and corrective actions taken for the spiral pitting observed on the tube OD 

surface. 

magnification and the Manager, Quality Assurance put forth an analysis that 

the pitting was actually fretting that occurred while the tubes were in 

transit from SSM to ABB CENO. 

The team examined the spiral pitted tube sections with a 40x 

Fretting occurs when two metal surfaces move against one another with an 

oscillating motion, and forms abrasive particles. These metallic particles 

react with the air, are compacted, and eventually are impressed into the metal 

surfaces. 

The team also witnessed the spiral pitted tube section processed through the 

automated visual inspection process to determine whether the automated 

electro-optical evaluation system would detect the spiral pitting. The test 

proved to the team's satisfaction that the automatic visual inspection 

process, performed in accordance with SSM procedure ETA 641, "Post Grind OD 

Surface Inspection," Revision 0, dated August 30, 1991, would detect the 

spiral pitted surface condition and reject the tubing. The proposed SSM 

corrective actions included revising Quality Assurance Procedure NDT-V-57, 

"Final Visual and Packaging," Revision 15, dated December 8, 1993, to enhance 

the final packaging inspections and the possibility of adding packing material 

in certain instances. 

- 11 - 

$4 
On the basis of its evaluation of the SSM problem analysis and corrective 

actions, the team concluded that the SSM evaluation appeared to be thorough, 

competent, and conclusive as to the root cause of the fretting. 

- 12 - 

45 
APPEND I X 

PERSONS CONTACTED 

The NRC staff participating in the inspection of SSM; SSM’s personnel 

contacted during the inspection; and the personnel attending the entrance and 

exit meetings are listed below. A bullet (0) indicates that person attended 

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

meeting. 

Sandvi k Sped a1 Metal s Corporati on: 

t Banko, W. W. 0 t Bowles, K.C. 

Bradley, Dr. E.R. 0 t Brewer, T.M. 

Erickson, P. 

t Galbraith, K.P. 

Hollibaugh, C.L. 

Hoxie, D. 

Johnson, R.T. 0 Knapp, G.R. 

King, C. 

Luebke, J.T. 0 t Martenson, Dr. A.J. 

Vice President, Operations & Plant Manager 

Manager, Quality Assurance 

Senior Development Metal 1 urgi st 

Quality Assurance Engineer 

Technician (Burst Test) 

President & CEO 

Internal Auditor 

Technician (Tensile Test) 

P1 ant Superintendent 

Coordinator, Total Qual i ty 

Manager, QA and Lab Services 

Manager, Facilities Engineering 

Vice President, Research & Engineering 

U.S. Nuclear Regulatory Commission: 

0 t Brewer, D.H. Metallurgical Engineer, VIS/TSIB 0 t Czaj kowski , C. J. Metal 1 urgical Engineer, Brookhaven 

0 t Matthews, S.M. Quality Assurance Specialist, VIS/TSIB 

National Laboratory 

- A-1 - 

46 
' *****+ * 

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-0001 

January 31, 1995 

Mr. Albert E. Riesen, President 

Teledyne Wah Chang 

P.O. Box 460 

Albany, Oregon 97321-0460 

SUBJECT: NRC INSPECTION NO. 99901229/94-01 

Dear Mr. Riesen: 

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

(NRC) inspection of Teledyne Wah Chang (TWC), Albany, Oregon, conducted 

November 28 through December 2, 1994. The NRC inspection team, led by 

Steven M. Matthews and comprising other inspectors named in the report, 

conducted a performance-based evaluation of TWC management, staff, and quality 

programs and the implementation of those programs related to the fabrication 

of zirconium alloys in various product forms. The inspection was conducted to 

provide a basis for confidence that TWC zirconium alloy products supplied to 

the U.S. nuclear industry in fuel assemblies would perform their safety 

funct i on. 

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

representative records, (b) conducted interviews, (c) held discussions, 

(d) listened to presentations, and (e) made various observations. On the 

basis of this inspection, the NRC team determined that the implementation of 

the TWC quality assurance program, documented in the TWC Quality Manual, 

Revision 1, dated March 8, 1994, either met or exceeded the requirements of 

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

B to 10 CFR Part 50). 

discussion of the areas examined and the NRC team observations. 

The enclosed inspection report contains a detailed 

The NRC team identified the following strengths in TWC's production 

operations: (a) the products were custom fabricated in accordance with the 

processes and specifications established by the customer and all aspects of 

the production process were documented, from order entry through the execution 

of each process step by operators and quality control personnel; 

(b) training and requalification of personnel was in order and process 

operating personnel were considered competent, in most instances having long 

experience with TWC; and (c) technical problem analysis and modifications to 

equipment and processes to improve quality were performed by competent 

engineering personnel. 

I 47 
A. Riesen -2- 

For the nonconformances cited in the NRC report of its previous inspection of 

TWC (NRC Report No. 99901229/91-Ol), the NRC determined that TWC corrective 

actions were adequate and had been effectively implemented. 

report documents closure of those nonconformances. 

The enclosed 

No response to this letter or its enclosure is required. 

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

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

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

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

Sincerely, 

Special Inspection Branch 

Division of Technical Support 

Off i ce of Nuc I ear Reactor Regul at i on 

Docket No.: 99901229 

Enclosure: Report No. 99901229J94-01 
REPORT NO.: 

ORGANIZATION: 

ORGANIZATIONAL 

CONTACT : 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF TECHNICAL SUPPORT 

NUCLEAR INDUSTRY 

ACTIVITY : 

INSPECTION DATES: 

LEAD INSPECTOR: 

OTHER INSPECTORS: 

REVIEWED BY: 

APPROVED BY: 

99901229/94-01 

Teledyne Wah Chang (TWC) 

1600 North East Old Salem Road 

P.O. Box 460 

Albany, Oregon 97321-0460 

Robert C. Bartcher 

Director, Quality Assurance 

TWC produces zirconium alloys for use in fuel 

assemblies by the nuclear power industry. 

November 28 through December 2, 1994 

Steven M. Matthews Date 

Vendor Inspection Section (VIS) 

Special Inspection Branch (TSIB) 

David H. Brewer, VIS/TSIB 

John R. Weeks, Brookhaven National Laboratory 

Ggegorf g% Cwalina, Chief, VIS/TSIB Date 

I Robert M. Gallo, Chief, TSIB Date 

49 

Enclosure 
1 SUMMARY OF INSPECTION FINDINGS 

During this inspection, the NRC inspection team evaluated TWC management, 

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

to the fabrication of zirconium alloys in various product forms. 

inspection was conducted to provide a basis for confidence that TWC zirconium 

alloy products supplied to the U.S. nuclear industry in fuel assemblies would 

perform their safety functions. 

The 

The inspection basis consisted of the following: 

TWC Quality Manual (QM), Revision 1, dated March 8, 1994 

Appendix By "Quality Assurance Criteria for Nuclear Power Plants and 

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

Federal Requlations (10 CFR Part 50) 

Part 21, "Notification of Failure to Comply or Existence of a 

Defect," of 10 CFR 

No violations or nonconformances were identified during this inspection. 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

During an inspection of TWC conducted on October 7 through 10, 1991, the NRC 

inspection team determined that certain TWC activities were not conducted in 

accordance with NRC requirements contractually imposed on TWC by purchase 

orders (POs) from nucl ear fuel manufacturers. The fol1 owing nonconformances, 

issued by the staff on December 4, 1991, and the associated TWC corrective 

action, were evaluated by the NRC team during this inspection. Based on the 

evaluation of TWC corrective actions, the NRC team determine that each of the 

nonconformances was closed. 

2.1 Nonconformance 91-01-01 (CLOSED) 

Contrary to Criterion V, "Instructions, Procedures, and Drawings," of 

Appendix B to 10 CFR Part 50, the TWC Quality Control Manual (QCM), 

Revision 1, dated February 15, 1991, had not stated that activities affecting 

quality were to be prescribed by documented procedures and performed in 

accordance with those procedures. 

To evaluate the TWC corrective action that addressed this nonconformance, the 

NRC team reviewed the TWC implementation of (a) Section F, "Instructions, 

Procedures, and Drawings, I' of the TWC Quality Manual, Revision 1, dated 

March 8, 1994; (b) Quality Control Instructions (QCI), document QCI-F-1, "Work 

Instructions," Revision 6, dated June 6, 1994; and (c) QCI-F-4, "Process Flow 

Outlines," Revision 1, dated August 29, 1994. 

TWC corrective action was adequate to ensure that activity affecting quality 

was documented and performed in accordance with those procedures. 

The NRC team determined that 

-2- 

!50 
2.2 Nonconformance 91-01-02 (CLOSED) 

Contrary to Criterion XV, "Nonconforming Materials, Parts, or Components," of 

Appendix B to 10 CFR Part 50, and Section P.3 of the TWC QCM, dated 

October 23, 1989, TWC had not issued a Product Condition Information Request 

(PCIR) to document deviations in samples from Zircaloy-2 heat number 22881 

following nodular corrosion testing. 

The TWC QCM in effect during the October 1991 inspection had been replaced. 

To evaluate the TWC corrective action that addressed this nonconformance, the 

NRC team reviewed the TWC implementation of QCI-P-2, "Production Condition 

Information Request," Revision 9, dated November 24, 1993. The NRC team 

determined that TWC corrective action was adequate to ensure that a PCIR was 

issued for a material deviation. 

3 INSPECTION FINDINGS AND OTHER COMMENTS 

3.1 Backqround 

Wah Chang (a Chinese phrase meaning "great development") was founded in New 

York in 1916 for international ore trading, engineering and tungsten 

extraction. In 1956, to operate a government zirconium plant, Wah Chang 

expanded to Albany, Oregon. Wah Chang was purchased in 1967 by Teledyne, Inc. 

Zircaloy-2 (Zr2) and Zircaloy-4 (Zr4) (collectively, zircaloy), used in U.S. 

nuclear fuel assemblies, were produced by TWC in strip for channel sections 

and grid spacers, rod for fuel rod end caps, and extruded tube hollows for 

fuel clad tubing. 

products, TWC custom fabricates them to meet customer negotiated requirements. 

To accommodate the various needs among customers for these 

3.2 Entrance and Exit Meetinqs 

During the entrance meeting in Albany, Oregon, on November 28, 1994, the NRC 

team met with members of TWC management and staff, discussed the scope of the 

inspection, and established contact persons for the NRC team within the 

management and staff of the TWC organization. 

During the inspection, the NRC team condycted a performance-based inspection 

of TWC organizations through technically directed observations and evaluations 

of processes, activities, and documentation. The NRC team (a) examined 

technical documentation, procedures, and representative records, (b) conducted 

interviews, (c) held discussions, (d) listened to presentations, and (e) made 

various observations. 

Inspection participants and contacts are listed in the Appendix. 

During the exit meeting on December 2, 1994, the NRC team summarized the 

inspection findings with TWC management and staff. 

-3- 

51 
3.3 Order Entry 

During this inspection the NRC team evaluated the following documentation: 

PO 100 751, dated November 17, 1993, from Sandvik Special Metals 

Corporation (SSM) for Zr4, Grade R60804, 3.20 inch outside diameter (OD) 

x 1.62 inch inside diameter (ID) x 0.790 inch wall extrusions (tube 

shells), produced in accordance with SSM's specification SSM 1021, 

Revision 10, and TWC Quality Control Plan and Process Flow Outline 

(PFO), PFO-271, "Zircaloy 2 and 4 Extruded Tube Shells (3.2" OD x 

0.79" W)," Revision 1, dated December 3, 1993. 

PO 9392494-D5602, dated September 23, 1993, from ABB Combust ion 

Engineering Nuclear Operations (ABB CENO) for Zr4 centerless ground bar 

material produced in accordance with ABB CENO's specification 00000-PD304, "Zircaloy 4 Bar Material for Fuel and Burnable Poison Rod End 

Caps," Revision 3, dated July 17, 1991, and TWC PFO-251, "Zircaloy 

Extruded Plug Stock Cold Worked, Annealed, and Centerless Ground," 

Revision 1, dated October 22, 1992. 

PO R/073263, dated April 7, 1994, from Siemens Power Corporation - 

Nuclear Division (SPC-ND) for Zr4 strip, 0.17 inch thick x 9.5 inch 

wide, produced in accordance with SPC-ND's specification EMF-S35081, 

"Formable Zircaloy 4 Strip or Sheet for PWR Applications," Revision 4, 

dated November 14, 1991, and TWC PFO- 147, "Zircaloy Spacer Strip-PWR," 

Revision 6, dated October 26, 1989. 

PO 027266, dated August 8, 1994, from Carpenter Technology Corporation, 

Special Products Division, for Zr4, Grade R60804, triple melted, alpha 

annealed, 0.083-0.086 inch thick x 20.572-20.582 inch wide strip 

produced in accordance with Carpenter's specification S-70000-1, 

"Zircaloy Strip for BWR Channels," Revision 11, dated November 30, 1993, 

and TWC PFO-250, "Zircaloy Channel Strip," Revision 0, dated 

June 9, 1992. 

The NRC team evaluated these packages and determined that TWC customers 

established the purchase specifications which included sampling and testing to 

be performed and the requirement to approve all revisions to applicable PFOs. 

The NRC team also determined that TWC material certifications, for the items 

reviewed from the POs listed above, were detailed in their description of the 

following items: 

a 

a 

a 

a 

a 

a 

a 

a 

0 

a 

heat number 

ingot chemistry analysis 

ingot impurities analysis 

product chemistry analysis 

ingot ultrasonic test 

ingot hardness 

grain size 

tests, as specified by the customer 

cumulative anneal parameter range 

metallographic tests 

-4- 

52 
3.4 Zirconium Alloy Manufacturins Process 

TWC produced zirconium, niobium, hafnium, titanium, and various refractory 

metals and alloys. 

(zirconium orthosilicate) from Australia, the major supplier. 

Zirconium metal and alloys were produced using zircon sand 

The NRC team observed various stages of the TWC production process which is 

described here in some detail through completion of ingot production. 

After separating zircon from several other naturally occurring minerals, the 

zircon was mixed and ground with an appropriate amount of carbon in the form 

of coke to produce an intimate mixture having a specified particle size range. 

The zircon sand/coke mixture was reacted with chlorine at about 1200°C 

(2200°F) to produce zirconium tetrachloride and silicon tetrachloride. 

Zirconium tetrachloride and silicon tetrachloride were separated by partial 

condensation of the chlorinator off-gas. 

Zirconium tetrachloride, which contained approximately two percent hafnium 

tetrachloride, was dissolved in water and filtered to remove suspended solids 

(fine sand and carbon). 

although they are similar chemically, they are quite different in their 

nuclear properties. Hafnium absorbs large quantities of thermal neutrons 

while zirconium does not. 

columns. 

The zirconium and hafnium were separated, for 

Separation was accomplished by a series of resin 

The aqueous zirconium tetrachloride was precipitated with sulfuric acid to 

form hydrous zirconium sulfate and filtered. The filter cake was treated with 

ammonium hydroxide to form hydrous zirconium oxide and filtered and rinsed 

again to remove sulfate ions. The hydrous zirconium oxide was calcined at 

approximately 1000°C (1832°F) to produce zirconium oxide. 

Zirconium oxide and the appropriate amount of carbon in the form of coke were 

intimately blended and reacted with chlorine to produce zirconium 

tetrachloride. Aluminum, iron, and titanium and others were removed as 

impurities. 

elevated temperature to produce zirconium metal and magnesium chloride. The 

zirconium metal was heated to distill excess magnesium and produce a zirconium 

regulus, a porous sponge-1 i ke materi a1 , 

Zirconium tetrachloride was reacted with metal1 ic magnesium at 

The zirconium sponge regulus was broken into chunks suitable for crushing to 

pass a three-quarter inch screen. 

chemistry requirements, sponge from various runs was blended together. 

Blended zirconium sponge and alloying elements were combined and pressed into 

14" diameter by 5" thick compacts. 

Based on chemical analysis, computer selected combinations of sponge compacts 

and recvcled billets were electron beam welded to form electrodes. The 

y the consumable vacuum arc melting 

mpurities and improved homogeneity. TWC 

few mi nor exceptions, a1 1 zi rconi um a1 1 oys 

Fines were removed. To achieve ingot 

electrohes were me1 ted into ingots 

process. Mu1 ti pl e me1 ts decreased 

technical staff stated that, with a 

-5- 

53 
supplied to the commercial nuclear power industry were triple arc melted. 

Ingots were given a heat number by which they were identified through 

subsequent processing. 

Ingots were ultrasonically tested to determine the extent of pipe, hardness 

tested, and sampled for chemical analysis at multiple locations along the 

length to assure homogeneity and compliance with customer specifications. 

The NRC team reviewed a qualification report researched and written during the 

early period of plant history demonstrating that the triple arc melting 

process produces ingots that are uniform in chemical composition throughout 

within a few percent of the mean value. On this basis, TWC considers chemical 

samples from various locations on one side of the ingot as representative of 

the who1 e. 

Ingots meeting customer criteria were identified as suitable for assignment to 

the customer's requirements. The circumference of the ingots was machined and 

ingots were fabricated into wrought product by forging, hot rolling, cold 

rolling, or extrusion. 

sequence depends on the alloy, final product form, and customer requirements. 

The choice of working methods and heat treatment 

The NRC team observed processes in which (a) sponge was crushed, sorted, and 

blended in preparation for assembling compacts, (b) sponge compacts and 

recycled billets were electron beam welded into an electrode suitable for arc 

melting, and (c) vacuum arc melting was performed, producing an ingot. 

TWC demonstrated to the NRC team its ability to control various alloying 

elements within a relatively narrow band of the range specified by 

ASTM B 350-93, "Standard Specification for Zirconium and Zirconium Alloy 

Ingots for Nuclear Application." 

perform a range of thermal and mechanical treatments, TWC customers specify 

the combination of chemical composition, thermal, and mechanical treatments 

that will produce material properties that best suit their needs. TWC is, 

therefore, a custom fabricator, negotiating product and process details with 

each customer. 

With this capability and the ability to 

3.5 Beta Quenching 

In discussion with the NRC team the TWC technical staff stated that beta 

quenching was the single most significant metallurgical factor relating to all 

product forms in both Zr2 and Zr4. The reason for this significance was the 

affect beta quenching has on improving the corrosion resistance of the 

materi a1 s. 

3.5.1 Fundamental s 

Beta quenching is a thermal treatment in which the workpiece is heated above 

about 1000°C (1832"F), held in the beta phase temperature range, and rapidly 

quenched in water. At the temperature of about lOOO"C, the crystalline 

structure changes from hexagonal close packed to body centered cubic. This 

-6- 

54 
change in crystalline structure is referred to as the alpha to beta 

transformation. Above the transformation temperature, alloying elements 

dissolve from particles of intermetallic compounds and go into solid solution 

more rapidly. 

The effect of this thermal treatment is to more uniformly disperse alloying 

elements throughout the microstructure. TWC personnel stated that evidence of 

the effectiveness of the beta quench process had been demonstrated 

microscopically by differences observed in the size and distribution of 

intermetallic compound particles dispersed in the sol id solution matrix. 

The uniformity of the temperature to which the workpiece is heated, typically 

between 1016-1200°C (1861--2192"F), is not critical provided it is above the 

transformation temperature. 

The upper temperature limit depends on the time at temperature. If acceptable 

limits of time at temperature were exceeded, too much of the alloying elements 

would go into solid solution. This would result in insufficient sites of 

intermetallic compounds available for nucleation and growth during quenching. 

Although in the alpha phase after quenching, precipitation would occur at 

prior beta grain boundaries, resulting in areas of parallel platelet 

microstructure which TWC personnel stated would be readily observable in 

metallographic examination. Parallel platelet microstructure is not as 

ductile as solid solution with islands of intermetallic compounds. A 

microstructure containing areas of parallel platelets is subject to cracking 

during subsequent hot and cold working and is to be avoided. 

TWC stated that all zircaloy they produced received an "early" beta quench 

and, in addition, some Zr2 received a "late" beta quench performed by 

customers believing that enhanced corrosion resistance would result. 

3.5.2 Corrosion Resistance 

Research' 

how different heat treatments must be performed on Zr2 and Zr4 to obtain 

optimum corrosion resistance for each alloy in light water reactor 

appl i cat i ons . 

referenced by TWC personnel and reviewed by the NRC team, showed 

' Garzarolli, F., Stehle, H., Steinberg, E., and Weidinger, H., 

"Progress in the Knowledge of Nodular Corrosion," Zirconium in the Nuclear 

Industry: Seventh Internationa7 Symposium, ASTM STP 939, R. B. Adamson and 

L. F. P. VanSwam, Eds., American Society for Testing and Materials, 

Philadelphia, 1987, pp. 417-430. 

and Corrosion Studies for Optimized PWR and BWR Zircaloy Cladding," Zirconium 

in the Nuclear Industry: Eighth International Symposium, ASTM STP 1023, 

L. F. P. VanSwam and C. M. Euken, Eds., American Society for Testing and 

Materials, Phi lade1 phi a, 1989, pp. 202-212. 

Garzaroll i, F., Steinberg, E., and Weidinger, H. G., "Microstructure 

-7- 

55 
At the ingot stage and during subsequent open die forging, alloying elements 

segregate on a microscopic basis by growing intermetallic compound 

precipitates. Beta quenching of slabs, logs, or extrusion billets homogenizes 

the chemistry by dispersing the alloying elements into solid solution. 

chemical homogenizing resulting from beta quenching is essential to the 

enhanced corrosion resistance of Zr2 and Zr4 as described in the references. 

The 

Zircaloy-2, generally used in boiling-water reactors (BWRs), is more subject 

to nodular corrosion. 

produced by an "early" beta quench performed in the billet reduction process. 

Subsequent thermal treatments are performed in a way to control heat input 

below a specified value, as cited in the referenced research, to ensure 

optimum corrosion resistance. 

The best corrosion protection for this alloy is 

TWC stated that some of their customers that manufacture Zr2 fuel clad tubing 

perform a "late" beta quench prior to the final mechanical forming processes. 

Thermal apd mechanical processes performed on the material after the "early" 

beta quench promote additional precipitation of alloying elements from solid 

solution. The metallurgical effect of the "late" beta quench is to return the 

optimum amount of the alloying elements to solid solution. Higher 

concentrations of alloying elements in solid solution for Zr2 have 

demonstrated enhanced resistance to nodular corrosion in BWR service. 

In contrast, Zr4, generally used in pressiirized-water reactors (PWRs), is more 

subject to uniform corrosion. The best corrosion protection for this alloy is 

produced by an "early" beta quench performed in the billet reduction process. 

Subsequent thermal treatments are performed to expose the material to a 

relatively higher range of accumulated heat input than for Zr2, as cited in 

the referenced research. The metallurgical effect of subsequent mechanical 

and thermal treatments is to precipitate more intermetallic compounds and grow 

them larger, leaving less alloying elements in solid solution. Lower 

concentration of alloying elements in solid solution for Zr4 has demonstrated 

enhanced resistance to uniform corrosion in PWR service. No "late" beta 

quench is performed which would increase the alloy content of the solid 

solution. 

3.5.3 Furnaces 

TWC used several electrical resistance and induction furnaces to heat zircaloy 

for beta quenching. Electrical resistance furnaces were used to heat material 

that had been open die forged to slab (for plate, sheet, and strip products), 

and material rotary forged to log (for bar, rod, wire, and about 95% of the 

material for fuel clad tubing or other tubin? applications). 

furnaces were used to heat extrusion billets only; about 5% of the material 

for fuel clad tubing or other tubing applications. 

Induction 

An extrusion billet is very much like a log that has been reduced to a 3 

round cross section by various forging operations but with a hole trepanned 

down the centerline. 

-8- 

56 
The NRC team noted that a variety of induction furnaces were available for 

installation on the beta quench tower. 

size depending on the size of the billet it was to heat. 

ceramic sleeve mounted between the induction coil and the workpiece space. 

The sleeve prevented the workpiece from contacting the coil and maintained it 

in the approximate center of the coil. 

Each induction furnace was a different 

Each furnace had a 

An induction furnace heats a workpiece more rapidly than an electrical 

resistance furnace. 

more desirable for heating Zr2 for nodular corrosion resistance 

considerations. Induction heating produces heat at or near the surface of the 

workpiece while power is flowing in the induction coil. 

off to the coil, heat induced when power was on conducts to the interior of 

the workpiece. TWC personnel selected furnace operating parameters to produce 

the desired heating characteristics. 

documented in the TWC profile report evaluated by the NRC team. 

For this reason the induction furnace may be considered 

When power is turned 

Furnace operating parameters were 

3.5.4 Temperature Measurement 

The NRC team observed a variety 

methods used by TWC; thermocoup 

temperature indicatinq crayons. 

and Control 

of temperature measurinq and i ndi cat ng 

es, optical pyrometers,-Temp lsticks and 

However, only thermocouples with ca ibration 

traceable to the NatiGnal institute for Standards Technology (NIST) were 

observed being used to control beta quenching processes in both the electrical 

resistance and induction furnaces, 

The NRC team observed a single color optical pyrometer in the area where 

induction heating was occurring but it was not in use. TWC personnel stated 

that it was accurate only under tightly controlled circumstances. 

emissivity of a workpiece surface influenced the temperature reading produced 

and since workpiece emissivity varies considerably depending on surface finish 

and oxidation thickness, TWC used it only where temperature measurement 

accuracy was not considered critical. 

The 

Calibration of the thermocouple temperature measuring and controlling 

equipment had been performed periodically as required by MIL-STD-45662A, 

"Calibration Systems Requirements," dated August 1, 1988. A similar method 

was used to calibrate the thermocouples and controller used in the induction 

furnaces. The NRC team observed that all thermocouple measuring and 

control1 ing equipment used in both electrical resistance and induction 

furnaces were current in calibration. 

3.5.4.1 Electrical Resistance Furnace Temperature Surveys 

Furnace temperature surveys are performed to determine the working zone or 

volume within the furnace that maintains the required temperature control. 

Furnace surveys are performed by placing a number of thermocouples throughout 

the furnace to collect temperature information as the furnace operates. 

temperature information is evaluated to determine the working zone of the 

furnace within which acceptable temperature limits are maintained. 

heat treatment activity is required to be performed within the working zone. 

The 

Subsequent 

-9- 

57 
The NRC team reviewed the furnace surveys for electrical resistance furnaces 

used for beta quenching. Furnace surveys had been performed on the furnaces 

periodically as required by MIL-STD-45662A, "Calibration Systems 

Requirements," dated August 1, 1988. In each case the NRC team determined 

that the acceptable working zone was being used and the surveys were current. 

3.5.4.2 Induction Furnace Temperature Profiles 

For an induction furnace a temperature profile performed using a profile 

billet is comparable to performing a temperature survey on an electrical 

resistance furnace. 

TWC personnel stated that a temperature profile was performed on each furnace 

before and after beta quenching each lot of material. 

consistent furnace performance during beta quenching of the lot. 

This process verified 

The NRC team reviewed documentation of a temperature profile of the Mark I1 

Induction Furnace performed September 27, 1994. The temperature profile had 

been performed using a "profile" billet of Zr2 having dimensions like the 

billets that were to be heated. The "profile" billet was instrumented with 

six thermocouples embedded; two, inch from the top of the billet, one of 

these 4 inch in from the OD and the other 4 inch out from the ID. The other 

four were similarly placed; two, 4 inch from the bottom of the billet and two 

midway along the axis. The NRC team observed that the calibration of each 

thermocoupl e was veri f i ed and documented by TWC personnel . Temperature 

profile documentation reviewed by the NRC team was within specified limits 

with the exception of one control thermocouple which was noted to be out of 

proper position. TWC personnel stated that they considered the control 

thermocouple out of position as insignificant since the profile thermocouples 

had registered within acceptable limits. 

condition and agreed that for temperature profiling purposes, it was not 

si gni f i cant. 

The NRC team reviewed the discrepant 

3.5.4.3 Electrical Resistance Temperature Control 

Just as induction furnace profiles compare and contrast to electrical 

resistance furnace surveys, so control thermocouples have similarities and 

differences between the two types of furnaces. 

use the same type of control thermocouples and their calibration is checked 

the same way, the way they are used is different. In the electrical 

resistance furnace the control thermocouple is permanently mounted at a 

location within the working zone. 

While both types of furnaces 

3.5.4.4 Induction Furnace Temperature Control 

In the induction furnace three process control thermocouples were placed in 

the ID of each billet quenched; two, % inch from each end and one midway. 

Billet soak temperature was specified as 1050 to 1200°C (1922 to 2192°F) and 

soak time as 2 to 5 minutes. 

to allow for variations in heating that can occur in the induction heating 

process. TWC personnel stated that standard practice had not produced such a 

large range of variation. 

The wide range on soak temperature was specified 

- 10 - 

58 
The temperature data from the September 27, 1994, profile reviewed by the NRC 

team showed the maximum temperature recorded by the six embedded thermocouples 

ranged from 1096 to 1167°C (2005 to 2133°F). 

throughout the billet as the induction furnace cycled on and off. 

temperature variation was considered normal by TWC. The NRC team reviewed the 

temperature variation and found it within the limits discussed in 

Section 3.5.1 of this report. 

Temperature variation occurred 

This 

3.5.5 Quench! ng Operat i ons 

According to TWC, selection of the beta quenching process used on a customer's 

product was made by that customer. Since each process produces various 

advantages and disadvantages, the selection depends on customer requirements 

for its product. 

TWC strived for a quenching rate of 20-40°C (68-104"F)/second. 

was neither specified in controlling documentation nor measured. To 

accomplish this quenching rate however, the following conditions were 

specified in a number of specifications for quenching from electrical 

resistance and induction furnaces. 

This parameter 

Water temperature at the beginning of quenching less than 37°C (99°F) 

The volume of quenching water greater than twenty times the volume of 

the metal being quenched 

Except for the pit tank used to quench slabs, recirculating 

capability in all quenching tanks 

All beta quenching operations observed by the NRC team complied with these 

requirements. All thermocouples indicating the temperature of water in the 

quench tanks were observed to be in calibration. Quenching water was taken 

from a nearby river and contained no chlorine. Water from all quench tanks 

was recirculated through a cooling tower. 

The NRC team observed beta quenching operations at various times during the 

inspection involving the following materials. 

Material from Zr2 heat number 2368614, open die forged to 4 inch thick 

slab, was electrical resistance furnace heated for Carpenter Technology 

Corporation, Special Products Division, to produce channel strip for use 

in a German nuclear reactor. The specified soaking time at the 1052°C 

(1926°F) set temperature was 20 to 60 minutes; soaking time was observed 

to be 35 minutes. 

the floor immediately in front of the furnace. A motor lift with a 

pincer arm pulled the slab from the furnace and dropped it into the 

quench tank. 

furnace door to immersion in the water) was 25 seconds; actual transfer 

time as recorded by the operator's stopwatch was 16.9 seconds. 

Quenching was performed in the pit tank located in 

The specified maximum transfer time (from opening of the 

- 11 - 

59 
(2) Material from Zr4 heat number 236830Q, extruded to 3.2 inch diameter 

bar, was electrical resistance furnace heated for SSM to produce fuel 

clad tubing. The specified maximum soaking time at the 1052°C (1926°F) 

set temperature was 60 minutes; soaking time was observed to be 40 

minutes. Quenching was performed in the recircul ating quench tank 

located approximately 12 meters (40 feet) from the furnace. A motor 

fork lift carried two logs from the car bottom furnace to the quench 

tank for immersion. The logs were retained on the fork of the lift 

during immersion. The specified maximum transfer time was 60 seconds; 

actual transfer time as recorded by the operator’s stopwatch was 39.8 

seconds. 

(3) Material from Zr2 heat number 2367464, rotary forged and trepanned to an 

extrusion billet approximately 6-7/8 inch OD X 2.4 inch ID X 18-1/4 inch 

long, was heated in the Mark I1 Induction Furnace Quenc) Tower for Kobe 

Steel to produce the outer shell of barrier coextrusion . Specified 

soaking time in the 1050 to 1200°C (1922 to 2192°F) temperature range 

was 2 to 5 minutes; soaking time was observed to be approximately 3 

minutes 15 seconds. Quenching was performed by lowering the support 

platform and sliding it from under the billet, allowing the billet to 

fall into the quench tank. 

5 seconds. The quench tank had an alignment cone made of a rod assembly 

to direct the billet to a position in which forced circulation was 

provided through the hole in the axis. 

Transfer time was observed to be less than 

TWC did not produce fuel clad tubing but supplies an intermediate product form 

to customers that do. 

product. 

The following customers required tower quenched 

GE Nuclear Energy, Nuclear Fuel and Components Manufacturing (USA), 

for Zr2 coextrusion outer shell 

Nuklearrohr-Gesellschaft MBH (Germany) for Zr4 tube hollows 

Kobe Steel (Japan) for Zr2 coextrusion outer shell 

The NRC team reviewed documentation produced in the process of using the Mark 

I1 Induction Furnace Quench Tower for beta quenching 33 billets from heat 

number 236531, Zr2, on shop order 3773, item 1, reference numbers 34 through 

66. 

a record of its temperature profile. 

determine that the billet had received the required thermal treatment. Two 

billets failed to receive the proper temperature profile prior to the first 

quench as shown on the induction furnace controller/recorder. These billets 

were reheated and requenched and the proper temperature profiles were 

recorded. 

Each billet quenched was instrumented with three thermocouples to produce 

Each temperature profile was reviewed to 

Barrier coextrusion is a tubular product with a commercially pure 

zirconium liner placed inside an outer shell of Zr2. 

- 12 - 

60 
Each billet was allowed a maximum of three attempts to obtain the proper heat 

treatment and quenching. TWC data reviewed by the NRC team did not show any 

billet beta quenched three times; TWC stated that in general less than 5% of 

the billets required beta quenching twice to obtain an acceptable thermal 

prof i 1 e. 

3.5.6 Operator Qualification 

The NRC team observed TWC personnel performing beta quenching operations and 

reviewed the training records of those persons. 

considered appropriate and personnel qualifications were current. 

The training recorded was 

3.5.7 Corrosion Resistance Testing 

Corrosion resistance testing was performed by TWC according to the customer 

purchase specification requirements. These requirements generally coincided 

with the requirements of a specified revision of ASTM G 2, "Standard Practice 

for Aqueous Corrosion Testing of Samples of Zirconium and Zirconium Alloys." 

The sampling plan for selecting corrosion resistance specimens was also 

specified by the customer purchase specification. A typical sampling plan 

randomly selected one to three specimens from the tail end of extrusions 

comprising a lot, typically one third of an ingot, about 2724 kg (6000 lbs). 

One corrosion resistance test specimen failure typically required two retest 

specimens. If one of the retest specimens failed the lot was rejected. 

3.6 Cracking Analysis 

The NRC team reviewed the TWC report dated June 13, 1994, regarding evaluation 

of cracks observed by SPC-ND in grid spacer components produced by Caran of 

Paramount, California, from strip supplied by TWC. The report made five major 

recommendations for changes in material processing, grid spacer forming 

tooling, and grid spacer design. 

and SPC-ND were being negotiated. The NRC team considered the quality of the 

evaluation and recommendations contained in this report as an instance of the 

strength of the TWC technical staff. 

Changes that could be affected between TWC 

3.7 Laboratory Facilities 

The NRC team visited production and research laboratory facilities. 

extensive metallography capabilities for polishing, etching, viewing, and 

photographing metallographic specimens. Numerous examples of award winning 

metallographic work were exhibited. 

were demonstrated to be particularly effective in identifying metallurgical 

phases, intermetallic compound compositions, and grain texture developed by 

various forming schedules. 

TWC had 

Heat tinting and color photomicrography 

Numerous autoclaves were observed for performing a variety of corrosion 

resistance tests discussed in Section 3.5.7 of this report. Chemical analysis 

equipment was present for determining gaseous and metal elements in the alloys 

produced by TWC. Tensile and bend testing facilities were also observed. 

- 13 - 

61 
APPENDIX 

PERSONS CONTACTED 

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

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

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

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

Teledyne Wah Chang: 

t 

t 

t 

Arbelbide, G.L. 

Bartcher, R.C. 

Brown, D.J. 

Denham, J.H. 

Doerfler, D. 

Graham, R.A. 

Harvey, D. 

Jackson, L. 

McNabb, M.D. 

Riesen, A.E. 

Sauder, P. 

Scal treto, A. 

Schraa, J. 

Tarver, T. 

Webb, B. 

Wilson, M. 

Manager, Process Assurance 

Director, Quality Assurance 

Engineer, Process Assurance 

General Counsel 

Calibration/Furnace Control Technician 

Technical Manager, Nuclear Products Group 

Ultrasonic Test Engineer 

Manager, Production Assurance and Inspection 

Vice President, Business Units 

President 

Contract Expediter 

Metallography Supervisor 

Laboratory Assistant 

Product Manager 

Metal 1 ographer 

Department Assistant, 

Production Assurance 

U.S. Nuclear Regulatory Commission: 

t 

t 

t 

t 

Brewer, D.H. 

Cwal ina, G.C. 

Matthews, S.M. 

Weeks, J.R. 

Metal 1 urgi cal Engineer, 

VIS/TSIB/DOTS/NRR 

Chief, TVIS/SIB/DOTS/NRR 

Quality Assurance Specialist, 

VIS/TSIB/DOTS/NRR 

Metallurgical Engineer, 

Brookhaven National Laboratory 

- A-1 - 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 205554001 

January 12, 1995 

Mr. Nicholas J. Liparulo 

Nuclear Safety and Regulatory Activities 

Westinghouse Electric Corporation 

P.O. Box 355 

Pittsburgh, Pennsylvania 15230 

SUBJECT: NRC INSPECTION NO. 99900404/94-01 

Dear Mr. Liparulo: 

This letter addresses the inspections at the Oregon State University (OSU) 

Advanced Plant Experiment (APEX) Test Facility in Corvallis, Oregon and at the 

Societa' Informazioni Esperienze Termoidrauliche (SIET) SPES-2 Test Facility 

in Piacenza, Italy, conducted by Richard P. Mclntyre and Uldis Potapovs of the 

Nuclear Regulatory Commission's (NRC's) Special Inspection Branch, George 

Thomas and Alan E. Levin of the Reactor Systems Branch, Robert M. Latta of the 

Quality Assurance and Maintenance Branch, David E. Bessette and Harold H. 

Scott of the Office of Nuclear Regulatory Research, and David T. Tang of the 

Standardization Project Directorate. The inspection at OSU APEX was conducted 

August 30 through September 1, 1994 and the inspection at SPES-2 was conducted 

October 11 through 14, 1994. The details of the inspection were discussed 

with management at each test facility and with your staff members present 

during the inspection and at the exit meetings on September 1, 1994, at OSU 

and on October 14, 1994, at SPES-2. 

The purpose of the inspections was to determine if testing activities 

performed at the OSU APEX facility and at the SIET SPES-2 facility to support 

design certification of the Westinghouse AP600 advanced reactor were conducted 

under the appropriate provisions of WCAP-8370, Revision 12A, the most recent 

Westinghouse Quality Assurance Plan (topical report) that has been approved by 

the NRC. This was implemented at OSU APEX by LTCT-GAH-001, "AP600 Long Term 

Cooling Test Project Quality Plan" and at SIET SPES-2 by 00006-QQ-92, "Quality 

Pl an Re1 ati ve to Nucl ear Orders. 'I 

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

enclosed inspection report. The inspection consisted of an examination of 

procedures and representative records, interviews with personnel, and 

observations by the inspectors. 

The results of the inspection indicate that Westinghouse, in general, was 

adequately implementing the AP600 Project quality assurance program at OSU and 

SIET, with the exception of a few findings in certain areas. Specifically, 

the team identified Nonconformances with program implementation with respect 

to (1) the calibration of instrumentation, (2) the accuracy of the facility 

as-built drawings, and (3) the lack of adequate control of drawing revisions. 

63 
Mr. Nicholas J. Liparulo -2- January 12, 1995 

Also, an Unresolved Item concerning the acceptance of test results by both OSU 

and SIET that failed to meet established test acceptance criteria, without an 

evaluation and disposition being included in the test design record file was 

identified. Westinghouse stated that it, not the testing organization, make 

the final determination as to what test results are acceptable and 

documentation of this evaluation and disposition would reside in the official 

design record file at Westinghouse offices in Monroeville. 

review Westinghouse design record files in Monroeville during a future 

inspection to determine if Westinghouse appropriately evaluated and 

dispositioned the test results that did not meet the acceptance criteria. 

The NRC will 

The responses requested by this letter and the enclosed Notice of 

Nonconformance are not subject to the clearance procedures of the Office of 

Management and Budget as required by the Paperwork Reduction Act of 1980, 

Public Law No. 96-511. 

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

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

Room. 

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

to discuss them with you. 

Si ncerel y , 

R. W. Borchardt, Director 

Standardization Project Directorate 

Associate Directorate for Advance,-I Reactors 

Office of Nuclear Reactor Regul ation 

and License Renewal 

Docket No. : 52-003 

Encl osures : 

1. Notice of Nonconformance 

2. Inspection Report No. 99900404/94-01 

cc w/encls: See Next Page 

64 
Mr. Nicholas J. Liparulo 

Westinghouse Electric Corporation 

cc: Mr. B. A. McIntyre 

Advanced Plant Safety & Licensing 

Westinghouse Electric Corporation 

Energy Systems Business Unit 

P.O. Box 355 

Pittsburgh, Pennsylvania 15230 

Mr. John C. Butler 

Advanced Plant Safety & Licensing 

Westinghouse Electric Corporation 

Energy Systems Business Unit 

Box 355 

Pittsburgh, Pennsylvania 15230 

Mr. M. D. Beaumont 

Nuclear and Advanced Technology Division 

Westinghouse Electric Corporation 

One Montrose Metro 

11921 Rockville Pike 

Suite 350 

Rockvi 11 e, Maryl and 20852 

Mr. Sterling Franks 

U.S. Department of Energy 

Washington, D.C. 20585 

NE-42 

Mr. S. M. Modro 

EG&G Idaho Inc. 

Post Office Box 1625 

Idaho Falls, Idaho 83415 

Mr. Steve Goldberg 

Budget Examiner 

725 17th Street, N.W. 

Room 8002 

Washington, D.C. 20503 

Mr. Frank A. Ross 

U.S. Department of Energy, NE-42 

Office of LWR Safety and Technology 

19901 Germantown Road 

Germantown, Maryl and 20874 

Docket No.: 52-003 

AP600 

Mr. Ronald Simard, Director 

Advanced Reactor Programs 

Nuclear Energy Institute 

1776 Eye Street, N.W. 

Suite 300 

Washi ngton , D . C . 20006-3706 

STS, Inc. 

Attn: Lynn Connor 

Suite 610 

3 Metro Center 

Bethesda, Maryland 20814 

Mr. Patrick W. Marriott, Manager 

Advanced P1 ant Technol ogi es 

GE Nuclear Energy 

175 Curtner Avenue, M/C 781 

San Jose, California 95125 

Mr. John E. Leatherman, Manager 

SBWR Design Certification 

GE Nuclear Energy, M/C 781 

San Jose, California 95125 

Barton Z. Cowan, Esq. 

Eckert Seamans Cherin & Mellott 

600 Grant Street 42nd Floor 

Pi t t s burg h , Pennsyl van i a 152 19 

Mr. Victor G. Snell, Director 

Safety and Licensing 

AECL Technol ogi es 

9210 Corporate Boulevard 

Suite 410 

Roc kvi 1 1 e, Maryl and 20850 

65 
NOTICE OF NONCONFORMANCE 

Westinghouse Electric Corporation 

Pittsburgh , Pennsylvania 

Docket No.: 52-003 

99900404 

Based on the results of Nuclear Regulatory Commission (NRC) inspections 

conducted on August 30 through September 1, 1994, at Oregon State University's 

(OSU's) Advanced Plant Experiment (APEX) low pressure integral system test 

facility and on October 11 through 14, 1994, at SIET's SPES-E full height, 

full pressure test facility, related to the AP600 certification activities, it 

appears that certain activities were not conducted in accordance with NRC 

requirements as identified below: 

A. Section 7.5.2 "Methods of Acceptcnce" of Section 7, "Control of Purchased 

Items and Services" of Westinghouse Electric Corporation Energy System 

Business Unit and Power Generati on Business Unit Qual i ty Assurance P1 an, 

(WCAP-8370, REV. 12A), dated April 1992, which has been accepted by the 

NRC as implementing the requirements of 10CFR50, Appendix 6, states, that 

when a Certificate of Conformance is used for acceptance of items or 

services, verification of the supplier's certification process is 

performed during audits or survei 11 ance, or both, at interval s 

comnensurate with the suppl ier's past performance. 

Contrary to the above, The Industrial Company which performed 

thermocouple calibration at the OSU APEX Test Facility, subcontracted the 

calibration of their primary standards to Industrial Instruments, Inc. 

and accepted their calibration certificates without having performed 

audits or surveillance of that company. 

validity of the flowmeter calibration certificates received from Foxboro 

Company, an unaudited commercial supplier. (94-01-01) 

Also, OSU failed to verify the 

B. Section 2, "Instructions, Procedures and Drawings" of WCAP-8370 REV. 12A, 

states that activities affecting the quality of items and services are 

accomplished in accordance with documented instructions, procedures or 

drawings that include appropriate quantitative and qualitative means of 

verifying quality. 

Procedure DP-5.0, Rev. 3, of Westinghouse Electric Corporation Nuclear 

and Advanced Technology Division Quality Assurance Program (WCAP-9565) 

states that instructions, procedures and drawings shall include or 

reference the appropriate quantitative or qualitative acceptance criteria 

for determining that prescribed activities have been satisfactorily 

accomplished. 

responsibilities for the preparation, review and approval of instructions 

and procedures. 

Procedure DP-5.0 also identifies organizational 

Enclosure 1 

-1- 

66 
Contrary to the above: 

1. 

2. 

3. 

4. 

No procedures or instructions were available or utilized to 

identify methods to be used or the accuracy and the acceptance 

criteria for determining the as-built elevations and dimensions of 

the Oregon State University (OSU) APEX Test Facility. 

The Industrial Company (TIC) Cal i bration Procedure TIC C. P. #19, 

which was used to calibrate thermocouples for the long term 

cooling tests at the OSU facility, did not identify who 

originated, reviewed, or approved this document. 

No instructions or procedures were available at the SIET SPES-2 

Test Facility to verify critical dimensions or configuration of 

comnercial manufacturing drawings before accepting these drawings 

as representing the as-built design condition and placing them 

under the SIET quality assurance system. 

The procedures for determining system and component elevations and 

arrangement at the SPES-2 test facility did not prescribe the 

required accuracy or i ncl ude any acceptance cri teri a for these 

measurements. (94-01-02) 

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

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

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

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

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

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

for each nonconformance: (1) a description of the steps that were or will be 

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

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

preventative measures were or will be completed. 

'Dated at Roc ville, Maryland 

This ia &- day of ='*'7 , 1995 

-2- 

67 
ORGANIZATION: Westinghouse Electric Corporation 

Pittsburgh, Pennsylvania 

REPORT NO. : 

CORRESPONDENCE 

ADDRESS : 

ORGAN I ZAT I ONA L 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY : 

INSPECTIONS 

CONDUCTED: 

TEAM LEADER: 

OTHER INSPECTORS: 

REV1 EWED : 

APPROVED : 

INSPECTION BASES: 

INSPECTION SCOPE: 

PLANT SITE 

APPLICABILITY: 

99900404/94-01 

Mr. Nicholas J. Liparulo 

Nucl ear Safety and Regul atory Act i vi ties 

Westinghouse Electric Corporation 

P.O. Box 355 

Pittsburgh, Pennsylvania 15230 

John Butler, Principal Engineer 

Advanced P1 ant Safety and Licensing 

(412) 374-5268 

Nuclear steam supply system components and services 

August 30 through September 1, 1994 at OSU APEX 

October 11 through 14, 1994 at SIET SPES-2 

c 

Richard P. McIntvre 

Vendor Inspection Section (VIS) 

Special Inspection Branch (TSIB) 

Uldis Potapovs, TSIB 

Robert M. Latta, TQMB 

Alan E. Levin, SRXB (SIET only) 

George Thomas, SRXB (OSU only) 

David E. Bessette RPSB/RES 

Harold H. Scott, RPSB/RES (OSU only) 

David T. Tang, PDST (SIET only) 

cu & 

L 

Gregodg. Cwal ina, Section Chief, VIS 

Robert M. Gallo, Chief, TSIB 

10 CFR Part 50, Appendix B and 10 CFR Part 21 

To determine if activities performed to support the design 

of AP600 and specifically, testing activities conducted at 

the Oregon State University (OSU) Advanced Plant Experiment 

(APEX) test facility in Corvallis, Oregon, and the Societa’ 

Informazioni Esperienze Termoidrauliche (SIET) SPES-2 test 

facility in Piacenza, Italy, were conducted under the 

appropriate provisions of WCAP-8370, Revision 12A, the most 

recent Westinghouse Quality Assurance Plan that has been 

approved by the NRC. 

None 

68 Enclosure 2 
1 INSPECTION SUMMARY 

0 Nonconformance 99900404/94-01-01 was identified and is discussed in 

Section 3.5.1. of this report 

Nonconformance 99900404/94-01-02 was identified and is discussed in 

Sections 3.4.1.4, 3.4.2.3, and 3.4.2.4 of this report. 

Unresolved Item 99900404/94-01-03 was identified and is discussed in 

Sections 3.4.1 and 3.6.2.2 of this report. 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

No previous inspections have been conducted in this area. 

3 INSPECTION FINDINGS AND OTHER COMMENTS 

3.1 Qualitu Assurance Proaram 

As a supplier of items and services to the nuclear industry, Westinghouse 

Electric Corporation (bl) has established a single quality assurance topical 

report that addresses the appl i cab1 e regul atory and industry requirements. 

The provisions of W's NRC approved topicai report which are identified in 

WCAP-8370, Rev. 12A, "Energy Systems Business Unit, Power Generation Business 

Unit Quality Assurance Plan," apply to all j! activities affecting quality of 

items and services including the design certification process for AP600. 

Accordingly, WCAP-8370 establishes W's commitments to meeting the requirements 

of 10 CFR 50 Appendix 8, ASME NQA-1 and NQA-2, and Regulatory Guide 1.28, 

Revision 3 "Quality Assurance Program Requirements (Design and Construction)" 

Section 17.3 of the AP600 Standard Safety Analysis Report (SSAR) states that 

activities supporting the design and design certification phase of the project 

are performed in accordance with bl Topical Report WCAP-8370 as supplemented by 

a project-specific Quality Plan. 

Design Quality Assurance Program Plan (QAPP), WCAP-12600, Rev. 2, affirms the 

commitments established in WCAP-8370 for the AP600 program and states that 

WCAP-12600 does not replace the WCAP-8370 document but establishes 

supplemental commitments for the AP600 Program. 

design certification test programs and related analyses are within the scope 

of this QAPP. 

The AP600 Advanced Light Water Reactor 

It also states that the AP600 

3.1.1 OSU APEX 

Specific quality provisions applicable to the OSU facility are identified in 

the AP600 Long Term Cooling Test Project Quality Plan (PQP), AP600 Document 

Number LTCT-GAH-001, Revision 1. This specific PQP establishes controls for 

the design and construction of the test facility as well as the conduct of 

testing activities. 

-2- 

69 
The inspection team reviewed the above documents to determine if testing 

activities performed at the OSU facility associated with the design of the 

AP600 advanced reactor were conducted in accordance with the appropriate 

provisions of W’s 10 CFR Part 50, Appendix B, quality assurance program. 

In particular, the inspection team examined the items and services covered by 

the quality assurance program to confirm that these activities were 

accomplished under suitably controlled conditions by properly trained 

personnel and that the pertinent test data used to furnish documentary 

evidence of activities affecting quality were appropriately recorded and 

mai nt ai ned . 

Based on these document reviews and discussions with project management 

personnel it was ascertained that the organizational structure defined in the 

PQP had been effectively implemented and that the overall quality assurance 

program administered by W provided for appropriate organizational independence 

and authority to identify conditions adverse to quality. Facil i ty personnel 

involved with managing and performing activities affecting quality were 

properly trained and knowledgeable of the quality elements embodied in the PQP 

and the associated W test control documents. 

3.1.2 SIET SPES-2 

The quality related activities associated with the AP600 full height, full 

pressure integral system tests performed at SPES-2 were conducted under the 

auspices of a cooperative agreement between W, ENEA, ENEL and SOPREN-Ansaldo. 

The purpose of this test program is to provide thermal hydraulic data for 

computer code validation and to simulate the operation of the AP600 plant 

passive safety systems. Specifically, the SPES-2 facility, which is operated 

by SIET, incorporates a full height, full pressure (FHFP), 1:395 volume scale 

simulation of the AP600 reactor vessel, reactor coolant system, and the 

passive safety injection system. 

The general features of SIET’s quality assurance system are described in their 

Quality Manual , 00001-QQ, Revision 3. Additional quality plan provisions 

specific to AP600 testing are also detailed in Procedure 00006-QQ-92, “Quality 

Plan Relative to Nuclear Area Orders,“ Revision 2, which includes a cross 

reference of European quality standards to the Criteria of NQA-1. In order to 

conduct these confirmatory tests, SIET has implemented an internal quality 

system which incorporates the requirements of both the International 

Organization for Standardization (ISO) Code Series 9000 (UNI-EN 29000) and the 

Italian Code UNI-8450 (1983). As described by SIET’s Quality Assurance 

Manager, these codes were derived from ASME NQA-1-1983 and IAEA No. SO-C-QA, 

Revision 1. 

The inspection team reviewed SIET’s quality system practices described in the 

Qual ity Manual and Procedure 00006-QQ-92 as well as other specific 

QA implementing procedures. 

inspection team was somewhat hampered by the unavai 1 abi 1 i ty of engl i sh 

translations of the subject documents (with the exception of Procedure 00006- 

QQ-92). However, using the interpretive services available at SIET, the 

inspection team successfully completed the examination of these documents. 

During the conduct of these reviews the 

-3- 

70 
Specific design requirements and quality elements associated with the SPES-2 

test facility are delineated in !# WCAP-14053, "AP600 FHFP Integral Systems 

Test Specification." These requirements include facility and test article 

design test matrix control, actuation setpoints, initial test conditions, test 

oversight and document approval , and qual i ty assurance requirements. Under 

the provisions of this cooperative effort ENEA contracted with SIET to perform SPES-2, AP600 test facil i ty modi f ications/construction, instrumentation 

calibration, QA plan development, facility operations and maintenance, and 

data acqui sition/transmi ttal . 

Based on the results of the team review it was generally determined that 

appropriate procedural controls had been establ ished at the SPES-2 test 

facil ity which properly incorporated the applicable quality provisions of 10 

CFR 50 Appendix B. 

the QA program had appropriately assigned functional responsibilities and that 

quality related activities were accomplished under suitably controlled 

conditions. 

It was also concluded that organizations participating in 

3.2 Instructions. Procedures and Drawincts 

3.2.1 OSU APEX 

The following 

test project were reviewed by the inspection team: 

qual i ty documents uti 1 ized for the AP600 Long Term Cool ing 

0 WCAP-9565, Nuclear and Advanced Technology Division (NATD) Quality 

Assurance Program Procedures (DPs) 

0 WCAP-12600, Revision 2, AP600 Quality Assurance Program Plan 

e WCAP-12601, Revision 13, AP600 Program Operating Procedures (APs) 

0 LTCT-GAH-001, Revision 1 , AP600 Long Term Cooling Test Project 

Quality Plan 

e PXS-T1P-001, Revision 2, AP600 Long Term Cooling Test 

0 DP-2.0, Revision 6, Quality Assurance Program 

Spec i f i cat i on 

0 DP-3.2.6, Revision 8, Preparation and Control of Drawings 

e DP-5.0, Revision 3, Instructions, Procedures and Drawings 

0 A-01, Revision 2, Testing Administration Procedure 

Based on the review of these documents it was determined that appropriate 

procedural controls had been developed and implemented to govern the conduct 

of quality related test activities at the OSU APEX facility. 

The inspection team also reviewed the following selected sample of test 

procedures to determine if these documents conformed to the provisions of 

-4- 

71 
procedure DP-5.0 and that they contained appropriate quantitative or 

qualitative acceptance criteria for determining if quality related test 

activities were satisfactorily accomplished: 

Test ID No. 

9400 1 

94002 

94003 

94004 

94005 

94007 

Test Procedure Title 

Pressurizer Vol ume Determination 

Primary Sump and Secondary Sump Volume Determination 

BAMS Moisture Separators Volume 

Incontainment Reactor Water Storage Tank (IRWST) 

Steam Generator 1 & 2 Secondary Side 

Reactor Vol ume Determination 

Vol ume 

As a result of these reviews it was concluded that the listed test procedures 

properly described test parameters including acceptance criteria. However, it 

was noted by the inspection team that the final acceptance of these documents 

which were characterized as in process records was to be conducted at W's 

Monroeville offices in accordance with Procedure AP-6.2, Revision 1, 

"Technical Document Release and Control . 'I 

3.2.2 SIET SPES-2 

The following quality documents used by SIET for the SPES-2 integral systems 

test were reviewed by the inspection team: 

Document No. 

00001 -QQ 

00002-QQ 

00003-QQ 

00004-QQ 

00005-QQ 

00006-QQ 

00007-QQ 

00008-QQ 

00009-QQ 

Title Revision No. 

Qual i ty Assurance Manual 3 

Document Control Procedure 2 

Instrument Management 1 

Quality Plan for Back Flow Prevention Test 3 

SI ET Organi zat i on 2 

Quality Plan Relative to Nuclear Area Orders 2 

Internal Audits Procedure 0 

Plant Instrumentation and Data Acquisition 

System Set-up Procedure 1 

Measurement Chain Check and Modification 

Procedure 0 

As previously noted all of the above documents except for Procedure 00O06-QQ7 

which was provided with an English translation, were prepared in Italian. As 

a result, the inspection team focused on examining SIET's site quality 

activities by conducting interviews with responsible personnel and evaluating 

the contents of each of the above documents which governed the SPES-E test 

program. This review indicated that, although some implementing procedures 

were st i 11 under development , the programat i c aspects i nvol vi ng qual i ty 

related activities were properly prescribed, performed, and documented in 

accordance with established instruction, procedures, and drawings. Results 

within this area were generally acceptable however, one concern was identified 

-5- 

72 
involving the lack of appropriate procedures and acceptance criteria for the 

confirmation of critical as-built dimensions. This concern is discussed in 

detail in Section 3.4 of this report. 

To gain additional insights into the implementation and control of quality 

related activities for the SPES-2 test program, the inspection team reviewed a 

selected sample of site design record files (DRFs), which for a typical matrix 

test, contained an organized set of information including the following items. 

Pretest QA checkl i st 

Pretest checkl ist 

Test procedure 

Deviation form 

Devi at i on report 

Day-of-test report 

Based on these evaluations it was determined that test information was 

properly recorded, reviewed, and approved for matrix tests in accordance with 

the comnitment described in WCAP-14053, "AP600 FHFP Integral Systems Test 

Specification. " With respect to design considerations and document control 

this review indicated that test procedures had been properly reviewed and 

approved by W's Monroeville office. Additionally, in order to determine the 

adequacy of administrative controls for testing activities at the SPES-2 

facility the inspection team performed a detailed review of the following 

procedures : 

Procedure No. Title 

00226PP93 

00293 PP94 

00227PP93 

00342PP94 

Cold Leg Break Plan and Procedures 

Cold Leg-Core Makeup Tank Balance Line Test 

P1 ant and Procedures 

Direct Vessel Injection Break Plan and 

Procedures 

Steam Line Break Plan and Procedures 

As a result of these reviews, the inspection team concluded that the above 

procedures properly described test parameters, incl uding acceptance criteria 

and QA activities. It was also noted that the final acceptance and closure of 

the design record files would be performed at y's Monroeville office in 

accordance with Procedure AP-6.2 , Rev. 1, "Technical Document Re1 ease and 

Control." 

3.3 Document Control 

3.3.1 OSU APEX 

The programnatic aspects of document control for activities affecting quality 

are described in Section 6.0 of the PQP. 

appropriate translation of these requirements into implementing procedures the 

inspection team reviewed the following procedures: 

In order to determine the 

-6- 

73 
DP-6.0, Rev. 4, "Document Control" 

AP-6.1, Rev. 0 , "AP600 Document Numberi ng" 

AP-6.2 , Rev. 1 , "Technical Document Release and Control 

AP-6.3, Rev. 0, "Preparation, Review and Approval of AP600 

Documents" 

As a result of these reviews and examination of the document control process 

it was determined that an effective system had been implemented to control the 

distribution and revision of quality related documents. In particular, the 

OSU records retention facility was efficiently operated with excellent 

controls over the distribution and maintenance of project documents including 

instructions, procedures and drawings. The effective administration of 

information resources of OSU was identified as a project strength. 

3.3.2 SIET SPES-2 

In order to evaluate the adequacy of the document control system at SIET the 

inspection team examined the administrative controls specified in Procedure 

000O2-QQy Revision 2 "Document Control Procedure" and Procedure 00006-QQ-92 , Section 4. 

associated with the preparation, issuance, and revision of documents which 

specify quality requirements or prescribe activities affecting quality. 

The inspection team also evaluated the implementation aspects 

Based on the results of these reviews and the examination of SIET's document 

control faci 1 i ty it was ascertained that appropriate procedural provisions had 

been developed for the identification and distribution of controlled documents 

including the proper assignment of responsibility for preparing, reviewing, 

approving and issuing qual i ty re1 ated documents. 

3.4 Desim Control 

3.4.1 OSU APEX 

The OSU APEX low pressure integral system test facility is a 1:4 linear scale 

model of the AP600 system, including a simulated reactor vessel and internals, 

two simulated primary coolant loops, a pressurizer and surge line, four 

simulated coolant pumps, two accumulators, and other associated piping and 

components. The facility is desl;ned to operate at 400 psi pressure and 450 

degrees F. 

Design control was reviewed by tracing the facility concept, scaling approach, 

generation of design drawings, facility construction, and characterization 

testing. Although the design and construction of the test facility was 

accomplished to comnercial quality standards, the scaling analysis documented 

in OSU-NE-9204 identifies certain critical attributes of the facility which 

would directly impact on the ability of the matrix test data to be used for 

benchmarki ng the computer codes devel oped for AP600 cert i f i cat i on. 

critical attributes are listed in Section 3 of the "OSU Implementation of the 

AP600 Long Term Cooling Test Project Quality Plan (LTCT-GAH-OOl)," document, 

dated May 5, 1994. These attributes are geometrical properties of the 

facility that must be measured to ensure that the important phenomena are 

adequately modeled in the facility and include the facility component (test 

The 

-7- 

74 
article) volumes, certain line resistances and the hot and cold leg piping 

inside diameters. Additionally, the !+! Long Term Cooling Test Specification 

states that all test articles and interconnecting piping shall be sized and 

located at the proper elevations based on the scal ing study. 

implementation plan states that the procedures to measure these critical 

attributes will be developed and implemented by W. 

equipment volumes are used as input parameters for computer codes: 

The PQP 

The following critical 

Core Makeup Tank (CMT) Volume 

Accumulator (ACC) Volume 

IRWST Volume 

Primary Sump Volume 

Secondary Sump Vol ume 

Automatic Depressurization System (ADS) 4 Moisture Separator 

Large Break Separator Volume 

Pressurizer (PZR) Volume 

Steam Generator (SG) shell side volume 

Reactor Volume 

Volume 

The team reviewed the results of several tests conducted to verify that the 

critical equipment volume tests were properly performed. The objective of 

these tests was to measure the volume of the major components of the test 

facility. The components were filled with water and drained. The drained 

water was weighed and the component volume was calculated. The calculated 

values agreed within 3% of the design values. These test results also agree 

with the values specified in the scalinq analysis report (Ref: OSU-NE-9204). 

conducted and the test 

lowing is a list of the test 

The team concluded that the tests were properly 

procedures were adequately implemented. The fo 

procedures reviewed: 

OSU-V-01, Accumulator Volume Determ 

OSU-V-02, CMT Volume Determination 

OSU-V-03. PZR Volume Determination 

nation 

OSU-V-04; IRWST Vol ume Determination 

OSU-V-05, Primary Sump and Secondary Sump Volume Determination 

OSU-V-06, SG #1 and #2 Secondary side Volume Determination 

OSU-V-07, ADS Moisture Separators and BAMS Moisture Separator 

Vol ume Determi nation 

The team also reviewed the documentation of the testing that measured the line 

resistances of the OSU model to verify that the tests were properly performed. 

The objective of these tests was to characterize the differential pressure 

versus flow for the piping connecting the major components of the test 

facility. The review consisted of: (1) determining if the tests had met the 

acceptance criteria (two had not); (2) checking a sample of the documentation 

of differential pressure (dp) measurement calibrations and verifying size 

marking on a sample of installed orifices; and (3) checking a sample of the 

hand-written values in the procedure tables with the typed numbers in the 

report (several typed values were incorrect). In two instances, @ reported 

that characterization tests had not met the acceptance criteria. The first 

instance was a test of ADS-1, -2, and -3 flow. The second was a test of 

-8- 

75 
primary sump njection flow. 41 stated that it, not OSU, make the final 

determination as to when test results are considered acceptable and that 

documentation of this evaluation and disposition would reside in the official 

design record file at W offices in Monroeville. The NRC will review 41 design 

record files n Monroeville during a future inspection to determine if 41 

appropriately evaluated and dispositioned the test results that did not meet 

the acceptance criteria and if sufficient supporting documentation exists. 

The team identified these failures to meet acceptance criteria without 

sufficient supporting documentation as part of Unresol ved I tem 94-01-03. 

3.4.1.1 Facility Scaling 

The program objective for the facility were originally specified by 41 in 

PXS-TlP-001, "Long Term Cooling Test," dated August 20, 1992. This document 

describes the background that led to the facility development and the 

objectives of the testing program. This report is a reasonable requirements 

document to guide the facility's conception. Following this document, OSU 

proceeded to develop the facility scaling. 

The facility scaling basis is described in "Scaling Analysis for the OSU AP600 

Integral System and Long Term Cooling Facility" (OSU-NE-9204) Draft 1993. 

This report describes the basic rationale for designing a scaled test facility 

representative of the AP600 at 1:4 height and 1:192 volume scale. This report 

has been previously reviewed in some detail by NRC staff and the ACRS. 

Revisions and additions are still being made based on comments received and 

the final staff review must await receipt of the final report. 

The scaling report describes the general scaling methodology and its 

relationship to Ishi i-Kataoka scal ing (NUREG/CR-3267) and Hierarchical Two 

Tier Scaling (NUREG/CR-5809). The report was discussed with its author, the 

OSU Program Manager. 

circulation and depressurization processes, which are relatively distinct * 

portions of the experiments that comprise the test program. 

similarity scaling groups are identified and compared between the scaling 

facility and the prototype plant. 

measure of the degree to which the process is distorted. 

The report addresses the scaling approach for natural 

Nondimensional 

The deviation of the ratio from one is a 

Most facility dimensional requirements follow from three basic decisions made 

to fix the facility design. The first is the 1:4 height decision and is a 

qualitative, engineering judgement. It takes into consideration the minimum 

height required to retain reasonable representation of gravity driven natural 

circulation forces, the size of the existing building to house the facility, 

and the cost. The second decision is the diameter scale (=1:7), which is 

determined by the minimum dimension required to obtain representative flow 

regime transitions in the hot and cold legs. The l/d (length to diameter) 

ratio of the hot and cold leg was adjusted slightly (shortened and fattened). 

Similarly, the surge line l/d ratio was adjusted to the extent reasonable. 

The third decision was the pressure scale (400 psi) which is an economic 

deci si on. 

-9- 

76 
3.4.1.2 Facility Design 

Following these basic decisions, the next set of design decisions follow from 

piping friction resistances or Friction Number scaling. 

was present during the inspection and the design process was discussed, 

however, the associated files could not be examined since the work was 

performed by W in Monroeville. 

constructor (TIC) for the purposes of meeting Section VI11 of the ASME code 

and State of Oregon requirements. 

The chief designer 

The design was independently verified by the 

The reactor coolant pump is not fully representative of the prototypic pump 

since it has a different stopped rotor resistance. This resistance is, 

however, a relatively small component of the total loop flow resistance, which 

is dominated by the core and the steam generator. The core design has been 

simplified to reduce cost, but the resistance was matched. 

The facility dimensional tolerances were generally specified to be & 0.01 

inches. 

placement of components was 1/4 inch in the vertical direction and 1 inch in 

the horizontal direction. 

built review. 

Standard sizes were used for piping and tubing. The tolerance for 

The placement of components was verified by an asThe basic objectives and rationale for instrumenting the facility is given in 

PXS-T1P-001, "Long Term Cooling Test Specification." The instrumentation is 

intended to provide redundant means to determine mass and energy balances. 

Generally, there are two flow meters in series to provide a check. Discharge 

mass flow is measured by a flow meter and by a load cell on the catch tanks. 

The discharge flow passes through a separator, with the liquid flowing to the 

catch tank and the vapor to atmosphere. Both flow paths are instrumented to 

measure flow rate and temperature. The discharge flow measurement as it is 

instrumented would appear to provide good measurement of mass and energy loss 

to determine system mass and energy balances. The mass balance was stated to 

be within & 2%. 

3.4.1.3 Facility Construction 

The OSU low pressure integral system test loop was procured as a commercial 

grade facility and constructed to the requirements of Section VI11 of the ASME 

Code. The facility constructor, The Industrial Company (TIC) , Northwest is an 

ASME Section VI11 Certificate holder and provided on-site QA/QC manager to 

follow the construction quality. All process control and test instrumentation 

was also procured to comnercial quality standards. Approximately 750 

instruments for the measurement of flow, pressure, heat flux, power, 

temperature, and other parameters input into the data acquisition system. 

3.4.1.4 Verification of As-built Dimensions and Elevations 

According to the 4? Project Manager, the as-built dimensions of the facility 

were determined during a system walkdown by contract personnel (Jacobsen 

Engineering) under H's supervision. The as-built dimensions were documented 

on isometric drawings which became a part of the system turnover package. 

-10- 

77 
I 

Selected isometrics were reviewed to assess the as-built verification process. 

Drawing No. ADS 1-3, Rev. 1, Automatic Depressurization System, for example, 

was an isometric sketch of this system with detailed dimensional information 

on the specific vertical and horizontal pipe runs, outside and inside pipe 

diameters, and specific notations regarding the type of connections. 

Title Block contained several sign-off spaces which were initialed or signed, 

however, the title or organizational affiliation of the respective signers was 

not identified. Most of the layout dimensions were measured to fractions of 

inches with the smallest fraction being 1/4 inch. 

generally indicated as nominal pipe size although some inside diameters were 

recorded to three decimal places. 

diameter dimensions were obtained by actual measurement. 

arrangement drawing (GAS-AP600-01) showed system elevations in fractions of 

inches with the smallest fraction being 1/16 inch. 

The 

Piping dimensions were 

A general 

It was not indicated whether the inside 

No procedures or instructions had been prepared to provide guidance for the 

verification of the as-built dimensions. The failure to define significant 

parameters such as the measurement technique and the required accuracy or 

acceptance criteria for the as-built dimensional verification was identified 

as an example as part of Nonconformance 94-01-02. 

As-built drawings were also reviewed to verify that the OSU model "actual" 

values assumed in the scaling report OSU-NE-9204 are correct. 

components (such as CMT and IRWST) physical dimensions such as height and 

diameter that are given in the scaling report were compared with the values 

from the as-built drawing. 

within a reasonable range of values given in the scaling report. 

Major 

The as-built drawing dimensions were found to be 

A sample of facility items that have been changed to keep current with the 

AP600 design changes were also checked on the drawings, e.g., the direct 

vessel injection nozzle, core bypass holes, the rerouted cold leg pressure 

balance line, and piping orifices. 

on the drawing, but the value in the database was different. Also, as-built 

drawings show orifices that are not installed, however, the database appeared 

correct, i.e., only installed orifices are listed. 

One orifice (ORI-451) was marked correctly 

3.4.1.5 Facility Characterization 

Facility characterization testing was performed to verify line resistances, 

component volumes and to measure facility heat losses. The facility heat 

losses were measured in integral terms by heating the facility to full 

operating temperature and adjusting core power until a steady state balance 

was achieved. The measured value of 30 KW can be compared to the maximum 

facility power of 600 KW, which represent 5 percent of scaled full power. 

Therefore, heat losses will be in the range of 5 to 15 percent during the 

course of an experiment and should result in an acceptable level of distortion 

from the standpoint of providing an experimental representation of a 

transient. 

During startup testing, some differential pressure instruments were found to 

be improperly ranged. The ranging was based initially on NOTRUMP 

-11- 
calculations. A substantial effort was required during startup to ensure 

grounding of instruments and to eliminate signal disturbances. 

3.4.2 SIET SPES-2 

SPES was originally commissioned by ENEA to simulate a 

pressurized water reactor with Italian specific design features uti1 izing a 

volume scale of 1:427, full height, and full pressure. For SPES-2 testing, 

the facility was modified to model the AP600 design using, when possible, the 

components of the existing plant (for example, steam generators). 

in early 1992, planning began to modify the facility for AP600 testing. 

Construction was completed in mid-1993, after which a period of shakedown and 

characterization testing comnenced. Matrix testing began in early 1994 and 

was completed in October 1994. 

312 (3-loop) 

Beginning 

3.4.2.1 Facility Scaling 

The SPES-2 facility is 1:395 scaled volume with respect to the AP600. 

scaling criteria maintained selected significant parameters including the 

fluid thermodynamic conditions, vertical elevations, power to volume and flow 

rate ratios, transit time of fluid, and thermal flux. Additional scaling 

criteria were applied to the design of selected components such as lines in 

which friction pressure drops were relevant. 

SPES-2 is full height and so its diameter scale is approximately 1:20. 

ratio distorts the surface to volume ratio and, therefore, heat loss to the 

The 

This 

environment. The scaled heat loss is about 8 kW while the actual 

about 150 kW, therefore, the distortion is approximately 140 kW. 

a very severe problem for larger LOCAs (i .e. greater than break s 

3 inches) but it is a problem for smaller LOCAs. This is because 

proceed longer and decay heat decreases to less than the facility 

Therefore, in order to run sensible small break LOCAs, heat loss 

compensated by adding the heat loss to the core power. 

The heat transfer area in the SPES steam generator is underscaled 

approximately 50% compared to AP600. 

primary system conditions, it is necessary to lower the secondary 

pressure during the start of the experiments. 

To obtain the correct heat 

heat loss is 

This is not 

zes of about 

small LOCAs 

heat loss. 

S 

by 

ransfer and 

side 

3.4.2.2 Facility Design 

The facility design is described in WCAP-14053, "AP600 SPES4 FHFP Integral 

Systems Test Specification." The design parameters were then converted into 

design requirements described in WCAP-13277, Rev. 1, "Scaling, Design, and 

Verification of SPES-2. " WCAP-13277 requirements were then used for 

preparation of design drawings, procurement activities, and facility 

construction. The completed facility, then, is described in WCAP-14073, 

"SPES-2 Facility Description Report," (May 1994). Several updates to this 

last report were noted in the master copy at SIET, but had not been reissued. 

The design modifications to the facility were prepared interactively by SIET 

and W. SIET drafted drawings and sent them to W, where they were reviewed and 

-12- 

79 
adjusted. The iterations continued informally for some period of time until 

the design drawings reached a final draft stage, at which time they were 

placed under change and approval control. 

against the scaled dimensional requirement to assure conformance. 

W checked the final drawings 

Much of the previously existing SPES facility was retained in SPES-2. 

the steam generators were retained, as was the reactor vessel, pressurizer, 

and reactor cool ant pumps. tube sheet to the hot leg and cold leg centerlines was preserved. The 

pressurizer elevation was changed to represent proper scaling of its bottom 

elevation with respect to AP600. 

provides some distortion in volume versus elevation. 

based on the Wilson bubble rise model and should preserve level swell. A new 

surge line was added representing the correct AP600 geometry. 

this line (1 1/2 inch schedule 160) is small in terms of maintaining flow 

regimes. 

tanks, accumulators, passive residual heat removal (PRHR) system, and 

automatic depressurization system. 

Two of 

The elevation difference from the steam generator 

The pressurizer is 60% of full height, which 

Its aspect ratio was 

The size of 

New components added to the SPES-1 facility included core makeup 

A number of design changes and their rationale were reviewed. Much of the 

scaling and design work associated with modifying the facility consisted of 

maintaining elevations and vertical dimensions at 1:l in order to preserve 

gravity driving force and to maintain piping friction resistances, or Friction 

Number scaling. By obtaining the correct driving force and the correct 

resistance, natural circulation should be well represented. 

New hot legs and a new surge line were added for correct geometric 

representation. 

the transition to stratified flow. 

the reactor vessel was replaced. 

and four cold legs, with center line elevations preserved. 

injection lines also enter this part of the vessel. 

region is represented as an annulus. 

represent the circumferential flow resistance present in the full scale plant, 

however, matching the flow resistance between adjacent cold legs was not an 

objective. 

The hot leg was scaled according to Froude number to preserve 

The upper plenum/upper downcomer region of 

It includes the arrangement of two hot legs 

The direct vessel 

The annular region includes fins to 

The upper downcomer 

At the lower end of the upper plenum/upper downcomer, the downcomer converts 

from an annulus to an external pipe, which parallels the core region of the 

vessel and reenters in the lower plenum. Natural circulation flow patterns 

that may result from this geometry were calculated using a computational fluid 

dynamics code. 

thermocouples. 

reasonable agreement obtained. 

The results were used to help determine locations for 

The results were also compared with a RELAP calculation, with 

The ADS4 valves have been oversized by about a factor of three in SPES 

compared to the scaled value from AP600. 

performed by Ansaldo using RELAP5. The calculations indicated that, because 

the metal mass of SPES is substantially overscaled, the resultant distortion 

in vapor generation during the approach to IRWST injection was large. 

This was done based on analyses 

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80 
The inspector reviewed the "master copy" of the facility description report 

for SPES-2. This copy includes a penciled-in list of corrections and 

modifications to the original report. The pencil corrections did not appear 

to have been incorporated as yet into the as-built design drawings. 

Additionally, the SPES-2 facility description report contains a list of 

instrumentation. From examining the list of instruments, one can infer that 

the underlying rationale is to attempt to track mass and energy in the system 

during experiments. 

states to be measured and the method of measurement is described. 

In addition, the relation between the processes and 

The team noted that, for all design documentation, such as reports and 

drawings, relevant information was properly archived and easily retrievable by 

SIET personnel. 

3.4.2.3 Facility Construction 

The SPES test facility was designed and constructed to commercial standards. 

The existing plant modifications were performed by commercial contractors or 

SIET also using commercial standards. 

During the early phases of the SPES facility modification, component and 

construction drawings were not covered by the SIET quality assurance program 

and design changes to manufactured components or plant system arrangement were 

not controlled under the SIET quality system provisions. Several revisions of 

the facility drawings were made during this period. 

recorded by letter designations (A,B,C, etc.). After the facility design was 

"stabilized", the drawings were placed under the SIET QA program controls and 

further revisions were issued with numerical revision designations. According 

to SIET management, the first numerical revision of each drawing was verified 

by the technical manager for conformance to design requirements by checking 

as-built dimensions. However, there was no formal record of such verification 

and this process was not controlled by any procedures or instructions. 

These revisions were 

A1 1 revi si ons of Drawing No. 25.01.02 , SPES-2 Annul ar Downcomer, were reviewed 

to assess the traceability of modifications through the construction phase of 

the facility and to verify the validity of the as-built dimensions. This 

drawing encompassed three sets of letter revisions: B (October 20, 1992), C 

(December 3, 1992), and D (October 22, 1993). The revisions incorporated 

significant changes to the drawings which could be identified by examining 

file copies of marked up drawings. 

specifically identified in the revised drawing either by summarizing or 

describing them in the appropriate space in the drawing title block or by 

highlighting the actual changes on the drawing. 

These changes, however were not 

The first numerical revision of drawing No. 25.01.02 (Revision 0) was issued 

on October 29, 1993, and was marked "issued" in the revision b ock. According 

to the system described above, Revision 0 was issued under the SIET QA system 

and represented the as-built status of the facility as of that date. It could 

not be determined by examining Revision 0 and the previous rev sion (Revision 

D) which dimensions of the drawing were actually verified before placing this 

drawing under the SIET QA system as representing the as-built status of the 

facility. Several dimensions on a copy of Revision D of the drawing were 

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circled, others were underlined, and some were both circled and underlined. 

The significance of these marks was not explained and it could not be 

determined which dimensions were actually verified or whether the verified 

dimensions included the critical dimensions of this particular assembly. 

Revision 1 of this drawing was issued on November 3, 1993, and appeared to be 

identical to Revision 0. The revision block contained the notation "As-Buil t" 

with no indication whether it incorporated any changes from the previous 

revision (Revision 0). The failure to provide instructions or procedures for 

the verification of critical as-built dimensions when placing uncontrolled 

drawings under the QA system was identified as an example as part of 

Nonconformance 94-01-02. 

3.4.2.4 Verification of As-built Dimensions and Elevations 

The AP600 SPES-2 FHFP Integral System Test Specification (WCAP-14053) states 

that all test articles and interconnecting piping shall be sized and located 

at the proper elevations based on a scaling study using prototypical AP600 

dimensions as the basis. Since the test facility was constructed to 

comnercial standards without significant QA involvement, conformance with 

critical system elevations, dimensions, and volumes was established by 

performing verification of these attributes under the SIET QA system. 

determination of critical dimensions was accomplished by uti1 izing a 

combination of as-built drawing dimensions and measured layout/elevation 

dimensions. According to the SIET Technical Manager, critical elevations were 

determined by establishing a reference level (the top of the steam generator 

tube plate), stringing horizontal wires at that elevation, and using plumb line/liquid level techniques to obtain other reference planes from which 

specific elevation measurements were taken. 

The 

A1 though a general SIET procedure ddscri bing different measurement techniques 

(00216P093) was available, this procedure did not identify specific 

measurement methods to be used for a particular application or prescribe the 

measurement accuracy required for each type of measurement. A review of the 

system elevation drawings and isometrics shows linear dimensions and 

elevations in millimeters with an apparent accuracy to one millimeter over a 

distance of 20,000 millimeters. 

measurements. 

acceptance criteria for the verification of the as-built facility dimensions 

was identified as an example as part of Nonconformance 94-01-02. The team 

reviewed several design drawings for the SPES-2 facility, including 

preliminary, issued, and as-built versions. All drawings appeared to be 

appropriately signed off by SIET personnel. Some were missing the "final 

approval" signature; SIET's explanation for this omission was that the final 

approval was done by W, and the final approval is thus documented by 

correspondence, but not shown on the drawing itself. 

find the records of communications between and SIET indicating review and 

approval of drawings in the appropriate volumes of the DRF. 

No tolerances were indicated for these 

The failure to specify accuracy requirements or include any 

The team was able to 

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3.5 Procurement and Cal i bration of Test Instrumentation 

3.5.1 OSU APEX 

The OSU PQP states that the procurement of instrumentation and calibration 

services shall i ncl ude documentation of cal i brat ion traceabl e to national 

standards. 

of appropriate cal i bration procedures and that these procedures shall be 

approved by the OSU Program Manager and submitted for Quality Assurance review 

and Project Manager information. Section 5.0 of the PQP also states that 

activities affecting quality shall be performed in accordance with documented 

instructions or procedures and that such instructions or procedures shall be 

prepared in accordance with DP-5.0 which establishes specific responsibilities 

and requirements for the preparation, review, and approval of such 

procedures. 

It also states that OSU shall be responsible for the development 

All instrumentation for the OSU test facility was procured by either OSU or 

TIC to comnerci a1 standards. Acceptabi 1 i ty of this equipment for qual i ty 

related activities was typically demonstrated through on-site calibration 

after delivery. Instrumentation is carefully arranged and fully labeled, 

including channel number, serial number, and calibration sticker. TIC 

performed the initial instrumentation calibration and installation. Each type 

of instrument was procured from a single manufacturer. For example, the 

pressure and differential pressure instruments (p and dps) are from Rosemount, 

thermocouples are from Gordon, and flow meters are from Foxboro. Record 

keeping of instrumentation calibration was thorough. The current record of 

each instrument is kept in a well organized notebook. Prior records are 

stored in a file cabinet. The test equipment is identified and the next 

calibration due date is noted. 

Transcat or Fluke. Selected records of procurement- and calibration activities 

were reviewed as sumnarized below. 

The test equipment is checked annually by 

Most of the thermocouples used in the test facility were calibrated by TIC at 

ambient and two elevated temperatures (225F and 450F) and checked for 

linearity of signal prior to installation using certified calibration 

equipment. The calibration procedure (TIC C.P. #19) consisted of a half page 

document which was not dated and did not identify who originated, reviewed, or 

approved the document. According to this procedure, the ambient temperature 

calibration was to be accomplished using a certified temperature calibrator 

while the elevated temperature calibrations were to be obtained using a dry 

well calibration device. The procedure did not identify specific well sizes 

to be used for different thermocouple configurations but stated that the 

"proper well size" was to be used. 

preparation, review and approval chain on calibration procedure TIC C.P. #19 

was identified as an example as part of Nonconformance 94-01-02. 

Thermocouple cal i brat i on 

The failure to identify and record the 

Review of the calibration records for the temperature calibrator and the dry 

well calibrator indicated that the calibration certificates for these devices 

were issued by Industrial Instruments, Inc. of Sandy, Utah. These 

certificates stated that the instruments had been calibrated using standards 

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a3 
whose accuracies meet or exceed acceptabl e cal i brat i on ratios and are 

traceable to either The National Institute of Standards and Technology (NIST) 

or to nationally accepted consensus standard. The certificates were signed by 

a metrologist and QA technician and were stated to be effective for one year 

from the calibration date (December 1993). 

TIC did not invoke any QA requirements for the calibration/certification 

services performed by Industrial Instruments, Inc. and had not performed any 

audits or surveillance of that company. Section 7.5.2 of WCAP 8370, which is 

referenced in the PQP as applicable to activities important to the test data, 

states that "When a CoC is used for acceptance ... verification of the 

supp7ier's CoC and certification process is performed during audits or 

surveillance." The acceptance of calibration certifications from an unaudited 

company was identified as an example as part of Nonconformance 94-01-01. 

Flow Meters and Pressure Transmitters 

Review of selected purchase orders for flow meters and pressure transmitters 

indicated that these devices were procured by OSU as commercial grade 

equipment to a catalogue model number without referencing any special quality 

or technical requirements. 

can independently assure the desired output accuracy (cal ibration) through onsite qualification, it was noted that the flowmeter calibration was also 

performed by a commercial suppl ier (Foxboro) . The failure to verify the 

validity of the calibration certificates received from an unaudited commercial 

supplier is identified as an example as part of Nonconformance 94-01-01 as 

discussed above. 

Whi 1 e such practice may be appropriate when OSU 

Review of the OSU procedure for the calibration of commercial grade Rosemount 

pressure transmitters indicated that this procedure had been prepared in 

accordance with the applicable PQP requirements. It contained detailed 

technical instructions as well as the appropriate reviews and approvals. 

3.5.2 SIET SPES-2 

The calibration of all measuring instruments at SIET is controlled in 

accordance with procedure 00003-QQ. 

identified with a registration number. 

instrument, including all calibration information is reported on an instrument 

card traceable to the instrument through the registration number. 

instrument configuration for each test is reported in each test procedure. 

Exh instrument is individually 

The technical data pertaining to this 

The 

The primary test instruments at the SPES-2 facility include thermocouples, 

pressure transmitters, and flowmeters. The flowmeters are generally venturi 

or calibrated orifice type although turbine meters are used in some 

applications. These instruments are typically procured to commercial 

standards (UNI) and calibrated by SIET. The instruments are required to meet 

Ita1 ian national consensus standards and are also calibrated to accuracy 

ranges established by these standards for specific instrument types. 

accuracy ranges prescribed by these standards have reportedly been reviewed by 

SIET, and found to meet or exceed the accuracy requirements stated in the FHFP 

Integral Systems Test Specification. 

The 

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84 
SIET has a functional metrology laboratory which is capable of calibrating all 

test instrumentation used in the integral systems tests. All primary 

standards used in the metrology laboratory are periodically calibrated to 

Ita1 ian national standards under the Systema Italian0 Taratura (SIT). 

system includes three organizations classified according to the type of 

calibration (mechanical measurement, electrical measurement, etc.) which are 

empowered to accredit specific calibration laboratories. 

Institute Meterologico Gustava Colonnetti is the organization empowered to 

accredit 1 aboratori es performing cal i brat i ons of mechanical and thermal 

measurement devices. 

in a register of accredited laboratories. 

in this register to calibrate their primary calibration standards. The SIT 

system, in turn, is audited by the Western European Calibration Cooperation 

(WECC). 

Thermocouples are calibrated by the SIET metrology laboratory using a 

comparison method. 

salt) are available depending on the temperature range of interest. The 

primary standard is a resistance thermometer calibrated by an accredited 

laboratory with the calibration traceable to Italian national standards. The 

cal i bration is performed in accordance with an approved procedure (00165P092) 

specific to this activity. Pressure transmitters are calibrated using a dead 

weight method to SIET procedure 00118P091, while flow meters are calibrated to 

SIET procedure 00211PP92 using a gravimetric method. 

were written in Italian, a detailed review of their contents was not made. A 

limited translation, combined with observation of calibration activities and a 

review specific calibration records indicated that the calibration process was 

adequately control 1 ed. 

This 

For example, 

Laboratories accredited under this system are identified 

SIET uses only laboratories listed 

No user audits are performed by organizations utilizing this system. 

Three constant temperature baths (water, oil, or molten 

Since these procedures 

3.6 Test Control 

3.6.1 OSU APEX 

Westinghouse Engineering Procedure AP-3.11 , "AP600 Testing," provides Specific 

AP600 test engineering instruction for implementing the requirements of the 

AP600 QAPP (WCAP-12600) and the NATD QA Program (WCAP-9565). AP-3.11 requires 

the preparation of a test specification for all safety-related AP600 tests. 

The team reviewed PXS-T1P-001 , Rev.2, "AP600 Long-Term Cool ing Test 

Specification," to verify that provisions of AP-3.11, including purpose, test 

objectives, test facility requirements, test articles, instrumentation 

requirements, data acquisition, test operation, test reports, data 

requirements, and quality assurance requirements required far the tests, were 

adequately covered in the test specification. The team determined that the 

PXS-T1P-001 test specification included all the appropriate provisions and was 

found to be adequate for the tests. 

AP-3.11 also gives guidance for the preparation of the test procedure used for 

the conduct of the tests. 

the preparation of the test procedure. 

for the tests SB-2, SB-3, SB-4, and SB-5, to verify that the test procedures 

used for the conduct of the tests were prepared according to Appendix C of 

AP-3.11. The team determined that test objectives, prerequisites, initial 

Appendix C of AP 3.11 is a sample format given for 

The team reviewed the test procedures 

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85 
conditions, instructions, acceptance criteria, and post test activities for 

the conduct of the tests were included in the test procedures. 

procedures used for the test were found to be acceptable. 

The test 

3.6.1.1 Test Matrix 

Following completion of facility characterization, a matrix testing program 

comprised of 14 experiments was conducted. 

break sizes in the cold leg, direct vessel injection line, and pressure 

balance line. Then, a series of 12 additional tests were defined with seven 

having been performed at the time of the inspection. The remaining five were 

to be completed by the end of September, 1994. The 42 rationale for the 

additional 12 tests was not reviewed, however, the test matrix in total was 

reviewed. 

These tests investigate a range of 

The test matrix appears to provide a reasonable coverage of varying break 

sizes, break locations and break orientations. 

cold leg, the PRHR cold leg, the PRHR hot leg, the direct vessel injection 

line, and the pressure balance line. A two inch break in the bottom of the 

CMT cold leg was chosen to perform a number of parametric variations in 

operation of the ADS and non-safety systems. 

Breaks-are included in the CMT 

3.6.1.2 

The instrumentation system operates normally with data recorded at eight 

second intervals. 

frequently, however, the interval over which this can occur is limited because 

of the capacity for data storage. 

are changing slowly, the limited recording frequency is a concern because it 

may not catch higher frequency phenomena. The data must be reviewed carefully 

to determine whether there is evidence that this may have occurred. 

uncertainty evaluation was performed by W. 

inspection team. 

OSU performs the data reduction from the data acquisition system (DAS) to 

optical discs. 

data. The reduction converts voltage signals to engineering units (with the 

exception of thermocouple data which is recorded on the DAS directly as 

temperature) and parses the data in about 1/2 hour. 

of level differential pressure cell to produce a true collapsed liquid level 

is performed by y and not OSU. The software was subjected to a verification 

process that appeared to be very thorough. 

measurement type was performed. Then every channel was verified for several 

recording points at the beginning and end of a test, which basically amounted 

to a complete channel by channel verification. A check was made for roundoff 

error as part of the software validation, with none found. 

any reading that indicate a clearly failed instrument, i.e. very high or very 

low. 

test. 

Test results and Data Processing 

It can operate in the "burst mode" to record data more 

While this may be adequate when conditions 

A data 

This was not examined by the 

OSU developed a software program called REDUCE to process the 

Temperature compensation 

First, one channel from each 

The software flags 

Differential pressure zero points are checked before and after each 

Thermocouples are checked against each other. 

The OSU APEX operations team does not perform any formal data analysis. 

work is performed by W in Monroeville so, therefore, was not examined. The 

This 

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86 
testing is conducted in a production mode, with normally two tests performed 

per week. The timing restricts the extent to which the data can be reviewed 

while the testing is in progress. This permits limited feedback from data 

analysis to the testing program. Often, data review can highlight the need 

for additional instruments or tests to evaluate a phenomena or process. The 

production mode of testing permits limited opportunity for such feedback. 

addition, the data review will continue after the current test operation 

disperses. Thus, there may be some concern over whether 

able to consult with operations personnel when questions arise on suspect data 

indications. 

In 

analysts will be 

Nevertheless, some feedback is occurring. Additional thermocouples are being 

placed in the cold legs to measure thermal stratification. Also, instruments 

are being placed in the cold legs, lower plenum and upper plenum to measure 

condensation events which have occurred in several of the tests. The 

instruments are fast response pressure transducers with a high frequency 

recording system. Currently, there are apparently no plans to instrument the 

PRHR to measure condensat ion despite the probable occurrences of condensation 

events in APEX. 

3.6.1.3 Staff Training and Test Conduct 

The operations team at OSU is comprised of OSU, TIC and personnel. The team 

experience was stated to exceed 100 years. 

operators were reviewed. 

administered covering a wide range of subjects including instrument theory and 

practice. The instrumentation personnel were experienced with calibration and 

troubleshooting. 

The training records of the OSU 

Each has extensive experience and exams were 

Test acceptance criteria are simple but consistent with the contractual 

obligations of OSU. 

data is recorded. 

with a DAS computer such that the data was not recorded. 

The tests are run according to the test procedure and the 

Repeat of tests has been limited to when problems occurred 

Early in the testing program, problems were encountered with failures of 

magnetic flow meters in the cold legs. 

that normally occurs during experiments due to operation of the PRHR. 

thermal stratification cracks allowed water to leak into the meter's 

electronics. While the instrument was tested to withstand thermal shock 

causing uniform change in flow temperature, it evidently cannot withstand 

highly non-uniform temperature. Seven magnetic flow meters failed before' W 

decided to forego the measurement. The manufacturer (Foxboro) is currently 

trying to resolve the problem. Aside from the cold legs, the magnetic flow 

meters have been used elsewhere in the system without problems and have 

provided excellent data (when conditions are nearly single phase). 

The cause was thermal stratification 

The 

3.6.1.4 Witness of Matrix Test SB-31 

The team witnessed Matrix Test SB-31 on August 31, 1994. 

test was to obtain thermal hydraulic data with spurious "S" signal, but 

without actuation of the ADS valves. 

available for use at the facility. The team briefly reviewed the test 

The objective of the 

A copy of the test procedure was made 

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87 
procedure OSU-MATRIX-SB31, Revision 0, "Spurious 'S' Signal Without ADS 

Actuation and with No Break Simulation, Long-Term Cooling Operation." The 

test supervisor gave a short pre-test briefing on procedures and expected 

system behavior. 

The "S" signal scrams the facility, opens the CMT isolation valves and the 

PRHR outlet valves. 

Cold CMT water drains into the core, expands, and consequently increases the 

Pressurizer level. Meanwhile, PRHR cools the core and shrinks the pressurizer 

level. Consequently, the pressurizer level was oscillating. From the control 

room observations, the system interactions and the system behavior were as 

expected. The test was performed for about one hour and 30 minutes. 

Prior to the conduct of the test, the inspection team witnessed the completion 

of the test prerequisites including the briefing of operations personnel. The 

inspection team determined that the specified test parameters were properly 

established and that initial conditions and system 1 ine-ups were appropriately 

performed. 

excel lent comnunications and the resultant test data was accurately recorded. 

No discrepancies were identified by the inspection team and the observed 

activities indicated appropriate test control measures. 

Since no pipe break was simulated, ADS stays closed. 

Throughout the test the operations personnel demonstrated 

3.6.2 SIET SPES-2 

3.6.2.1 Test Matrix 

The test matrix for SPES-2 was specified by W. 

approved by the Test Engineer and the Test Requester. The procedures state 

the test objectives, specify initial conditions with acceptable tolerances, 

and specify critical instruments that must be operable. The procedure 

contains a check list with signature blocks for the key steps. 

procedure also includes test termination criteria, although in practice, a 

test may be continued longer. A day of test report is issued normally within 

a day of the conduct of the test. W reviews the test outcome to determine 

whether the test was valid or must 6e repeated. Three tests have had to have 

been rerun. The first matrix test was repeated due to AP600 design changes, 

the pressure balance line break (test 7) was repeated due to a valve being 

inadvertently left closed at the start of the test, and the main steam line 

break (test 12) was repeated due to a leak that developed. Dispositions for 

these tests were located in the respective design record files. 

All the experiments are run with initial conditions representative of normal 

full power conditions in the AP600. 

variations in initial conditions might affect the results. Pre-test 

predictions are performed by Ansaldo for W using the RELAPS code. 

calculated results are used to help determine how to operate SPES-2 in order 

to best match the target initial conditions. The calculations are available 

during the test and can be checked against the experimental data as the test 

proceeds. 

witnessed on October 15, 1994. 

The test procedures are 

The test 

No tests were planned to explore how 

The 

This was done during the one inch cold leg break test that was 

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88 
I 3.6.2.2 Test results and Data Processing 

Similar to the arrangement between 4? and OSU, SIET is not responsible for data 

analysis, which is performed instead by W. In addition to quick look reports 

for each experiment, y plans to issue: a Final Test Report that describes the 

experimental data (March 1995); a Data Analysis Report that compares the code 

with the experimental data (May 1995); and a Code Validation Report that 

support the application of the code to AP600 analysis (September 1995). 

The team reviewed the part of the DRF called "Approvals." Clear evidence was 

found of close communication between SIET and W, and that 

approval authority for test procedures for a1 1 preoperational and matrix 

tests. Final acceptance of completed tests also is H's responsi bil ity, but 

there was no indication in the SIET DRF of final acceptance. W stated that 

this information should be included in the official DRF in Monroeville. 

Documentation of tests in the SIET DRF includes a DOT report, indicating 

pertinent information about the test, and other documentation, such as signedoff procedures check1 ist; instrumentation 1 isting; and pre-test acceptance 

criteria. According to 8's test engineer, SIET is responsible for completing 

DOT reports only for those tests that are initiated (although they may not be 

completed). Any test for which the transient is initiated is assigned a 

unique number. For tests that are aborted due to a system failure (e.g., 

leaks, power problems, malfunctioning equipment) prior to initiating the test, 

no unique number is assigned, and no DOT reports are required. However, any 

problems, maintenance to the facility, ana so forth, occurring since the 

previous numbered test, are supposed to be recorded on the DOT report. 

addition, for tests that are initiated but not completed, or for tests that 

are found on post-test examination not to meet acceptance criteria, test 

deviation reports are required by QA procedures. 

had direct 

In 

The inspector noted some inconsistency in documentation of completed tests in 

the DRF. 

after test initiation, was not always documented and deviations also were not 

always documented on the DOT report or subsequently in the DRF. 

"deviations" are indicated on the DOT reports as "unexpected events" or 

"observations." While most problems were documented somewhere in the DRF, it 

appeared that the procedures for indicating deviations and documenting their 

resolution were unclear and the practice was inconsistent. The following are 

examples of this practice. 

Indication of the reasons for aborting tests, either prior to or 

Some 

Test SO0605 (volume on DVI line breaks): Stage 4 ADS valves failed to 

open automatically as specified. 

seconds later than the automatic signal. This was not documented as a 

"deviation," but rather as an "unexpected event." 

They were opened manually about 90 

Test SO0401 (volume on cold leg breaks): Stage 4 ADS valves failed to 

open automatically as specified. 

minutes after the automatic signal. 

"deviation," but rather as an "observation." 

Test SO0807 (volume on pressure balance line breaks): 

repeated because a manual valve on the break line was inadvertently left 

They were opened manually about 4 

This was not documented as a 

This test was 

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89 
in the closed position during the beginning of the test. 

the test began, anomalous behavior was observed, and the closed valve was 

determined to be the cause. 

continued to completion, but it was determined (apparently by y Test 

Engineering) that the closed valve was sufficient cause to rerun the 

test. 

to be invalid, and the closed valve is not indicated as a "deviation," 

but rather as an "unexpected event." 

Shortly after 

The valve was opened manually and the test 

The Day of Test report does not indicate why the test was declared 

The above examples are identified as part of Unresolved Item 94-01-03 and the 

NRC will review the W design record files in Monroeville during a future 

inspection to determine if jt! appropriately evaluated and dispositioned the 

test results. 

3.6.2.3 Witness of Matrix Test 

On October 15, 1994, conduct of a one inch cold leg break test was observed. 

This test was a variation on Matrix Test 1, and was intended to evaluate the 

effect of PRHR cooling. The latter test included full (three tubes) PRHR 

cooling while the former included reduced PRHR capacity (one tube). The use 

of three tubes is representative of "best estimate" plant conditions while one 

tube is used because it represents approximately the scaled PRHR heat transfer 

assumed by Pre-test 

calculations were performed by Ansaldo for three cases of heat loss 

compensation: 150 kW, 100 kW, and 50 kW. Examination of the results 

indicated that overall, the 150 kW compensation appeared to be the most 

prototypic, so the experiment was run accordingly. 

Ansaldo RELAP code predictions with the test data was very good for the data 

channels observed. 

Ansa1 do pre-test cal cul ations. 

in its AP600 SSAR analyses of most design basis events. 

The agreement of the 

The design record file includes a folder with all the 

SIET delineated the functions of test engineer, test conduct, and data 

analysis and is responsible for qualifying the data. Attaining initial 

conditions was complicated by the need to supply steam to the feedwater 

heaters from the steam line, since the adjacent fossil plant, which normally 

supplies steam to SPES-2 was out of service. The use of a different steam 

supply and procedures was properly executed/managed and the experiment was 

successfully performed. 

Through various discussions and observations, the subject personnel were 

considered very competent and the test was performed smoothly and 

professionally. 

the observed activities indicated the use of appropriate test control 

measures. 

No discrepancies were identified by the inspection team and 

3.7 Corrective Actions 

3.7.1 OSU APEX 

The procedural controls for the corrective action program at OSU are described 

in Section 13.0 of the site PQP. 

ident i f i cat i on and resol uti on of conditions adverse to qual i ty are del i neated 

Specific requirements regarding the 

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in Procedure DP-15.3, Revision 3, "Quality Performance Feedback." This 

document establishes the provisions for the control of non-conforming i tems 

including the description of functional/organizational responsibilities and 

identification of the management process for assessing appropriate resolution 

of discrepancies. 

In order to evaluate the effectiveness of the corrective action program at OSU 

the inspection team examined a selected sample of deviations identified during 

the design, construction, and operation of testing activities. As a result of 

these reviews it was generally concluded that appropriate provisions had been 

establ ished and documented to assure that conditions adverse to qual i ty were 

properly identified and corrected. 

process controls had been implemented for the documentation and reporting of 

significant adverse conditions to appropriate levels of management. 

It was also determined that appropriate 

3.7.2 SIET SPES-2 

SIET's corrective action program is described in Section 16 of their Quality 

Manual. The inspection team evaluated the pertinent portions of the Quality 

Manual which addressed nonconforming equipment and test samples, technical 

concerns, and the corrective action processes. Additionally, the applicable 

sections of Procedure 00003-QQ involving nonconforming instruments and the 

relevant aspects of Procedure 00007-QQ associated with internal audit findings 

were examined. 

Collectively the administrative provisions of these procedures combined with 

the training received by department personnel involving test deviation 

documentation indicated that appropriate provisions had been established for 

the prompt identification of conditions adverse to quality. Relative to the 

identification of root cause determination and corrective actions for 

significant conditions adverse to quality, the inspection team reviewed the 

results of Quality Status Reports 00269-VV-93 Revision 0 and 00333-VV-94 

Revision 0. 

Based on the results of these reviews it was generally concluded that 

appropriate provisions had been established for the documentation of 

conditions adverse to quality and that corrective actions had been initiated 

by appropriate levels of management regarding significant conditions adverse 

to quality. However, one potential area of concern was identified involving 

deviation reports which did not properly establish the basis for resolving 

discrepancies in test parameters. 

Section 3.6.2.2 of this report. 

This issue is discussed in detail in 

3.8 Qual i tv Assurance Records 

3.8.1 OSU APEX 

Quality Assurance records are prescribed in Section 14.0 of the PQP and 

describes typical qual i ty records associated with the Long Term Cool ing Test 

project. Additionally, requirements involving the retention of quality 

records at the test site pending transfer to the AP600 Control File are 

delineated in the PQP. The inspection team reviewed OSU's implementing 

-24- 

91 
procedures A-01, Rev. 2, "Testing Administration" and A-03, Rev. 0, "Control 

of Test Data" to determine the adequacy of program controls. 

Based on the inspection teams reviews within this area, it was generally 

determined that the inprocess records associated with test activities were 

properly stored and control led and were easily retrievable. Procedure A-03 

contained detai 1 ed instructions re1 ated to the development and control of 

electronically recorded test data and provisions for the proper identification 

and maintenance of documenting evidence of activities affecting quality had 

been establ i shed. 

- 

3.8.2 SIET SPES-2 

Qual ity assurance records requirements associated with the AP600 integral 

system test program at the SPES-2 facility were instituted for the collection, 

identification, and preservation of test records pending their transmittal to - W. The administrative controls associated with these activities are 

delineated in Section 12.0 of the Quality Manual and Section 4.0 of Procedure 

00006-QQ. 

and determined that appropriate provisions had been developed for the 

assimilation of documents that had been designated to become test records. 

Specifically, provisions had been established in SIET's document plan for the 

proper maintenance of test related records which identified the document 

title, identification number, revision status, initiating organization, 

approval authority, and effective issue date. Additionally, the document plan 

properly specified the requirements for external control and distribution of 

project documents. 

The inspection team examined the referenced administrative controls 

In order to evaluate the implementation of SIET's document control plan the 

inspection team reviewed a selected sample of QA documents including test 

reports, instrumentation calibration reports, plant log reports, and the 

facility description. The results of these reviews indicated that documents 

which were designated as QA records were legible, accurate and complete, and 

that they were properly indexed and readily retrievable. 

were identified and it was determined that the QA records control process was 

well established and properly implemented. 

No discrepancies 

3.9 

3.9 

osu 

fac 

the 

- wc 

Audits 

1 OSU APEX 

performed no internal audits of the test activities at the APEX test 

lity. 

1994 

rporate Projects QA audit at OSU. 

This function was conducted by 41 for OSU APEX. The team reviewed 

readiness review activities and reports as well as the August 1994 

The team concluded that it appeared 

that conducted an appropriate QA overview sight of OSU activities at the 

APEX test facility. 

OSU stated that no external supplier audits were performed of equipment 

suppliers since all equipment was purchased as commercial grade from known 

suppliers. This area will be reviewed further at W Monroeville. 

-25- 

92 
3.9.2 SIET SPES-2 

The team reviewed Section 17, "Internal Quality System Audits," o SIET'S 

Qual i ty Manual , 00001-QQ, and implementing Procedure OOO07-QQy "Internal 

Audits Procedure." The team reviewed the lead auditor qualification and 

certification records for the SIET QA Manager and his staff, and all 

documentat ion appeared current. The team a1 so reviewed re1 ated internal audit 

documentation such as: annual audit plans, audit checklists, audit reports and 

responses to audit report findings. The team veri f ied that appropriate 

corrective action was performed by the responsible personnel for audit report 

findings. 

External supplier audits are not within the scope of the testing contract with 

W for SPES-2. 

was procured from reputable, known Italian suppliers and accepted for use with 

an in house receipt inspection. 

reputable, internationally known suppliers of instrumentation (e.g. Honeywell) 

who are used as suppliers if they provide the instrumentation with calibration 

certificates. SIET a1 so purchases thermocouples from local suppl iers who 

maintain an IS0 9000 QA program based on this certification. 

The team concluded that SIET was implementing an appropriate internal audit 

program for SPES-2 test activities. 

SIET organization was not examined. 

SIET stated that purchased equipment (such as pipe and valves) 

SIET maintains a list of supposedly 

The conduct of audits external to the 

4 PERSONNEL CONTACTED 

Oreson State Universitv (OSU) APEX 

* Jose N. Reyes, Program Manager, OSU * John Groome, Facility Operating Manager, OSU * * 

* Louis Lau, Senior Engineer, * * 

Westinghouse Exit Heeting Participants on Conference Call 

Moshe A. Mahlab, OSU Test Project Manager, 

Carl Dumsday, OSU Test Manager, W 

Bob Tupper, AP600 Project Engineer, !j 

Kenneth A. Kloes, Quality Assurance Engineer, 

* 

* 

* 

Brian McIntyre, AP600 Licensing Manager 

Eugene Piplica, AP600 Test Manager 

David A1 sing, AP600 Qual i ty Assurance Manager . * John Butler, AP600 Licensing Engineer 

Societa' Informazioni EsDerienze Termoidrauliche (SIET) SPES-2 

4 Carlo Medich, Technical Manager, SIET 

4 Gustavo Cattadori, Head of Component Qualification Division, SIET 

4 Marina Bacchiani , Nuclear Reactor Division, SIET 

+ 

4 Marco Rigamonti, Test Engineer, SIET 

Alberto Musa, Quality Assurance Manager, SIET 

-26- 

93 
-27- 

94 

+ Eugene J. Pip1 ica, AP600 Test Manager, W 

+ Larry E. Conway, AP600 Test Engineer, W 

+ Kenneth A. Kloes, Quality Assurance Engineer, y 

GE Nuclear Enerqv 

+ Paul F. Billig, SBWR Test Engineer 

Nuclear Requlatorv Commission 

* + Richard P. McIntyre, Team Leader, Special Inspection Branch (TSIB) * + Uldis Potapovs, TSIB 

* + Robert M. Latta, Quality Assurance and Maintenance Branch (TQMB) 

+ Alan E. Levin, Reactor Systems Branch (SRXB) * George Thomas, SRXB 

* David E. Bessette RPSB/Research (RES) * Harold H. Scott, RPSB/RES 

+ David T. Tang, Standardization Project Directorate 

* Attended the exit meeting at OSU on September 1, 1994 

+ Attended the exit meeting at SIET on October 14, 1994 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON D C 2'?'53-%1 

February 6, 1995 

Wyle Laboratories 

Attn: Sherwyn Hyten, 

7800 Governors Drive 

Huntsville, AL 35807 

Senior Vice President 

SUBJECT : NRC INSPECTION NO. 99900902/95-02 

Dear Mr Hyten: 

This letter refers to the inspection led by Mr. K.R. Naidu, of this office on 

January 25-28, 1995. The inspection included a review of activities conducted 

at your facility in Huntsville, Alabama. At the conclusion of the inspection, 

the inspectors discussed their findings with Dr. C. Thibault and other members 

of your staff identified in Section 4 of the enclosed report. 

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

enclosed report. The inspectors reviewed the actions taken to correct a 

vi ol at i on and nonconformances i dent i f i ed i n Inspecti on Report 99900902/92-01, 

and evaluated selected areas of the program established and executed by Wyle 

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

Requlations (10 CFR Part 21) for reporting defects and noncompliance, and 

Appendix B to 10 CFR Part 50. The inspectors also observed selected seismic 

tests being performed to qualify 4.16-kV switchgear for installation at 

Pacific Gas and Electric Company's Diablo Canyon Power Plant. Within these 

areas, the inspection consisted of an examination of procedures and 

representative documents, interviews with personnel, and observation by the 

inspectors. 

Glithin the scope of this inspection, we found no instance where you failed to 

meet NRC requirements. Therefore, no response to this letter is required. 

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

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

Sincerely,- 

Robert M. Gallo, Chief 

Special Inspection Branch 

Division of Technical Support 

Office of Nuclear Reactor Regulation 

Docket No.: 99900902 

Enclosure: As stated 

cc: See Next Page 

95 
cc: Mr. Gregory M. Rueger 

Nuclear Power Generation 

B 14A 

Pacific Gas 81 Electric Company 

77 Beale Street, Room 145 

P.O. Box 77000 

San Francisco, CA 94106 

Mr. D. R. Michaud 

Products Manager 

NTS 

533 Main Street 

Acton, MA 01720 

Mr. J. L. Bachmann 

Project Manager 

Power Distribution Services, Inc. 

9870 Cresent Drive 

West Chester, Ohio 45069 

96 
REPORT NO. : 

ORGANIZATION: 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY: 

INSPECTION DATES: 

LEAD INSPECTOR: 

OTHER INSPECTORS: 

REVIEWED BY: 

APPROVED BY: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF TECHNICAL SUPPORT 

99900902/95-02 

Wyle Laboratories 

7800 Governors Drive 

Hunt svi 1 1 e , A1 abama 35807 

C. Thibault, Director, 

Nuclear Engineering Services Division 

(205) 837-4411 ext. 255 

Third party qualification of equipment manufactured by 

others for use in nuclear power plants. 

January 25-28, 1995 

/? 

n4A (&/& t -?/&/kKamalakar R. Naidu. Team Leader Date 

Vendor Inspection Section (VIS) 

Special Inspection .Branch (TSIB) 

Stephen D. Alexander, Reactor Engineer, VIS/TSIB 

Rogelio Mendez, Reactor Engineer, VIS/TSIB 

/t 

TSIB 

Gregory c.&walina, Chief, VIS Date 

TSIB 

Enclosure 97 
1 INSPECTION SUMMARY 

During this inspection, the inspectors reviewed the actions taken by Wyle to 

correct one violation and three nonconformances identified in Inspection 

Report 92-01, and found them acceptable. The inspectors also observed the 1 

implementation of Wyle's quality assurance program in selected areas during 

the seismic qualification tests of 4.16-kV circuit breakers and drawout 

cubicles supplied by National Technical Systems (NTS), Acton, Massachusetts. 

Pacific Gas and Electric Company (PG&E) issued a purchase order to NTS to 

supply the assembly for installation at its Diablo Canyon Power Plant to 

replace existing circuit breakers. Power Distribution Services (PDS), 

West Chester, Ohio, manufactured vertical lift drawout cubicles containing the 

4.16-kV circuit breakers (with Sulfur Hexaflouride (SF-6) interrupters) 

manufactured by Yasgawa Company, Ltd., of Japan. 

The inspection basis consisted of the following: 

0 

0 

0 

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

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

of Federal Requlations (10 CFR Part 50) 

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

Wyl e Qual i ty Assurance Manual and Procedures 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

2.1 Violation 99900902/92-01-01 (Closed): 

Contrary to Section 21.21 of 10 CFR Part 21, Wyle Scientific and Systems Group 

Laboratories (Wyle) failed to incorporate the new time limits and other new 

requirements that were specified in the revision of 10 CFR Part 21, that 

became effective October 1991. 

The inspectors reviewed Wyle's Quality Directive (QD) No. XIX-1, Revision A, 

"Reporting Defects and Noncompliance Per 10 CFR Part 21," and identified the 

need for some clarification. 

the conclusion of the inspection. The inspectors reviewed the final draft of 

QED No. XIX-1, Revision B, dated January 31, 1995, and determined it to be 

acceptabl e. 

The Wyle QA manager revised the procedure before 

2.2 Nonconformance 99900902/92-01-02 (Closed): 

Contrary to Criterion I of Appendix B to 10 CFR Part 50, Wyle failed to 

adequately delineate in writing the duties and authority of the QC Level I and 

Level I1 inspectors in its QD 11-2, "Engineering, Verification, Quality 

Assurance and Testing Personnel Qualification Program," dated August 1, 1988. 

As stated in its response letter of September 30, 1992, Wyle revised the 

following Nuclear Environmental Qualification (NEQ) procedures that delineate 

the responsibilities of Levels I and I1 Third Party Qualification (TPQ) 

inspectors: NEQ No. 403, Revision C, "TPQ Receiving and Component 

2 

98 
Homogeneity," dated May 6, 1994, and NEQ No. 409, Revision D, "The Wyle Third 

Party Qualification Process," dated May 6, 1994. Additionally, Wyle revised 

QD 11-2. Revision E of QD 11-2, dated November 15, 1993, delineates the 

general requirements of Levels I, 11, and I11 personnel who perform functions 

in the disciplines of engineering, verification, quality assurance and 

testing. 

The Wyle response stated that on May 22, 1992, personnel from TPQ Engineering, 

Inspection, and Quality Assurance and Testing were given training on the 

duties and authority of the Quality Control Levels I and I1 inspectors. Wyle 

staff informed the inspectors that the training was verbal. On the basis of 

interviews with members of Wyle staff and a review of some of their personal 

notes, the inspectors concluded that the reported training had been completed, 

but not documented. Before the conclusion of the inspection, the Director of 

Nuclear Engineering Services repeated and documented the training for his 

entire staffon the following subjects: 

0 Duties and authority of TPQ Verification Level I and I1 Inspectors. 

0 NEQ 404, Qualification by Similarity. 

e Timeliness of issuing a certificate of conformance. 

0 Requirement to issue a Nonconforming Material Report (NMR) when 

components are observed to be nonconforming and are rejected. 

2.3 Nonconformance 99900902/92-01-03 (Closed): 

Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Section 5, 

"Instructions, Procedures and Drawings," of Wyle's quality assurance manual 

(QAM), two examples were identified where Wyle's procedure did not ensure that 

activities affecting quality were adequately prescribed in QA program 

documents. 

During this inspection, following Wyle's formal written response to this 

nonconformance that had been forwarded to the NRC in a Wyle letter of 

September 30, 1992, the inspectors reviewed Wyle's implementation of 

corrective actions and preventive measures described in the letter. The 

inspectors found Wyl e's corrective act i ons and prevent i ve measures acceptabl e 

with the exception of documentation of retraining that was reported completed. 

As stated previously, before the conclusion of the inspection, the Director of 

Nuclear Engineering Services repeated and documented the training. 

2.4 Nonconformance 99900902/92-01-04 (Closed): 

Contrary to 10 CFR Part 50, Appendix B, Criterion V, Wyle personnel did not 

follow documented procedures. 

The inspectors reviewed the corrective action taken by Wyle as outlined in its 

formal written response to this nonconformance dated September 30, 1992, and 

found it acceptable. 

Three examples were cited. 

3 

99 
3 INSPECTION FINDINGS 

3.1 Entrance and Exit Meetinq 

During the entrance meeting on January 25, 1995, the NRC inspection team 

discussed the scope of inspection, the areas to be reviewed, and established 

the persons to contact within Wyle management and staff. 

meeting on January 27, 1995, the NRC inspection team summarized its findings 

and concerns. 

Section 4. 

During the exit 

Persons contacted during this inspection are identified in 

3.2 Review of Procurement Documents 

Pacific Gas and Electric (PG&E) issued purchase orders (PO) Nos. 068497 and 

068517 to National Technical Systems (NTS) for the supply of vertical lift 

drawout cubicles with 4.16-Kv, 350-MVA (million volt-ampere) circuit breakers 

manufactured by Yasgawa Company, Ltd., of Japan, for installation at the 

Diablo Canyon Power Plant (DCPP). The drawout cubicles are designed to 

replace the existing drawout cubicles that were made for GE AM-4.16, 250-MVA 

circuit breakers. PG&E selected NTS as the project manager to oversee the 

design, manufacture and qualification of the drawout cubicles on which the 

Yasgawa circuit breakers and other auxiliary equipment will be mounted for 

Class 1E applications. Power Distribution Services, Inc. (PDS), was selected 

to manufacture the drawout cubicles. NTS issued PO No. 37796, and supplement 

37796A, dated June 28 and November 28, 1994, respectively, to Wyle to subject 

three 4.16-Kv drawout cubicles containing Yasgawa 4.16-Kv circuit breakers to 

triaxial seismic multi-frequency random tests. The tests were designed to 

simulate plant specific seismic conditions in accordance with NTS Test 

Procedure No. 60431-95N-ST, "Triaxial Seismic Multi-frequency Random Test on 

1200/2000 Amp SF-6 Conversion Breaker for Use in the Diablo Canyon Site for 

Pacific Gas & Electric Company." The inspectors reviewed the NTS PO and its 

supplement, which were designated safety-related services, and determined that 

they required compliance with the following: 

A quality assurance program meeting the requirements of 10 CFR Part 50, 

Appendix B. 

Wyle's reporting requirements to meet 10 CFR Part 21; Wyle to notify NTS 

if a reportable condition is observed. 

e Requirements of MIL-STD-45662A for the calibration of measuring and test 

equipment. 

Wyle's responsibility for notifying NTS of any nonconforming items and 

actions taken to remedy defects, or suspected defects found in the 

switchgear during testing. 

The inspectors did not identify adverse findings in this area. 

4 

100 
3.3 Review of NTS Test Procedure 

The inspectors reviewed NTS’s Test Procedure No. 60431-95N-5T and determined 

that it contained the following: 

a. The test specimen was described as follows: 

The test specimen consisted of three GE stationary metal-clad 

cubicles similar to those that are installed at the DCPP. Each 

stationary cubicle was equipped with a PDS-fabricated, vertical 

lift drawout cubicle on which a Yasgawa 350-MVA interrupting 

capacity circuit breaker was mounted and suitably modified to 

replace the existing GE Magne-Blast 250-MVA circuit breaker. 

of the three 4.16-Kv circuit breakers in the drawout cubicles were 

rated for 1200 amperes (A), and one was rated for 2000 A 

continuous current. 

including a potential transformer, miscellaneous relays, terminal 

blocks, dummy loads in the stationary cubicles to simulate the 

switchgear installed at DCPP. 

Two 

PDS also mounted auxiliary equipment, 

b. The steps to seismically qualify the drawout cubicles and breakers 

were: 

A visual inspection and functional tests were performed to obtain 

baseline information. 

sequence addressed the various positions and operations of the 

circuit breaker during six operating basis earthquakes (OBEs) and 

one safe shutdown earthquake (SSE). Full scale shock spectrums 

for the test runs and locations were specified: i.e., vertical, 

side to side, and front to back for OBE, SSE, and fragility 

1 eve1 s . 

The seismic test procedure and test 

The inspectors did not identify adverse findings in this area. 

3.4 Review of Wvle Test Documentation 

The inspectors reviewed the test control record for the qualification of 

DCPP’s 4.16-kV drawout cubicles and breakers prepared by Wyl e and determined 

that it reflected the NTS test procedure requirements and included the 

following: 

1. Receipt inspection of the 4.16-kV switchgear shipped from PDS. 

This step had been ompleted. Wyle’s receipt inspection 

documented minor sh pping damage to the side of one of the 

stat i onary cubi cl es 

2. Test Set-Up and Equ 

3. Resonance search. 

4. Qual i f i cati on Level 

OBE and SSE. 

pment Sheet. 

(Random Mu1 ti-Frequency (RMF)) tests for the 

5 

101 
5. Fragility Level tests. 

6. Post Test Review (Data Review). 

Wyle QA imposed hold points on Items 1, 2, and 6. A hold point requires a QA 

representative to witness and acknowledge the satisfactory completion of the 

line item. The inspectors observed that a QA inspector signed the hold point 

for item 1, concurring with the identification of the shipping damage and 

completion of the receipt inspection. 

The inspectors did not identify adverse findings in this area. 

3.5 Observation of Seismic Qualification Tests 

The inspectors noted that 42 response accelerometers were attached to the test 

specimen (stationary and drawout cubicles with breakers) at 14 locations. 

Records indicated that NTS had performed operability tests on the various 

auxiliary equipment mounted on the switchgear test specimen to establish the 

baseline conditions. PDS had also manufactured a control panel to operate the 

circuit breakers from a remote location during the tests. 

Wyle welded a structural steel frame to the triaxial shaker table and the test 

specimen was welded to the frame to simulate the switchgear mounting at DCPP. 

Contact chatter detection devices to monitor the chatter of the auxiliary 

contacts of the breaker and selected relays were connected. 

The inspectors observed a total of seven tests conducted on the test specimen. 

Tests 1, 2, and 3 consisted of low level (approximately 0.2g) sine sweeps from 

1 to 100 Hz and back to 1 Hz in each orthogonal axis to establish major 

resonances of the test specimen. Tests 4, 5, 6, and 7 consisted of multifrequency random tests to simulate an OBE. The inspectors observed that NTS 

personnel were documenting the anomalies and that PG&E personnel will 

participate in their evaluation. The inspectors observed that the tests were 

performed in accordance with the approved procedures and did not identify any 

adverse findings in this area. 

3.6 Review of PDS Instrument Calibration 

The inspectors reviewed the calibration data of the instruments used during 

the seismic testing of the breakers and found no problems with instrumentation 

involved directly with the breaker qualification tests; however, the 

inspectors noted potential problems with the use of measuring and test 

equipment (M&TE) by NTS/PDS. 

On January 25, 1995, the inspectors observed NTS/PDS using a Multi-Amp test 

instrument with a January 11, 1995, calibration due date. The Multi-Amp 

instrument was used for verifying protective relay baseline data (obtained at 

PDS) prior to and upon completion of the breaker seismic tests. The NTS/PDS 

representatives informed the inspectors that they had discussed this issue 

with PG&E and obtained their approval for using M&TE with an expired 

calibration date because it would not affect the results of the seismic 

6 

102 
testing of the breakers. The NTS/PDS representatives further stated that the 

relay data were for information only and not directly related to the seismic 

testing of the breakers. However, the inspectors were concerned because NTS 

testing procedure No. 60431-95N-ST stated that the relays were to be 

calibrated to specified settings. The inspectors also noted that this MultiAmp test instrument was not the same as the instrument used for taking the 

baseline test data. 

In addition, the inspectors reviewed the calibration data of 18 protective 

relays taken using the Multi-Amp set with the expired calibration and observed 

that 5 of the relays were outside the tolerances specified in the test 

procedure. All 18 protective relays had been originally set to within the 

tolerances specified in the test procedure. The inspectors could not 

determine whether the Multi-Amp unit with the expired calibration date used at 

the test site caused the relays to appear to be out of calibration, whether 

there was a drift in the relay set points, or whether the Multi-Amp test 

equipment used for the baseline data was out of calibration. 

The relay calibration issue does not necessarily ivalidate the seismic testing 

of the circuit breaker and drawout cubicle design; although, excessive relay 

contact chatter experienced during the testing severly hampered the qualification effort. Nevertheless, this issue will be re-examined if future 

activities by NTS or PDS utilize the relay information. 

4 PERSONS CONTACTED 

Wvle Laboratories 

T. Brewington 

D. Compton 

G. Duthi 

*t R. Francis 

S.D. Hyten *t J.N. Matzkiw 

R.B. Pinkerton 

*t E.R. Schum 

t T. Stinson 

D. Smith 

P. Stover 

*t C. Thibault 

*t R.G. Thomas 

P. Wadsworth 

Director of Nuclear Services, International 

Test Technician 

Test Technician 

Manager, Electrical Equipment Retrofit 

Senior Vice President 

Manager, TPQ Components 8. Production 

Senior Test Supervisor 

Manager, EQ/TPQ Assemblies 

Director of Administration 

Manager, Nuclear Qualification Testing 

Test Technician 

Director, Nuclear Engineering Services 

Manager, Quality Assurance 

Test Technician 

* Attended the entrance meeting on 1/25/95 

t Attended the exit meeting on 1/27/95 

National Technical Systems 

M. Lilly 

D.R. Michaud 

Quality Assurance 

Products Manager 

7 

103 
Pacific Gas & Electric Company 

M. Basu 

S.J. Float 

M.R. Khan 

T.W. Packy 

B. Supremo 

Senior Electrical Engineer 

Electrical Engineer 

Consulting Mechanical Engineer 

Lead Auditor 

El ectri cal Engineer 

Power Distribution Services, Inc. 

J. L. Bachmann Project Manager 

J. E. Bachmann Quality Assurance 

B. Kogin Project Engineer 

D.R. Robling Project Engineer 

8 

104 
Sel ected Generic Correspondence on the Adequacy of 

Vendor Audits and the Quality of Vendor Products 

Identifier Title 

Information Notice 95-02 Problems with General Electric CR2940 Contact 

Blocks in Medium-Voltage Circuit Breakers 

Information Not 

Information Not 

Informat on Not 

ce 95-12 Potenti a1 ly Nonconforming Fasteners Suppl i ed by 

A&G Engineering 11, Inc. 

ce 95-19 Failure of Reactor Trip Breaker to Open Because 

of Cutoff Switch Material Lodged in the Trip 

Latch Mechanism 

ce 95-20 Failures in Rosemount Pressure Transmitters Due 

to Hydrogen Permeation Into the Sensor Cell 

105 
CORRESPONDENCE RELATED TO VENDOR ISSUES 

106 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555.0001 

January 19, 1995 

Mr. Joseph M. Tate 

General Manager 

ABB Service Company 

Regional Service Cent r 

5311 Conmerce Parkway West 

Cleveland, OH 44130 

SUBJECT: ABB RESPONSE TO A NOTICE OF VIOLATION (NOV) AND NONCONFORMANCE (NON) 

CONTAINED IN NRC INSPECTION REPORT 99901281/94-01 

Dear Mr. Tate: 

This letter addresses the ABB Service Inc. (ABB), January 5 and 13, 1O95, 

response letters to the U.S. Nuclear Regulatory Commission (NRC) reg,, ding the 

Notices of Viol ati on (NOV) , Nonconformance (NON) , and unresol ved i tems i ssued 

to ABB on December 5, 1994, with NRC Inspection Report 99901281/94-01. 

The NRC review of your January 5, 1995, NOV response and discussions with your 

Mr. John Webb on January 11, 1995, found that the ABB's basis for disputing 

Violation 94-01-01 appeared reasonable and the NRC staff has withdrawn the 

vi o l a t i on. 

The NRC review of your January 5 and 13, 1995, responses to the NON and 

Unresolved Items found that ABB's statements were responsive to our concerns. 

We will review the implementation of your corrective actions for the 

Nonconformances and Unresolved Items during a future NRC inspection to 

determine the adequacy of your corrective actions and ensure that full 

compliance has been achieved and will be maintained. 

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

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

staff contact Mr. Kamal Naidu (301) 504-2980 if you have any comments or 

questions. 

Please have your 

Si ncerel y , 

Robert M Gallo, Chief 

Special Inspection Branch 

Division of Technical Support 

Office of Nuclear Reactor Regulation 

Docket No. 99901281 

cc: See attached list 

107 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHISG’2K 7 C 20555-0001 

February 7, 1995 

Mr. Richard B. Kelley, Vice President 

Power Generation Services 

2146-6 F1 intstone Drive 

Tucker, Georgia 30084-3101 

SUBJECT: REQUEST FOR INFORMATION: 

FOR EXPORT TRADING FIRMS 

10 CFR PART 50, APPENDIX 6, APPLICABILITY 

Dear Mr. Kel 1 ey: 

By letter dated November 22, 1994, you requested the U. S. Nuclear Regulatory 

Conmission (NRC) to provide a ruling, accompanied by comments, precautions, 

procedures, and references, for the applicability of Appendix B to Title 10 of 

the Code of Federal Requlations (10 CFR) to a U. S. commercial firm, 

repr-xnting a U. S. manufacturer, where the commercial firm is an efport 

traaing company involved with only commercial aspects of transactions 1 icensed 

for export. 

1. 

In your letter, you asked the following specific questions. 

Can the comnercial firm solicit safety related orders in the name of the 

manufacturer? 

2. Can the comnercial firm accept orders from customers issued in the name of 

the manufacturer and re-issue the same order to the manufacturer 

unaltered, except for financial terms of sale? 

3. Can the comnercial firm make export shipping arrangements for safety 

related goods shipped directly from the manufacturer to the customer’s 

import broker? 

On December 20, 1994, Messrs. Larry Campbell, Robert Gramm, and Uldis Potapovs 

of this office contacted you to obtain additional information and the specific 

circumstances relating to your questions. The following example was 

discussed: A nuclear power plant owner in Spain places an order for ASME 

Section 111, Class I, nuclear grade pipe (and invokes the requirements of 

Appendix B to 10 CFR Part 50) with a commercial firm in Spain. This 

commercial firm then places an order with a U. S. import/export commercial 

firm. The U. S. firm then places the order (invoking the requirements of 

Appendix B to 10 CFR Part 50 and the ASME Section 111 Code) with a U. S. 

manufacturer accredited by ASME for the manufacture and supply of nuclear 

grade piping. During this conversation, you raised the question: When a 

comnercial import/export firm, located in another country and representing a 

foreign comnercial nuclear power plant owner, places an order with a 

commercial import/export firm in the U. S. and the order invokes the 

requirements of Appendix B to 10 CFR Part 50 and the ASME Section 111 Code, 

what NRC requirements and regulations apply to this U. S. commercial firm? 

108 
R. Kelley -2- 

Although contractually the U. S. commercial firm may have an obligation to 

meet the quality and technical requirements invoked by the foreign comnercial 

firm, the manufacturing and supply of safety related items for use by an owner 

of a nuclear power plant in another country is not within the jurisdiction of 

the NRC. Potential enforcement actions by the NRC, resulting from 

noncompliance to Appendix 8 to 10 CFR Part 50 or the ASME Section I11 Code, 

would not be appropriate for items supplied under the circumstances described 

in your letter and supplemented in your December 20, 1994, conversation with 

the NRC staff. 

Should you have any further questions regarding this matter, please feel free 

to contact Mr. Larry L. Campbell at (301) 504-2976 or Mr. Robert Gramm at 

(301) 504-1010. 

Sincerely, 

Suzanne C. Black, Chief 

Qual i ty Assurance and Maintenance Branch 

Division of Technical Support 

Office of Nuclear Reactor Regulation 
IRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER 

!%9) (Asslend bv NRC. Add Vol Su Rw., RCM 1102, ad Addendum Numbers, If hy.?' 

NU R EG-0040 201,3202 BIBLIOGRAPHIC DATA SHEET 

Vol. 19, No. 1 (see instructions on the reverse) 

, TITLE AND SUBTITLE 

Licensee Contractor and Vendor Inspection 

Status Report 

Quarterly Report 

January - March 1995 

DATE REPORT PUBLISHED 

MONTH 

Y 4. FIN OR GRANT NUMBER 

. AUTHOR(S) 6. TYPE OF REPORT 

Quarterly 

January - March 1995 

7. PERIOD COVERED Nnclusrwe Dates) 

. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC,provide Division, Officeor Region, U.S Nuclear Regulamv Commission, andmai1ingaddress;ifcontracmr. provide 

name and mailing added 

Division of Technical Support 

Office of Nuclear Reactor Regulation 

U. S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC. wpe "Same as above': ifcontracmr, provide NRC Division, Office or Region, U.S Nuclear Regulatory Commission, 

and mailing addred 

Same as above 

0. SUPPLEMENTARY NOTES 

1. ABSTRACT (200 words or Jess) 

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 1995. 

12. KEY W0RDSlDESCR:PTORS (Listwordsorphrarer that willarristresearchersin locating thereport.) 

Vendor Inspection 

13. AVAILABILITY STATEMENT 

Unl i mi ted 14. SECURITY CLASSIFICATION 

(This Page) 

Unclassified 

Unclassified 

(This Report) 

15. NUMBER OF PAGES 

16. PRICE 

IRC FORM 335 (289)