US, Nuclear Regulatory Commission 

Office of Nuclear Reactor Regulation 

OSTI 
AVAfLABlLlTY NOTICE 

Availability of Reference Materials Cited in NRC Publications 

Most documents cited in NRC publications will be available from one of the following sources: 

1. The NRC Public Document Room, 2120 L Street, NW., Lower Level, Washington, DC 

20555-0001 

2. The Superintendent of Documents, U.S. Government Printing Office, P. 0. Box 37082, 

Washington, DC 20402:9328 

3. 

Although the listing that follows represents the majority of documents cited in NRC publications, it is not intended to be exhaustive. 

The National Technical Information Service, Springfield, VA 221 61 -0002 

Referenced documents available for inspection and copying for a fee from the NRC Public 

Document Room include NRC correspondence and internal NRC memoranda; NRC bulletins, 

circulars, information notices, inspection and investigation notices: licensee event reports: 

vendor reports and correspondence; Commission papers: and applicant and licensee documents and correspondence. 

The following documents in the NUREG series are available for purchase from the Government 

Printing Office: formal NRC staff and contractor reports, NRC-sponsored conference proceedings, international agreement reports, grantee reports, and NRC booklets and brochures. Also available are regulatory guides, NRC regulations in the Code of Federal Regulations, and Nuclear Regulatory Commission Issuances. 

Documents available from the National Technical Information Service include NUREG-series 

reports and technical reports prepared by other Federal agencies and reports prepared by the 

Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission. 

Documents available from public and special technical libraries include all open literature 

items, such as books, journal articles, and transactions. Federal Register notices, Federal 

and State legislation, and congressional reports can usually be obtained from these libraries. 

Documents such as theses, dissertations, foreign reports and translations, and non-NRC conference proceedings are available for purchase from the organization sponsoring the publication cited. 

Single copies of NRC draft reports are available free, to the extent of supply, upon written 

request to the Office of Administration, Distribution and Mail Services Section, U.S. Nuclear 

Regulatory Commission, Washington, DC 20555-0001 . 

Copies of industry codes and standards used in a substantive manner in the NRC regulatory 

process are maintained at the NRC Library, Two White Flint North, 1 1545 Rockville Pike, Rockville, MD 20852-2738, for use by the public. Codes and standards are usually copyrighted 

and may be purchased from the originating organization or, if they are American National 

Standards, from the American National Standards Institute, 1430 Broadway, New York, NY 

1001 8-3308. 

A year's subscription of this report consists of four quarterly issues. 
NUREG-0040 

VOl. 20, No. 2 

Licensee Contractor 

and Vendor Inspection 

Status Report 

Quarterly Report 

April - June 1996 

Manuscript Completed: August 1996 

Date Published: August 1996 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 
. . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . 
DISCLAIMER 

Portions of this document may be illegible 

in electronic image products. Images are 

produced from the best available original 

document. 

DISCLAIMER 

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

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

thereof, nor any of their employees, makes any warranty, express or implied, or 

assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents 

that its use would not infringe privately owned rights. Reference herein to any spccific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise docs not necessarily constitute or imply its endorsement, mommendation, or favoring by the United States Government or any agency thereof. 

The views and opinions of authors expressed herein do not necessarily state or 

reflect those of the United States Government or any agency thereof. 

ABSTRACT 

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

Special Inspection Branch, Vendor Inspection Section, that have been 

distributed to the inspected organizations during the period from April 1996 

through June 1996, and also includes a report, issued in March of 1996, but 

not included in NUREG-0040, Vol . 20, No. 1. 

iii 

CONTENTS 

- PAGE 

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

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

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

Algor, Inc. 

Pittsburgh, PA 

AMP Products, Inc. 

Jacobus, PA 

Cegelec Automation, Inc. 

Pomona, CA 

Factory Mutual Engineering Association 

Norwood, MA 

GE Nuclear Energy 

San Jose, CA 

General Electric Nuclear Energy 

Nucl ear Energy Production 

Wilmington, NC 

PROMAlEC, Inc. 

Cypress, TX 

(99901294/96-01) ......... 

1 

2 

(99901295/96-01) ......... 11 

(99900734/96-01) ......... 20 

(99900603/96-01) ......... 35 

(99900403/96-01) ......... 50 

(99900003/95-01) ......... 74 

(99901 292/95-01) ......... 163 

Select Generic Correspondence on the Adequacy of Vendor .................. 214 

Audits and t,he Quality of Vendor Products 

V 

INTRODUCTION 

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

licensing and inspection program is that licensees are responsible for the 

proper construction and safe and efficient operation of their nuclear power 

plants. The Federal government and nuclear industry have establ ished a system 

for the inspection of commercial nuclear facilities to provide for multiple 

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

participates in a quality verification process in compliance with requirements 

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

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

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

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

industry within the framework of quality verification programs. 

The 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 a1 so have 

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

codes, and standards. 

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

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

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

laboratories performing equipment qualification tests, and holders of NRC 

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

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

probl ems are determined and appropriate correct i ve act i 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 i censees and vendors. 

These 

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

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

and review of equipment problems found during operation and their corrective 

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

the inspected organization is required to take appropriate corrective action 

and to institute preventive measures to precl ude recurrence. When gener i c 

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

through vendor reporting or by NRC generic correspondence such as information 

notices and bull et i ns . 

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 i nvol vi ng vendor i ssues . 

Each vendor inspection 

viii 
INSPECTION REPORTS 

1 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 205550001 

May 3, 1996 

Mr. Michael L. Bussler, President 

Algor, Inc. 

150 Beta Orive 

Pittsburgh, PA 15238-2932 

and Chief Executive Officer 

SUBJECT: NRC INSPECTION REPORT NO. 99901294/96-01 

Dear Mr. Bussler: 

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

(NRC) inspection of Algor, Inc., conducted by Robert L. Pettis, Jr. and Billy 

Rogers, of this office, on February 20 through 22, 1996. The NRC inspection 

team conducted an evaluation of the Algor quality assurance (QA) program and 

the implementation of that program as it relates to safety-related software 

supplied to the nuclear industry. 

The NRC inspection team reviewed documentation, procedures, and representative 

records, conducted interviews and held discussions with members of your staff. 

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

implementation of the Algor QA program failed to meet certain requirements of 

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

CFR Part 21. 

the areas examined. 

The enclosed inspection report contains a detailed discussion of 

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

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

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

discuss them with you. Thank you for your cooperation. 

Sincerely, n 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nucl ear Reactor Regul at i on 

Docket No.: 99901294 

Enclosures: 1. Notice of Violation 

2. Inspection Report No. 99901294/96-01 

2 
NOTICE OF VIOLATION 

Docket No. 99901294 

During an NRC inspection conducted at Algor, Inc., on February 20 through 23, 

1996, a violation of NRC requirements was identified. In accordance with the 

"General Statement of Policy and Procedure for NRC Enforcement Actions," 

NUREG-1600, the violation is listed below: 

Title 10 of the Code of Federal Requlations Part 21.21(b) states if the 

deviation or failure to comply is discovered by a supplier of basic 

components, or services associated with basic components, and the supplier 

determines that it does not have the capability to perform the evaluation 

to determine if a defect exists, then the supplier must inform the 

purchasers or affected licensees within five working days of this 

determination so that the purchasers or affected licensees may evaluate the 

deviation or failure to comply, pursuant to 21.21(a). 

Contrary to the above, Algor, Inc., did not inform NRC licensees of 

deviations within five working days of discovery when informing 

licensees of the deviations identified by the Algor, Inc., Quality 

Assurance Bulletins Nos. 68, 69, 71, 84, 100, 101, and 102. 

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

Pursuant to the provisions of 10 CFR 2.201, Algor, Inc., is hereby required to 

submit 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 Violation (Notice). 

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

violation: (1) the reason for the violation, or if contested, the basis for 

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

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

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

response may reference or include previous docketed correspondence, if the 

correspondence adequately addresses the required response. 

is shown, consideration will be given to extending the response time. 

This reply should be clearly 

Where good cause 

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

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

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

redaction. 

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

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

withholding the information from the public. 

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

Dated at R ckville, Maryland 

this 3"e day of May, 1996 

Enclosure 1 

3 
REPORT NO. : 

ORGAN I ZAT ION : 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

INSPECTION DATES: 

LEAD INSPECTOR: 

OTHER INSPECTORS : 

REVIEWED BY: 

APPROVED BY: 

U. S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF INSPECTION AND SUPPORT PROGRAMS 

99901294/96-01 

Michael L. Bussler, President 

and Chief Executive Officer 

Algor, Inc. 

150 Beta Drive 

Pittsburgh, PA 15238-2932 

Theresa Anania 

Director of Operations 

Suppliers of safety-related finite-element 

computer proprams. 

February 20 through 22, 1996 

/J&?L/+ii?%(~. 

Robert L. Pettis. Jr.. P.E. - Date 

Vendor Inspection Section (VIS) 

Special Inspection Branch (SIB) 

Division of Inspection and Support Programs (DISP) 

Office of Nuclear Reactor Regulation (NRR) 

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

//ALL 

~egoryfl.Awa1 i na, Chief 

Cwalina, Acting Chief 

Date 

Enclosure 2 

4 
1 SUMMARY OF INSPECTION FINDINGS 

During this inspection, the NRC inspection team evaluated the implementation 

of the Algor, Inc., quality assurance (QA) program related to the supply of 

safety-related finite-element computer programs to the nuclear industry. 

inspection was conducted to determine Algor’s compliance with the requirements 

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

(Appendix H) and the provisions of 10 CFR Part 21 (Part 21). 

team reviewed technical information, procedures and representative records, 

conducted interview and held discussions with members of Algor’s staff. 

The 

The inspection 

1.1 Violations 

1.1.1 Contrary to 10 CFR 21.21(b) which requires that if a deviation is 

discovered by a supplier of basic components and the supplier determines that 

it does not have the capability to perform the evaluation to determine if a 

defect exists, then the supplier must inform the purchasers or affected 

licensees within five working days of this determination so that the 

purchasers or affected licensees may evaluate the deviation, Algor, Inc., 

(Algor) did not inform NRC licensees of deviations within five working days of 

discovery. (Viol ation 96-01-01) 

1.1.2 Contrary to 10 CFR 21.6 which requires that entities subject to the 

regulations post 10 CFR Part 21, Section 206 of the Energy Reorganization Act 

of 1974, and the procedures adopted pursuant to 10 CFR Part 21, Algor had not 

posted Section 206 or the Algor 10 CFR Part 21 Procedure, QAPM Section 8, and 

the copy of 10 CFR Part 21 that Algor had posted was dated January 1988 which 

did not contain the numerous subsequent revisions contained in the later 

editions. (Non-Cited Viol ation) 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

This was the first inspection of Algor. 

3 INSPECTION FINDINGS AND OTHER COMMENTS 

3.1 Bac karound 

Algor, Incorporated, is a designer of various engineering fini te-element 

analysis software programs used both in nuclear and commercial applications. 

Algor’s product 1 ine includes both 1 inear and non-1 inear stress analysis 

programs designed to analyze various applications including dynamic, 

vibration, heat transfer and piping analysis. 

3.2 Entrance and Exit Meetinqs 

During the entrance meeting, held on February 20, 1996, the NRC inspection 

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

the inspection, and established organizational contacts. 

meeting, held on February 22, 1996, the inspection team summarized its 

findings with Algor management. Section 4 of this report lists the persons 

contacted during the inspection. 

During the exit 

2 

5 
3.3 10 CFR Part 21 Procedures and ImDlementation 

Section 10 CFR 21.6, "Posting Requirements," requires that entities subject to 

the regulations post 10 CFR Part 21, Section 206 of the Energy Reorganization 

Act of 1974, and the procedures adopted pursuant to 10 CFR Part 21. 

Quality Assurance Procedures Manual (QAPM) Section 8, "1OCFR21 - Error 

Reporting," Revision 4, dated February 21, 1995, Paragraph 8.6 "Posting," 

stated that 10 CFR Part 21 shall be prominently displayed on the employee 

bulletin board but did not address Section 206 or the Algor 10 CFR Part 21 

procedure, QAPM Section 8. 

The 

The inspectors reviewed Algor's posting and determined that Section 206 was 

absent and Algor's 10 CFR Part 21 Procedure was absent. 

of 10 CFR Part 21 that was posted was dated January 1988 and did not contain 

the numerous sibsequent revisions contained in the later editions. The 

inspectors provided Algor with current copies of 10 CFR Part 21 and Section 

206 and Algor indicated that they would correct their posting. The failure to 

meet the posting requirements specified in 10 CFR 21.6 constitutes a violation 

of minor significance and is being treated as a Non-Cited violation, 

consistent with Section IV of the NRC Eqforcement policy (NUREG-1600). 

In addition, the copy 

The inspectors reviewed the implementation of Algor's 10 CFR Part 21 program 

to determine compliance with 10 CFR Part 21. 

provided to NRC licensees indicated that deviations would occur as errors in 

the computer software. Algor had defined four error types in QAPM Section 7, 

"Qalipak Quality Assurance Package," Revision 4, dated February 21, 1995; 

Class A errors, Class B errors, Class C errors, and Class H errors. 

Discussion of Algor products 

Class C errors, defined as errors which produce results that may appear 

reasonable and correct, but are in fact erroneous; were difficult for the end 

user to detect because the program did not provide any clear indication that 

the results may be invalid; and provided no warning messages, error messages 

or other abnormalities despite responsible operation of the Algor product, 

relative to the error itself, which indicated that the results may be invalid. 

The inspectors concluded that Class C errors would constitute deviations, as 

defined by 10 CFR Part 21, and would therefore require evaluation by the 

supplier (Algor) or the user (NRC Licensee). 

would not evaluate Class C errors but would inform the user such that the user 

could perform the evaluation to determine whether the deviation was a defect 

or failure to comply. 

Algor had determined that it 

Algor had notified customers of Class C errors by providing them with Quality 

Assurance Bulletins (QAB) which described the error and provided customers 

with applicable information on software revisions which corrected the error or 

methods that users could follow to work around the error. 

review ,of QABs included a block of twenty, which were issued during 1995 and 

early 1996. The review determined that seven of the Class C error reports 

were provided to NRC licensees with greater than five working days between 

error classification (identification of the deviation) and providing the QAB 

to the licensee (informing the customer of the deviation). The length of time 

between error classification (identification of the deviation) and providing 

the QAB to the licensee (informing the customer of the deviation) ranged from 

3 

The inspectors 

6 
twenty-three to seventy-one days. 

issuance dates are listed as follows: 

The error classification dates and QAB 

QAB No. 68 error classified 4/11/95 - QAB issued 5/4/95 

QAB No. 69 error classified 4/11/95 - QAB issued 5/5/95 

QAB No. 71 error classified 4/11/95 - QAB issued 5/4/95 

QAB No. 84 error classified 6/1/95 - yAB issued 8/11/95 

QAB No. 100 error classified 6/16/95 - QAB issued 8/14/95 

QAB No. 101 error classified 6/16/95 - QAB issued 8/14/95 

QAB No. 102 error classified 6/16/95 - QAB issued 8/14/95 

The provisions of 10 CFR 21.21(b) require that if a deviation or failure to 

comply is discovered by a supplier of basic components and the supplier 

determines that it does not have the capability to perform the evaluation to 

determine if a defect exists, then the supplier must inform the purchasers or 

affected licensees within five working days of this determination so that the 

purchasers of affected 1 icensees may evaluate the deviation or failure to 

comply, pursuant to 21.21. 

Algor did not notify NRC licensee within five working days of identifying the 

deviations. This was identified ?I Vilation No. 96-01-01. 

Contrary to the requirements of 10 CFR 21.21(b), 

3.4 Review of OaliDak Quality Assurance Packaqe 

The inspectors reviewed QAPM Section 7, "Qal ipak Quality Assurance Package," 

Revision 4, dated February 21, 1995, which established the service to which 

Algor customers could subscribe to receive notification of errors which could 

produce significant, erroneous output data despite responsible operation of an 

Algor product. 

(1) Basic - a service which was available free, upon request, to all Algor 

customers. 

would provide a brief description of Class C errors and was to be issued 

monthly; and (2) Qalifax - a service in which Quality Assurance Bulletins 

(QAB) are issued to customers by telefax as soon as the QAB is completed. 

The Basic Qalipak Quality Assurance Package was discussed in QAPM Section 7, 

which stated that the Quality Summary Report would be provided to customers 

monthly, on request. Algor indicated that although QAPM Section 7 required 

the monthly issuance of the Qalipak Summary Report none had yet been issued. 

Algor indicated that they were in the process of developing the first issue of 

the Qalipak Summary Report, to be issued in the near future, and monthly 

thereafter. 

Algor offered two level s of the Qal ipak Subscription: 

The customer would receive a Qalipak Summary Report (QSR) which 

The Qalifax Quality Assurance Package was the method by which Algor notified 

NRC licensees of deviations (Class C errors) through the issuance of QABs. 

Algor provided Qalifax as a service sold in conjunction with the purchase of 

safety-related software. A Qalifax subscription was required to be in place 

(previously purchased and current) or purchased concurrently with safetyrelated software purchases. 

subscription in place which could support numerous safety-related software 

purchases. All QABs related to different software packages were provided to 

all Qalifax subscribers regardless of the software that the customer had 

purchased. 

A customer was only required to have one Qalifax 

4 

7 
The inspectors reviewed Algor's Qal ifax subscribers records and Qal ifax 

distribution records to verify that Algor had been providing notification of 

all Class C errors to all Qalifax subscribers including NRC licensees. 

Section 7, Paragraph 7.3, of the QAPM, "Qalipak Distribution" states that a 

Qalipak Transmittal Sheet would be included with each transmittal of a QAB. 

The Qalipak Transmittal Sheet was provided so that the customer could 

acknowledge receipt of the QAB. If the first Qalipak Transmittal Sheet was 

not received Algor would send a Second Notice and, if necessary, a Final 

Notice in conjunction with telephone calls, The inspectors reviewed the 

records for QABs Nos. 33 through 47, Class C errors, issued in September of 

1993. Algor's records showed that all Qal ifax subscribers, current in 

September 1993, had been notified and the records contained documentation of 

positive receipt acknowledgement from all Qal ifax subscribers (Qal ifax 

Transmittal Sheets signed by the customer and returned to Algor). The 

inspectors concluded that Algor had taken adequate measures to ensure that 

licensees had received notification of Class c errors. 

The inspectors noted that QAPM Section 7.1 stated that Qalifax subscribers 

would only receive QABs if the error was in purchased software while 

Section 7.3 stated that any QAB for Class C errors would be immediately 

distributed to all current subscribers of the Qal ifax subscription program. 

Algor indicated that although the procedure was contradictory the practice was 

to distribute QABs for all Class C errors to all Qalifax customers. The 

inspectors verified that QABs had been distributed to all Qalifax customers, 

regardless of software purchases. 

customer did not renew the yearly license for a piece of software and also did 

not renew the Qalifax subscription, all licensees would still be notified of 

all Class C errors which were applicable to the version of software that they 

had initially purchased as safety-related with a corresponding Qal ifax 

subscription. 

inconsistency in the Algor procedures, Algor had indicated an adequate 

position on the licensee notification of Class C errors. 

In addition, Algor indicated that, if a 

The inspectors concluded that a1 though there was an 

3.5 Purchase Order Review 

The NRC inspectors reviewed purchase orders (POs) to Algor from nuclear 

customers to determine the extent to which Appendix B and Part 21 requirements 

were imposed. The review determined that almost all the POs reviewed imposed 

such nuclear quality assurance requiremmts for software and re1 ated Qal ifax 

technical support. The following is a list of the POs which were selected for 

review during the inspection: 

Carolina Power & Liqht ComDanv 

PO No. 597206M-CR, dated December 9, 1993, ordered various finite-element 

analysis manuals including a Qalifax subscription, 

"Safety-Related" and imposed the requirements of Appendix B and Part 21. 

The corresponding Algor invoice for this order was Invoice No. 038558, 

dated December 23, 1993. 

The PO was marked 

30 No. 599637M-CR, dated March 31, 1994, ordered numerous unix platform 

stress and heat transfer software, including technical support. The PO was 

5 

8 
marked "Safety-Related" and imposed the requirements of Appendix B and 

Part 21. The corresponding Algor invoice for this order was Invoice No. 

040374, dated March 31, 1994. 

Florida Power & Liqht ComDanv 

PO No. B91536 01071, dated December 16, 1991, ordered various finiteelement analysis software including a Qalifax subscription. 

marked "Nuclear Safety-Re1 ated QL-1" and imposed the requirements of 

Appendix B and Part 21. 

Invoice No. 033730, dated February 26, 1993. 

The PO was 

The corresponding Algor invoice for this order was 

Enterqv ODerations 

Contract No. C-6135, dated November 30, 1994,'ordered various finiteelement analysis software, including a Qalifax subscription, for Arkansas 

Nuclear One, River Bend Station, Grand Gulf Nuclear Station and the 

Waterford Steam Electric Station, Unit 3, and imposed the requirements of 

Appendix B and Part 21. 

Invoice No. 044404, dated November 33 1994. 

The corresponding Algor invoice for this order was 

Public Service Electric and Gas CornDanv (PSE&G) 

PO P1-431.641, dated October 4, 1993; PO P3-0751638-1250-0000, dated October 

28, 1994, and PO P3-0826805-1240-0000, dated October 31, 1995. All three 

POs were for various software modules including a Qalifax subscription. 

All POs reviewed were for non safety-related material. The above PO review 

identified an extremely low level of nuclear safety-related activity 

processed by Algor over the past several years. 

Wolf Creek Nuclear ODeratinq CorDoration 

Algor Invoice No. 053195, dated December 28, 1995, referenced customer PO 

No. 566898 for various finite-element software including a Qal ifax 

subscription. 

but their quality Program Requirements Document No. AD17, Revision 00, 

dated December 27, 1995, identified in Section 3.1 that the reporting 

requirements of Part 21 applied. 

The Wolf Creek PO cw!d not be located during the inspection 

A review of revenue reports from Algor's nuclear customers identified only 

nine nuclear customers out of several thousand total customers purchasing 

Algor software and related services. However, not all nuclear customers may 

have purchased safety-related products and services from Algor pursuant to 

Appendix B and Part 21 requirements, as in the case of PSE&G. Algor's nuclear 

customers included Bechtel, Carolina Power & Light (CP&L), Crosby Valve, 

Entergy, EQE Engineering, Florida Power & Light (FP&L), Lockheed Idaho, 

Northern States Power, Proto-Power, PSE&G and Wolfcreek Nuclear. 

A review of all of the available audit reports of Algor from these customers 

only produced audit reports from FP&L, CP&L and Entergy. The review of the 

CP&L audit report, performed at Algor in September 1993, identified that it 

was a Nuclear Procurement Issues Council (NUPIC) based audit and as such, 

6 

9 
other NUPIC member 1 icensees could use the report for implementation purposes 

if it adequately covers their respective scope of supply. 

of the nuclear customers, purchasing activity preceded the CP&L inspection 

thereby making it impossible for them to take credit for the CP&L audit of 

However, for most 

A1 gor . 

The NRC inspectors concluded from its review that not all 

Appendix B and Part 21 software and technical services had 

quality program implementation compliance. This issue wil 

future NRC inspection at these licensees. 

4 PERSONS CONTACTED 

The following persons were contacted during the inspection 

noted, attended both the entrance and exit meetings. 

Alsor, IncorDorated: 

icensees purchasing 

audited Algor for 

be addressed in a 

and except as 

e M. Bussler President and Chief Executive Officer 

T. Anania Director of Operations R. Seebacher Quality Assurance Manager 

U.S. Nuclear Requlatorv Commission: 

R. Pettis Team Leader 

B. Rogers Reactor Engineer 

e Did not attended exit meeting 

7 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION .. . WASHINGTON, D.C. 206560001 

June 6, 1996 

Mr. William Zelner 

Qual i ty Manager 

AMP Products, Inc. 

250 Main Street 

Jacobus, PA 17407-4560 

SUBJECT: NRC INSPECTION REPORT NUMBER 99901295/96-01 

Dear Mr. Zelner: 

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

(NRC) inspection of your facilities performed April 29 through May 2, 1996, 

conducted by Mr. Ani1 S. Gautam of this office. The inspection involved 

activities authorized by Part 21, "Reporting Defects and Noncompliances," of 

Title 10 of the Code of Federal Resulations. 

2, 1996, during which we discussed our findings with you. 

The inspection was conducted to ascertain whether licensees effectively 

monitored the control of quality by AMP Products Inc. (AMP) for safety-related 

electrical terminals and splices purchased by licensees for nuclear power 

plants. The inspector assessed specific attributes and implementation of 

AMP's quality control program and the licensees' monitoring of these areas. 

An exit meeting was held on May 

We assessed attributes of your quality control program according to the 

criteria of 10 CFR Part 50, Appendix B, and the guidance in NRC Regulatory 

Guide 1.144, "Auditing of Quality Assurance Programs for Nuclear Power 

Plants." Details of the inspection are discussed in the enclosed copy of our 

inspection report. 

In general , AMP's qual i ty control program was effective and was appropriately 

monitored by licensees. The NRC inspector evaluated licensee audit findings 

and AMP corrective actions, commercial-grade item dedication, and monitoring 

of AMP's sub-vendors. 

In accordance with 10 CFR 2.790, a copy of this letter and its enclosures will 

be placed in the NRC Public Document Room. 

11 
W. Zelner -2- 

No response is required to this letter. 

inspection, we will be pleased to discuss them with you. 

If you have any questions about this 

Sincerely, - 

Robert M. Gallo, Chief 

Special Inspection Branch 

Division of Inspection and Support Programs Office of Nuclear Reactor Regul at i on 

Docket No. 99901295 

Enclosure: Inspection Report 99901295/96-01 

12 
U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF INSPECTION AND SUPPORT PROGRAMS 

REPORT NO. : 

ORGAN I ZAT ION : 

ORGANIZATIONAL 

CONTACT : 

INSPECTION DATES: 

INSPECTOR: 

APPROVED BY: 

99901295/96-01 

AMP Products, Inc. 

250 Main Street 

Jacobus, PA 17407-4560 

Mr. William Zelner, Manager 

Quality Assurance 

April 29 through May 2, 1996 

Ani1 S. Gautam 

Robert M. Gallo, Chief 

Speci a1 Inspection Branch 

Division of Inspection and Support Programs 

Enclosure 

13 
1 SUMMARY OF INSPECTION FINDINGS 

NRC conducted a special inspection to assess whether 1 icensees effectively 

monitored how well AMP Products, Inc. (AMP) controlled the quality of safetyre1 ated electrical terminals and splices purchased by 1 icensees for use in 

nuclear power plants. The inspection was conducted by an NRC inspector from 

the Special Inspection Branch of the Office of Nuclear Reactor Regulation at 

the AMP General Products Business Unit, Jacobus, Pennsylvania. The inspector 

assessed attributes of AMP's quality control program and its monitoring by 

licensees for the period of January 1994 through April 1396. 

In general, the AMP quality control program and its implementation were in 

compliance with the requirements of Appendix B to 10 CFR Part 50. 

effectively monitored AMP's quality control program through Nuclear 

Procurement Issues Committee (NUPIC) audit teams, questionnaires , and 

telephone surveys for safety-re1 ated electrical terminals and spl ices. 

inspector had no concerns about AMP's control of quality and licensees' 

monitoring of AMP. 

Licensees 

The 

Status of previous inspection findings are in Section 2. 

comments are in Section 3. 

attending the entrance and exit meetings are listed in Section 4. 

Findings and other 

The AMP personnel contacted, including those 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

This was the first NRC inspection of AMP Products, Inc. 

3 FINDINGS AND OTHER COMMENTS 

Licensees are required to monitor how well vendors control quality consistent 

with the importance, complexity, and quantity of products or services 

purchased from the vendors. NRC's evaluation of the licensee monitoring 

process falls, in part, under 10 CFR Part 50, Appendix B, Criterion VII, 

"Control of Purchased Material, Equipment, and Services," which requires 

licensees to establish specific measures to ensure that purchased material , equipment and services conform to the procurement documents, and comply with 

NRC guidance in Regulatory Guide 1.144, "Auditing of Quality Assurance 

Programs for Nuclear Power P1 ants, I' for auditing qual i ty assurance programs. 

The inspection was performed to ascertain whether licensees effectively 

monitored AMP's control of quality for the manufacture of safety-related preinsulated electrical terminals and splices purchased by licensees for use in 

Class 1E applications inside containments of nuclear power plants. The 

inspector assessed specific attributes of AMP's quality control program and 

the scope and the licensees' monitoring of these areas. 

The inspector examined AMP's quality program, including MP's qual ity control 

organization, conformance to procurement documents, evaluation and corrective 

actions in response to audit findings, validation of testing and certificates 

of conformance, commercial-grade item dedication, Part 21, "Reporting Defects 

and Noncompliances," of Title 10 of the Code of Federal ltequlations (10 CFR 

Part 21) evaluations, monitoring of sub-vendors, and self-assessment of 

performance. 

1 

14 
Li cen 

comprised of licensee staff. For example, AMP was monitored by Rochester Gas 

& Electric based on audits performed by NUPIC on July 24, 1995 (AMP supplied 

electrical terminals and splices for safety-related applications at the Ginna 

station). 

questionnaires and telephone surveys, and through audits conducted by other 

nuclear organizations. 

monitored AMP through audits conducted by NUPIC audit teams 

licensee monitoring was also accomplished through 1 icensee 

The inspector evaluated NUPIC audit reports, 1 icensee questionnaires and 

tel ephone surveys , audits conducted by Nucl ear Logi st i cs , Inc . and ABB Power 

T&D Company (vendors for nuclear power plants), nonconformance reports for 

safety-re1 ated terminals and splices, 1 icensee purchase orders, AMP purchase 

orders for sub-vendor, certification of conformance, commercial-grade process, 

responses to licensees' audits, AMP internal audit report, and AMP followup of 

NRC' information notices. 

3.1 AMP's Oualitv Control Proqram 

In general, the AMP quality control program and its 

compliance with the requirements of Appendix B to 10 

The inspector assessed AMP's quality pol icy, program in AMP's corporate quality specification 102-1 Revis 

mpl ementat i on were i n 

CFR Part 50. 

and standards described 

on K, "AMP Total Qual ity 

Management Process, li and supplemental specification GPBU 1-1002 Revision By 

"Addendum-Total Qual i ty Management Process. It 

attributes of AMP's quality program and how it conformed to 10 CFR Part 50 

Appendix B and NQA-1, "Quality Assurance Program Requirements for Nuclear 

Facilities." In catalog No. 82038, AMP described the safe 

preinsulated terminal and splices. (Note: As of April 1, 

Both documents described 

accepts purchase orders imposing 10 CFR Part 50 Appendix B 

but plans to maintain the existing quality program). 

AMP's senior management supported its qual i ty program, ass 

responsibilities. and remained involved in the implementat 

y-re1 ated 

1996, MP 

responsjb 

gned 

on of the 

no longer 

1 ities 

process. 

The' program was run by the quality manager and 6 quality auditors. The 

quality manager reported to the business unit director, who reported to the 

vice president and general manager of the Automoti ve/Consumer Business Group, 

who reported to the president/chief executive officer. The quality manager 

and auditors had the authority to stop production mtil nonconforming 

conditions were corrected. 

The inspector asked AMP for documents addressing any design errors and 

failures during manufacture requiring issuance of 10 CFR Part 21 reports 

during the past 3 years. AMP stated that it had found no design errors or 

deviations that required issuing a report in accordance with 10 CFR Part 21. 

The inspector requested information on any restrictions imposed by 1 icensees 

on AMP for the manufacture of electrical splices purchased by licensees during 

the past 3 years. AMP stated that no restrictions had been imposed on it and 

no stop-work orders had been issued by licensees. 

2 

15 
The inspector assessed AMP's commercial-grade dedication process. AMP 

purchased commercial-grade items from sub-vendors and qualified them to 

Appendix 6 requirements. The commercial-grade process comprised appropriate 

dedication activities, such as inspections, examinations, and witnessing of 

tests. Inspection findings were documented on data sheets and kept on file. 

Testing activities were identified in control plans and performed to test 

procedures. AMP evaluated test results for compliance to test requirements. 

The inspector examined AMP's cert i f i cate of conformance for i tems purchased 

under 1 icensee Purchase Order (PO) 65515004. The certificate attested that 

Kynar i nsul ated ring tongue terminal s and envi ronmental 1 y seal ed spl ices were 

qualified for use in nuclear power plants. The basis of the certificate was 

AMP Test Report 110-11004, Revision A; and Institute of Electrical and 

Electronic Engineers standards 323-1974, 344-1975, 383-1974, and Qual i ty 

Specification 102-1. 

qualified the test specimens mounted on terminal barrier blocks in specific 

configurations and for specific environmental conditions inside the 

containment. 

The inspector determined that Test Report 110-11004 

The AMP qual i ty program procedures required eval uati on of sub-vendors. 

31606201 for Kynar Grade 460 identified "Ausimont USA" as an AMP sub-vendor. 

The inspector questioned why AMP had not monitored Ausimont. The inspector 

also questioned why AMP had not monitored Grant Manufacturing, a sub-vendor 

for tin and lead anodes, and Hewlett Packard and Quality Technical 

Laboratories, sub-vendors for cal i bration services. The 1995 NUPIC audit team 

also noted that AMP had not audited these sub-vendors and issued audit finding 

SA-95-029-03 which identified, "Lack of audi tslsurveys for several sub-tier 

suppliers, and several performed surveys lacked objective evidence." The AMP 

quality manager stated that AMP planned to correct the situation by auditing 

Ausimont, Grant Manufacturing, and the providers of the necessary calibration 

services in late 1996 (eventhough AMP no longer accepts purchase orders 

imposing 10 CFR Part 50 Appendix B responsibilities). Since this finding was 

identified by NUPIC, AMP corrective action is in process, it was not a willful 

nonconformance, and it was not a repetitive problem, this nonconformance is 

being treated as a non-cited nonconformance, consistent with Section VII.B.l 

of the NRC Enforcement Policy. 

AMP PO 

The inspector reviewed 1 icensee PO 31606201 for AMP preinsulated splice part 

No. 53550-1. The PO required AMP to identify and correct/resQ!ve 

nonconforming conditions in accordance with the vendor's quality requirements 

and to submit any changes to the licensee for approval. 

verified that the safety-related terminals and splices conformed to 

procurement requirements. Original PO requirements were maintained to prevent 

inadvertent removal of requirements it was committed to meet. Certifications 

compared appropriately against the technical and quality requirements 

contained in the licensees' POs. 

The inspector 

AMP stated that it incorporated information on pertinent programmatic and 

hardware changes into service letters for distribution to appropriate 

customers. NRC Information Notice 88-81, "Failure of AMP Window Indent Kynar 

Splices and Thomas and Betts Nylon Wire Caps During Environmental 

Qualification Testing," alerted 1 icensees to the failure of AMP splices during 

licensee tests in which the splices touched each other or had a ground path. 

(Note: Test Report 110-11004 did not test the terminals and splices for 

3 

16 
configurations where they touched each other or had a ground path.) The 

inspector questioned if AMP issued any guidance to its customers addressing 

the failures noted in NRC Information Notice 88-81. The AMP quality manager 

could not identify any guidance issued to AMP’s customers to address these 

fail ures . 

The inspector reviewed selected AMP nonconformance reports to determine if 

adequate actions were taken to correct defects or weaknesses identified by AMP 

quality inspections or NUPIC audits. 

reports for AMP terminals and splices (part Nos. 53956-1, 53409-1, and 

53550-1) and determined that AMP tracked, handled, and took corrective action 

in response to the nonconformance reports. 

The inspector reviewed nonconformance 

The inspector assessed AMP’s internal audit Report No. 032096, dated March 20, 

1996. The audit identified 10 nonconformances, involving documentation 

deficiencies for design reviews, corrective actions, and maintenance 

procedures. 

completed. Eight observations were forwarded to management , among them an 

observation on effectiveness of corrective actions. 

Three corrective actions pertinent to the findings remain to be 

3.2 Licensee Monitorincl of AMP 

In general , 1 icensees effectively monitored AMP’s quality control program for 

safety-re1 ated electrical terminal s and spl ices. 

AMP identified two NUPIC audits conducted in October 1993 and July 1995 on 

licensee monitoring of how well AMP controlled the quality of safety-related 

electrical terminals and spl ices purchased by 1 icensees for use in nuclear 

power plants, and a 1994 NUPIC survey of AMP’s process for manufacture of 

commercial-grade non-nuclear items. 

questionnaires and telephone surveys conducted to monitor AMP. 

AMP also identified licensee 

AMP identified the following licensees to have purchased nuclear grade AMP 

terminals and spl ices qual ified to 10 CFR Part 50 Appendix B during 1993 - 1996: 

Licensee Remarks 

Arizona Pub1 ic Service Participated in 1993 NUPIC audit; 

conducted 1993 annual questionnaire. 

Boston Edi son Participated in 1993 NUPIC audit. 

Carolina Power & Light 

Duquesne Light 

Houston Lighting & Power 

I1 1 i noi s Power 

- 

Conducted 1995 telephone survey. 

- 

4 

17 
IES Utilities Conducted 1993, 1994, & 1995 annual 

questionnaire; conducted 1996 telephone 

survey. 

Northeast Uti 1 it i es 

Phi 1 adel phi a Electric 

Rochester Gas & Electric 

Southern Nuclear Operating 

Tennessee Val 1 ey Authority 

Texas Utilities Electric 

Washington Pub1 ic Power Supply 

Participated in 1993 & 1995 NUPIC audits; 

conducted 1996 telephone survey. 

Conducted 1993, 1994, & 1995 telephone 

survey. 

Conducted 1995 tel ephone survey. 

Participated in 1993 NUPIC audit; 

conducted 1993, 1994 & 1995 annual 

questionnaire; conducted 1993 telephone 

survey. 

Conducted 1993, 1994, & 1995 annual 

questionnaire. 

(Note: NUPIC sends questionnaires to licensees to plan areas for inspection of 

the vendor, and makes audit results available to all licensees. Licensee use 

of NUPIC for auditing vendors is based on 10 CFR Part 50, Appendix By and 

Regulatory Guide 1.144, which allow use of outside organizations to reduce the 

number of external audits as an alternative method for qualifying and 

monitoring vendors as long as all pertinent information is adequately 

eval uated . ) 

The inspector determined that the NUPIC audit reports were, in general, 

performed in accordance with written procedures and checklists. Audits 

comprised monitoring, witnessing, and observing activities, such as 

inspections, examinations, and performance tests. 

Arizona Public Service led the 1993 NUPIC audit regarding the control of 

quality by AMP at its facility in Mechanicsburg, Pennsylvania. The audit was 

conducted according to the standards of ANSI N45.2-1977, "Quality Assurance 

Program Requirements for Nuclear Facilities." The audit team uncovered no 

deficiencies or weaknesses, and the quality assurance program was found to be 

effectively implemented. However, the audit report gave insufficient detail 

of activities conducted by the audit team. 

IES Utilities led the 1995 NUPIC audit that was performed at five AMP 

facilities to review the AMP's QA program "from a performance-based aspect.' 

The team identified 12 deficiencies and made three observations. AMP 

responded to the findings and the licensee accepted AMP's responses. The 

report contained sufficient details of the review but did not state whether 

the audit criteria included verifying that AMP quality activities were in 

conformance with 10 CFR Part 50 Appendix B. 

ABB Power T&D Company Inc. (ABB Power) audited its sub-vendor AMP in November 

1994 to assure that AMP's quality program complied with their quality 

5 

18 
assurance programs. The audit focused on commercial-grade (non-nuclear grade) 

AMP products. 

the audit team. 

Nuclear Logistics Inc. (NLI) audited five facilities of its sub-vendor AMP in 

December 1995. The audit plan addressed 10 CFR Part 50 Appendix B criteria 

and critical characteristics. NLI monitored compliance to 10 CFR 50 Appendix 

B activities. The report did not state whether AMP was in compliance with 10 

CFR 50 Appendix B; however, when the inspector contacted the NLI lead auditor 

by phone the auditor verified that AMP was in compliance with Appendix B. No 

audit findings were identified. 

The report gave sufficient detail of activities conducted by 

No audit findings were identified. 

The inspector assessed the questionnaires and telephone surveys conducted by 1 i censees duri ng the period assessed. In general , 1 i censee eval uat i ons were 

appropriate, Areas evaluated by licensees included potential problems or 

changes to AMPy s product 1 i ne , qual i ty program, or procedures; purchase orders 

or procurement speci f i cat i ons; and sub-vendors , f aci 1 i ties , personnel , or 

quality involvement which could have an effect on product qualifications. 

To assess the staff's conclusions on the adequacy of licensees' monitoring 

AMP's control of quality over the past 3 years, the inspector interviewed the 

following AMP staff: the quality manager, two quality control auditors, and 

the manufacturing team leader. In general, the AMP staff considered NUPIC 

audits a 'lplus." The quality manager stated that the 1995 NUPIC team was very 

detailed and did not take credit for areas reviewed during previous NUPIC 

audits. 

4 PERSONS CONTACTED 

The NRC inspector and AMP staff contacted during the inspection are listed 

bel ow. 

AMP Products, Inc. 

* Will 

Ruth 

Car0 

Bill 

am Zelner 

Hershey Qual i ty Auditor, Team Leader 

Marshal 1 Manufacturing Team Leader 

Qual i ty Manager 

Presnell Auditor 

Nuclear Regul atory Commission 

* Ani1 S. Gautam Team Leader, NRR 

* Attended the entrance and exit meeting 

6 

19 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. #)565.oOol 

May 31, 1996 

Ms. Beth A. Barbone, Operations Manager 

Cegelec Automation Inc. 

2806 Metropolitan Place 

Pomona, California 91767 

SUBJECT: NRC INSPECTION NO. 99900734/96-01 

Dear Ms. Barbone: 

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

(NRC) inspection of Cegelec Automation Inc. at Pomona, California, conducted 

by Mr. R.C. Wilson of this office on May 7 through 9, 1996. The purpose of 

the inspection was to review activities conducted under your 10 CFR Part 50, 

Appendix By quality assurance program and 10 CFR Part 21 reporting program. 

The inspection consisted of an examination of procedures and records, 

interviews with personnel, and observations by the inspectors. 

The NRC inspector identified that certain of your activities appeared to be in 

violation of NRC requirements, as specified in the enclosed Notice of 

Violation (Enclosure 1). 

safety concerns under 10 CFR Part 21 did not address the evaluation of 

deviations from purchase order requirements, and the procedure did not reflect 

significant changes made to 10 CFR Part 21 in 1995. 

concern because deviations are the type of safety concern most likely to be 

identified by a vendor. 

Specifically, your procedure for reporting nuclear 

The violation is of 

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

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

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

additional actions you plan to prevent recurrence. After reviewing your 

response to this Notice of Violation, including your proposed corrective 

actions and the results of future inspections, the NRC will determine whether 

further NRC enforcement action is necessary to ensure compliance with NRC 

regulatory requi rements. 

In addition, the NRC inspector found that the imlementation of your qual 

assurance program failed' to meet certain NRC requirements. Specificaily, 

procedures and instructions for dedi cat ing commerci a1 grade i tems for nuc 

safety-related use were inadequate to ensure that the dedicated items wer 

suitable for the application. Furthermore, even though your actual dedication 

practices went beyond the requirements of the procedures and instructions, the 

dedication of a relay shipped for safety-related use was inadequate in two 

respects. 

are identified in the enclosures to this letter. 

The specific findings and reference to the pertinent requirements 

tY 

your 

ear 

20 
Ms. Barbone -2- 

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

statement in accordance with the instructions specified in the enclosed Notice 

of Nonconformance. 

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

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

Room. 

Sincerely , 

/ Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Docket No. 99900734 

Enclosures: 1. Notice of Violation 

2. Notice of Nonconformance 

3. Inspection Report 99900734/96-01 

21 
NOTICE OF VIOLATION 

Cegel ec Automation Inc . Pomona, California 

Docket No.: 99900734 

Buring an NRC inspection conducted on May 7 through 9, 1996, a violation of 

NRC requirements was identified. 

Policy and Procedure for NRC Enforcement Actions," 10 CFR Part 2, Appendix C 

(1992), the violation is listed below: 

In accordance with the "General Statement of 

10 CFR 21.21, "Notification of failure to comply or existence of a defect and 

its evaluation," requires, in part, that each corporation subject to the 

regul at i ons adopt appropriate procedures to ensure the eval uat i on and proper 

reporting of deviations and failures to comply. 

Contrary to the above, Procedure 16.1, "Significant Deficiency 'Substantial 

Safety Hazard,"' Revision 6, of the Ccgelec Quality Assurance Manual dated 

September 19, 1994, failed to address the evaluation of deviations. Instead 

the procedure, and the posted letter referring employees to the procedure, 

incorrectly focused on the terms "failure to comply" and "substantial safety 

hazard." Procedure 16.1 also failed to reflect significant changes that were 

incorporated into 10 CFR Part 21 in 1995. 

This is a Severity Level IV violation (Supplement VII). (99900734/96-1-1) 

Pursuant to the provisions of 10 CFR 2.201, Cegelec Automation Inc., is hereby 

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

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

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

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

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

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

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

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

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

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

compliance will be achieved. Where good cause is shown, consideration will be 

given to extending the response time. 

Dated at Rockville, Maryland 

this 3iJ- day of rfl -3- lgg6 

Enclosure 1 

22 
Cegel ec Automation Inc. 

Pomona, Cal i forni a 

NOTICE OF NONCONFORMANCE 

Docket No.: 99900734 

Based on the results of an inspection conducted on May 7 through 9, 1996, it 

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

NRC requirements as described bel ow. 

The definition of "dedication" in 10 CFR 21.3 specifies that dedication must 

provide reasonable assurance that a commercial grade item to be used as a 

basic component will perform its intended safety function, and is deemed 

equivalent to an item designed and manufactured under a 10 CFR Part 50, 

Appendix By quality assurance program. This assurance must be achieved by 

identifying the critical characteristics of the item and verifying their 

acceptabil i ty. 

Criterion I11 of Appendix B to 10 CFR Part 50, "Design Control," requires, in 

part, that measures shall be established for the selection and review for 

suitability of application of equipment essential to safety-related functions. 

Section 3.2 of Procedure 3.0, "Design Control," Revision 6, of Cegelec 

Automation Inc. Quality Assurance Manual dated September 19, 1994, requires, 

in part, that specified design requirements such as regulatory requirements 

shall be translated into drawings and/or procedures from customer design 

speci f i cat i ons . 

A. Contrary to the above, for nuclear safety-related purchase orders 

including Florida Power and Light Company order number 00015706 dated 

April 18, 1996, covering a Potter & Brumfield KRP-11AG-120 relay shipped 

to the Turkey Point nuclear plant on April 25, 1996, Cegelec performed 

dedication activities as specified in Engineering Department Work 

Instruction #4, "Dedication of Commercial Grade Material ,I' Revision 3, 

dated April 16, 1996. This instruction did not contain sufficient 

detail to adequately defice dedication activities, and it contained 

definitions that differ considerably from 10 CFR Part 21, including the 

statement that a critical characteristic provides reasonable assurance 

that the item received is the item specified. 

following nonconformance, dedication activities for this purchase order 

were in fact deficient. Dedication activities under other purchase 

orders examined by the tiRC inspector were considered acceptable only 

because they included efforts not specifically required by the work 

instruction. (99900734/96-01-02) 

As described in the 

B. Contrary to the above, Cegelec's dedication activities failed to 

establ ish reasonable assurance that the re1 ay del ivered to Florida Power 

and Light Company met the applicable requirements for its safety-related 

function, because there was no documented basis for the following: 

Enclosure 2 

23 
1. assurance that the purchased lot of relays was homogeneous, which 

was a necessary requirement for the sampling plan used to select 

relays for verification of critical characteristics. 

2. assurance that the purchased lot of relays were similar to the 

relays that were seismically and environmentally type-tested. 

(99900734-01 -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 Inspection 

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

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

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

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

will be taken to correct these items; (2) a description of steps that have 

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

actions and preventive measures were or will be completed. 

Dated at Rockville, Ma yland 

this 3,/k day of &f 

-+, lgg6 

-2- 

24 
REPORT NO. : 

ORGAN I ZAF I ON : 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY: 

INSPECTION DATES: 

INSPECTOR : 

REVIEWED BY: 

APPROVED BY: 

U S . NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF INSPECTION AND SUPPORT PROGRAMS 

99900734/96-01 

Cegelec Automation Inc. 

2806 Metropol i tan P1 ace 

Pomona, California 91767 

Eric Morales, Quality Assurance Manager 

909/593-8099 ext. 235 

Hydrogen and oxygen containment atmosphere 

monitors 

May 7 through 9, 1996 

Richard C. Wilson, Senior Engineer 

5/31/96 

Date - Vendor Inspection Section 

Speci a1 Inspect ion Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

5M94 

Date 

/p / A/ 

I n/-- 92 

aha,- Chie'f 

tion Section 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

& Date 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Enclosure 3 

25 
1 SCOPE OF INSPECTION: 

During the 1980s Cegelec Automation, then known as Comsip Del phi , supplied 

containment atmosphere hydrogen analyzers for nore than 40 nuclear pi ants, 

with oxygen analyzers a1 so suppl i ed for BWRs. Safety-re1 ated rep1 acement 

parts for these systems account for about 25% of current business. 

has 16 employees and occupies about 12,000 square feet. 

The hydrogen monitor pulls a sample of containment atmosphere through a heated 

3/8-in. line with containment isolation valve5 into a cabinet where the 

hydrogen concentration is measured in a thermal conductivity cell at about 

280°F. The sample is then returned to containment through another 3/8-in. 

line. The system is leak tested at 90 psig. 

Cegelec 

The NRC inspectors reviewed the implementation of selected portions of 

Cegelec's quality assurance (QA) program, and reviewed Cegelec's 10 CFR 

Part 21 program. The inspection bases were 10 CFR Part 50, Appendix By and 

10 CFR Part 21. 

1.1 Violation 99900734/96-01-01 (ODenl 

Contrary to 10 CFR 21.21(a), Cegelec did not have a procedure in place for 

evaluating deviations. 

changes in the 1995 revision of 10 CFR Part 21. 

see Section 3.4 of this inspection report.) 

1.2 Nonconformance 99900734/96-01-02 (Open) 

Cegelec's procedures a1 so did not reflect significant 

(Severity level IV violation; 

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

Cegelec's dedication procedures failed to specify appropriate quality 

standards for the review for suitability of application of commercial grade 

equipment dedicated and suppl ied under safety-re1 ated purchase orders (POs) from 1 icensees. 

definitions. 

The procedures 1 acked necessary detai 1 and used incorrect 

(See Section 3.2 of this inspection report.) 

1.3 Nonconformance 99900734/96-01-03 (ODenl 

Contrary to Criterion 111, "Design Control," of Appendix 8 to 10 CFR Part 50, 

Cegelec performed inadequate sampling of commercial grade items for testing 

and did not establish similarity between type test specimens and a relay 

delivered under a safety-related PO from a licensee. 

inspection report.) 

(See Section 3.3 of this 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

There were no open findings from previous NRC inspections of Comsip Delphi, 

which were performed in 1981 and 1982. 

2 

26 
3 INSPECTION FINDINGS AND OTHER COMMENTS 

3.1 Entrance and Exit Meetinss 

In the entrance meeting on May 7, 1996, the NRC inspector discussed the scope 

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

interfaces with Cegelec management. In the exit meeting on May 9, 1996, the 

inspector discussed his findings and concerns with Cegelec management. 

3.2 Quality Assurance Proqram 

The inspector selectively reviewed the Cegelec Quality Assurance (QA) Manual, 

Revision 7, dated September 19, 1994, the current departmental work 

instructions for the QA, Engineering, Procurement, and Service Departments, 

and several Shop Test Procedures (which also cover assembly and other 

operations). Cegelec's QA program was intended to meet the requirements of 

Appendix B to 10 CFR Part 50 for nuclear safety-related equipment. 

manager stated that the QA program and customer PO requirements were applied 

to a1 1 work performed. 

The QA 

QA Manual Procedure 6.0, "Document Contrcl," Revision 6, dated September 19, 

1994, appeared to adequately cover control of documents such as procedures and 

drawings. The inspector noted that QA Work Instruction #3, "Internal Audits," 

Revision 7, dated November 20, 1995, did not limit the time allowed for an 

audited department to respond to a finding. Cegelec's QA manager agreed to 

specify a limit in the next scheduled revision of the QA manual by August 15, 

1996. 

Most of Cege'lec's procurements were commercial grade. Piece parts were 

dedicated by receipt inspection and testing, and assemblies built from piece 

parts were further tested. 

not define a clearly acceptable commercial grade dedication process. For 

example, the principal dedication procedure was Engineering Department Work 

Instruction #4, "Dedication of Commercial Grade Material," Revision 3, dated 

April 16, 1996, which contained very little detail. It contained definitions 

that differ considerably from the 1995 revision of 10 CFR Part 21, including 

the statement that a critical characteristic provides "reasonable assurance 

that the item received is the item specified." 

The inspector found that Cegelec's procedures did 

The definition of "dedication" in 10 CFR 21.3 specifies that dedication must 

provide reasonable assurance that a commercial grade item to be used as a 

basic component will perform its intended safety function, and is deemed 

equivalent to an item designed and manufactured under a 10 CFR Part 50, 

Appendix B, QA program. This assurance must be achieved by identifying the 

critical characteristics of the item and verifying their acceptability. 

Cegelec's dedication procedures failed to specify appropriate quality 

standards for the review for suitability of application of commercial grade 

equipment dedicated and supplied under safety-related POs from licensees, as 

required by Criterion 111, "Design Control," of Apkendix B to 10 CFR Part 50. 

This constitutes Nonconformance 99900734/96-01-02. 

3 

27 
The inspector found that Cegelec's dedication practices, as reflected in 

detailed review of 1 icensee purchase order (PO) files, went beyond procedural requirements and were closer to being acceptable. 

dedication practices also contained deficiencies, as described in Section 3.3 

of this inspection report.) 

revision by August 15, 1996, and management agreed to consider the results of 

this inspection in performing the revisiop 

Instructions for hand1 ing nuclear safety-re1 ated procurements, while 

apparently adequate, were distributed among four procedures and were difficult 

to locate. Procurement Department Work Instruction #2 , "Control and Issuance 

of Purchase Orders," Revision 1, dated July 12, 1995, defined three levels of 

POs. Level I1 was defined as POs for "materials or services which include 

specific technical and/or quality requirements." POs for "critical materials 

or services are only placed with suppliers found on the CEGELEC Automation 

Inc. Qual i f i ed Vendors Li st. 'I Critical materi a1 and service suppl i ers were 

defined as suppl i ers of safety-re1 ated or qual i f i ed equi pment , speci a1 process 

(welding, NDE), or calibration services. QA Work Instruction #14, "Supplier 

Evaluation and Approval, Revision 3, dated December 23, 1994, required 

triennial vendor surveys of suppliers of "Critical Materials and Services." 

Vendor surveys required "direct observation of objective evidence at the 

suppliers facility which substantiates the documented program." 

inspection on an individual order was permitted in lieu of a survey. QA 

Manual Procedure 4.0, "Procurement Document Control, Revision 6, dated 

September 19, 1994, and Procurement Work Instruction #2 both required imposing 

10 CFR Part 21 in safety-related orders; Procurement Department Work 

Instruction #3, "Spare Parts Procurement and Order Entry, Revision 6 dated 

March 30, 1995, required imposing Part 21 "as applicable." The Cegelec QA 

manager agreed to consider clarifying the treatment of safety-related POs by 

the August 15, 1996, revision of the QA manual. 

The inspector concluded that the reviewed QA manual, work instructions, and 

procedures were satisfactory, except for the 10 CFR Part 21 violation 

discussed in Section 3.4 of this inspection report and the two dedication 

nonconformances described in Sections 3.2 and 3.3. 

(However, the actual 

Cegelec planned to issue the next QA manual 

Source 

3.3 Review of Licensee Purchase Orders 

The inspector selected five licensee PO files for review, covering the 

following items purchased commercial grade and supplied safety grade: a relay purchased as a commercial grade component; a hydrogen sensing cell 

manufactured from commercial grade piece parts; a pressure regulator purchased 

as a custom variation of a commercial grade component; and a flow regulator 

manufactured from commercial grade piece parts. Although no hardware had yet 

been processed for the fifth licensee PO file reviewed, it provided an example of Cegel ec' s ident i fying and correct i ng errors in 1 i censee POs . 

Turkev Point: 

Flwida Power and Light Company (FPL) PO No. 00015706 dated April 18, 1996, 

ordered a replacement Potter & Brumfield KRP-1lAG-120 relay for the Turkey 

Point nuclear plant. The PO stated that it was FPL Nuclear Safety Related 

4 

28 
(PC-I), and it invoked 10 CFR Part 21, IEEE 323-1974 mild cwironment, and 

IEEE 344-1975. A Certificate of Compliance was also required. The Cegelec 

Contract Review form sumnarized the PO requirements and recorded Cegelec's 

sales order numbers for the original system and the replacement part order. 

The Sales Order form identified that the relay supplied on the replacement 

sales order 44942 had been purchased on Cegelec PO No. 1230-849. A copy of 

that PO was in the file, showing that 25 KRP-11AG 120 volt relays had been 

procured from the Electronic Supply company, a local distributor. Cegelec 

drawing 31041-17 Rev. 7 specified the following critical characteristics: 

part/model number, overall dimensions, maximum pul I-in and minimum drop-out 

voltages, and energized and de-energized contact hi-pot testing. 

The Receiving Inspection Report addressed simple checks such as physical 

damage and count. The Material Dedication Inspection/Test Report, also 

completed by the receiving inspector, stated that 3 of the 25 KRP-11AG relays 

were tested to verify the critical characteristics specified on the drawing. 

The sample size was prescribed in QA Work Instruction #1, "Receiving 

Inspection," Revision 11, dated July 12, 1995, and based on MIL-STD-105D. The 

3 relays tested were not specifically identified. For each characteristic, 

"Acceptable" was recorded. The test equipment was identified by serial number 

and calibration due date (although the inspector did not pursue calibration 

traceability, Cegelec's procedures require triennial audits of calibration 

service suppliers, which the QA manager stated were performed). 

receiving testing, the lot of 25 relays was placed in controlled storage with 

a green tag designating safety-related material. 

The invoice from Electronic Supply included a printed "Certificate of 

Compliance" stating that all materials and/or parts are in accordance with the 

applicable manufacturer's specifications, and that inspection records and test 

data are on file and available for review at the manufacturer's plant. 

C of C was signed by the Director of Operations. Cegelec did not audit 

Electronic Supply or Potter & Brumfield. Cegelec provided to FPL a 

Certificate of Conformance specifying the shelf life of the relay and stating 

that it was the same in form, fit, and function and interchangeable with items 

supplied under the original FPL specification, and that it satisfied IEEE 

323-1974 and IEEE 344-1975 as documented in the qualification test report for 

the original analyzer system. 

After 

This 

The inspector had two principal concerns with the dedication process for the 

re1 ay: 

(1) There was no assurance that the lot of 25 relays was a homogeneous 

population, as required for the validity of the sampling method used. 

Procurement from an unaudited commercial grade distributor, with no 

attempt to either control or identify characteristics such as date code 

or serial number, and with no audit of or correspondence with the 

manufacturer, provides 1 i ttl e basis for concluding that the 25 re1 ays 

were essentially the same. Nuclear industry experience with relays from 

this particular manufacturer does not provide the necessary confidence, 

and the certification from the unaudited supplier is of little value for 

dedication purposes. 

5 

29 
(2) there was no assurance that any of the replacement relays were 

sufficiently similar to those that were environmentally and seismically 

type-tested in 1980-82; i .e. , the dedication process failed to establish 

traceability of design similarity from the replacement relays to the 

type-test specimens. 

Because of these concerns, dedication of the Potter & Brumfield relay for 

Turkey Point did not satisfy Criterion 111, "Design Control," of Appendix B of 

10 CFR Part 50, which requires appropriate quality standards for the review 

for suitability of application of commercial grade equipment dedicated and 

suppl ied under safety-re1 ated POs from 1 icensees. These two examples of 

inadequate design control constitute Nonconformance 99900734/96-01-03. 

The sampling concern could be overcome with 100% sample testing of the 

replacement relays. 

be addressed by disassembly and comparison if the original test specimens or 

detailed records were still available; possibly the manufacturer's records 

could be useful. 

new type tests on samples from the lot is another possibility. 

The problem with similarity to type-test specimens could 

Procurement of a demonstrably homogeneous lot and performing 

Palo Verde: 

Arizona Public Service Company PO No. 60282071 dated January 4, 1996, ordered 

(along with other parts) a replacement hydrogen sensing cell , Cegelec type 

1427-85, for the Palo Verde Nuclear Generating Station. The PO specified the 

qualification report for the original system, invoked 10 CFR Part 21, invoked 

the current Cegelec QA manual and licensee QA Requirement 003, and stated that 

all items are safety-related. A Certificate of Conformance to the 

requirements of the PO was also required. 

Applicable Cegelec documents included the "Bill of Materials for Assembly 

Number 1427-B5," Revision 4, dated January 18, 1994, and Procedure 10.11, 

"85 Thermal Conductivity Cell Block Assembly and Test Procedure," Revision 3, 

dated January 16, 1995. The cell is manufactured by Cegelec from commercial 

grade piece parts. A Material Requisition form for Sales Order 58265 

identified the PO for each of the 22 parts on the Bill of Material. Most 

parts were stainless steel, and the part drawing typically specified two 

critical characteristics: dimensions, and material as verified by test 

procedure 11.1, "Material Verification Test Procedure," Revision 1, dated 

December 20, 1990. This procedure also specified MIL-STD-105D sampling 

(typical samples were 32 of 530 diffusion discs and 2 of 8 cell blocks). 

Chemical and physical mill certifications were typically obtained for steel 

parts. The procurements were from distributors which Cegelec did not audit. 

Procedure 11.1 uses a Koslow 1599 alloy test kit to identify stainless steel 

when grade is not critical, or when type 316 is specified. In either case, 

electrical conductivity of the specimen is measured. A test solution color 

test for high chromium content typical of 300 and 400 series stainless steels 

is used for all types, and a second solution tests for approximately 3% 

molybdenum content for type 316 stainless. 

also provides material samples for visual comparison. Procedure 11.1 further 

specifies a flame test for inorganic materials, typically used for O-rings. 

A Koslow 1900 metal standard set 

6 

30 
The NRC inspector found that some parts drawings did not specify critical 

characteristics. An example was drawing 1018-C, "Model B5 O-ring Spacer 

Catalyst Bed," Revision 0, dated April 2, 1990. In these cases the Spare 

Parts Inspection Reports nonetheless showed dimensions and material as 

critical characteristics, and sampling test results were recorded. The QA 

manager explained that some older drawings had not been updated to define 

critical characteristics, but the receiving inspector would still perform 

typical tests. The NRC inspector also found no record of receipt testing for 

two parts, a hermetic seal and a resistor. 

these parts were dedicated during assembly and testing of the cell; the 

hermetic seal was tested during the pressure test and electrical testing, and 

the resistor value and temperature response were verified during electrical 

balancing of the cell. In these cases the NRC inspector again noted that 

Cegelec's actual performance went beyond the content of procedures, which did 

not address these variations. 

The QA manager stated that both of 

The NRC inspector again believed that the mill certifications and limited 

testing of sampled parts were not alone sufficient to dedicate the parts. 

However, the assembly operations and testing performed under procedure 10.11 

(including a one minute pressure test at 30 psig), with the other activities 

performed by Cegelec and the necessity for the customer to calibrate the cell 

after installation, are considered sufficient for dedication. 

Perry: 

Cleveland Electric Illuminating Company PO No. S 139804 dated December 13, 

1995, ordered two rep1 acement GH41XT-1833-1 pressure regulators for the Perry 

Nuclear Power Plant. The PO specified safety related procurement level I, 

required a Certificate of Conformance, invoked 10 CFR 50 Appendix B and 

10 CFR Part 21, and referenced the original system PO, specification, and 

qualification test report. 

model, custom engineered for Cegelec, and is procured commercial grade under 

Cegelec drawing 31041-27, "Downstream Pressure Regulator Set at 3 psig, 

Conoflow Mdl. # GH41XT--1833-1," Revision 11, dated April 18, 1995. The 

inspector did not review the Drawing Control Change Notices for this drawing. 

The regulator is a version of a standard Conoflow 

The two regulators suppl ied to Perry had been procured by Cegelec under two 

different POs, 30-808 and 30-823, from A. Biederman, Inc., a local 

representative. 

dimensions as critical characteristics, which were checked on 5 samples in 

each lot of 50. 

be checked. 

Procedure 6.10, "Model GH41XT-1833-1 Differential Pressure Regulator Bench 

Test Procedure," Revision 5, dated February 9, 1995. 

a ten minute leak test at 90 psig and functional performance testing. The 

inspector concluded that the dedication activities for this regulator were 

satisfactory, considering the application. 

The drawing speci f i ed part/model number and overall 

For PO 30-808 a damaged thread on one sample caused all 50 to 

The drawing also specified 100% testing of all regulators to 

This procedure includes 

D.C. Cook: 

American Electric Power PO. No. 66069-040-6X dated February 16, 1996, ordered 

(along with other parts) two replacement model 11727 R-1 downstream flow 

7 

31 
regulators for the Cook Nuclear Plant. The PO stated that it was nuclear 

safety-related, it invoked 10 CFR Part 21, and it required certification that 

the part is equal to or better than the original. The regulator is built 

according to drawing 11727, "Sample Flow Regulator R1 KIII and KIV," Revision 

10, dated May 2, 1995. 

Change Notices for this drawing. The drawing notes specified the applicable 

procedures for assembly and test: Procedure 10.6, "Cegelec Automation Inc. 

Model 11727 Series (R,) Downstream Pressure Regulator," Revision 8 dated May 

1, 1995, and Procedure 6.4, "Model 11727 Series (R,) Downstream Pressure 

Regulator Bench Test Procedure," Revision 7, dated May 1, 1995. 

parts drawings checked by the inspector typically identified critical 

characteristics as partjmodel number, dimensions, and materi a1 verification 

per procedure 11.1. 

The inspector did not review the Drawing Control 

In addition, 

The inspector checked the procurement of the regulator springs, which were 

ordered to drawing 11768, "Regulator Spring for Rl," Revision 2, dated 

December 12, 1989. The springs were ordered on PO No. 1270-209 from Superior 

Spring Company. 

identified, the drawing did specify AMS-5673 or AMS-5678, Grade 17-7 stainless 

steel, 0.035 inch wire diameter, and it specified a spring force (mislabeled 

"pressure") at a specific working height as measured in the test fixture shown 

on the part drawing. Superior Spring, an unaudited commercial grade supplier, 

provided certification of the following: 0.035 17-7SS material, specification 

AMS 5678, passivated, heat treated 900 deg./l hr., specifications MIL-S-5002 

and CH 900/MIL-H-6875. 

showed that a sample of 8 of 100 springs was tested for dimensions and 

material (stainless steel per procedure 11.1). The file gave no indication 

that the spring force was tested, even though the test fixture specified on 

the drawing was a simple cylinder. 

Although no critical characteristics were specifically 

The Material Verification Test Report for the PO 

The Procedure 6.4 test reports for the two regulators supplied to Cook showed 

that both passed the ten minute 90 psig leak test, and both maintained an 

acceptable vacuum for inlet pressures up to 60 psig, with acceptable 

repeatability. Even though critical characteristics for the piece parts were 

not consistently specified or verified, the inspector concluded that final 

testing adequately dedicated the regulator for its application. 

Millstone: 

Northeast Nuclear Energy Company PO No. 956726 dated November 1, 1995, ordered 

two model 11727 and two model 11728 pressure regulators for the Millstone 

Nuclear Power Station. 

but had sent a letter to the licensee correcting three errors in the PO: 

number of the original system PO, the date of Cegelec's QA manual , and the 

form number of the attachment designating QA requirements including nuclear 

safety-related. (The inspector noted that, a1 ternately, the wrong PO note and 

correct form may have been i ncl uded . ) These correct ions exempl i fy proper 

implementation of QA Manual Procedure 19.0, "Contract Review," Revision 0, 

dated September 19, 1994. 

on a PO reduces the chance that incorrect safety-related equipment may be 

suppl i ed. 

Cegelec had not yet supplied equipment under this PO, 

the 

Correction of such errors prior to beginning work 

8 

32 
3.4 10 CFR Part 21 Proaram 

The inspector reviewed Cegel ec's procedure for reporting in accordance with 

10 CFR Part 21: QA Manual Procedure 16.1, "Significant Deficiency 

'Substantial Safety Hazard'," Revision 6 dated September 19, 1994. 

procedure, and the letter referring to the procedure that was posted on a 

bulletin board, incorrectly focused on the terms "failure to comply" and 

"substantial safety hazard. I' A1 though the procedure correctly repeated the 

definition of "deviation" from 10 CFR 21.3, it did not address Cegelec's 

responsibility to evaluate deviations. Instead, Procedure 16.1 addressed the 

evaluation by Cegelec of "failures to comply" as possible substantial safety 

hazards. However, 

senses, equivalent to either deviations or defects, and Cegelec used the wrong 

sense, equivalent to defect. Procedure 16.1 (as we71 as Cegelec work 

instructions for commercial grade item dedication discussed in Section 3.3 of 

this inspection report) also failed to reflect significant changes in the 1995 

issue of 10 CFR Part 21. 

Both the 

10 CFR Part 21 uses the term "failure to comply" in two 

The inspector explained that Cegelec must have procedures for evaluating 

deviations and failures to comply under 10 CFR 21.21(a). 

rarely be in a position to perform thz required evaluation, the inspector 

suggested that the procedure focus on satisfying this requirement by notifying 

customers as allowed by Section 21.21(b) if the vendor is unable to perform 

the evaluation. The inspector also asked that Cegelec's revised procedure 

encourage employees to report possible deviations, either formally or 

informally, to their supervisors or QA. The inspector found no evidence that 

deviations requiring evaluation had ever gone unevaluated or unreported to 

licensees, and only one Part 21 concern had been reported (that by a licensee, 

as discussed below) since initial qualification testing in 1980-82. 

Since Cegelec wculd 

Failure to have procedures that require the evaluation or reporting of 

deviations, as required in 10 CFR 21.21(a), and failure to incorporate the 

significant changes of the 1995 revision of 10 CFR Part 21, constitute 

Severity Level IV violation 99900734/96-01-01. 

The inspector reviewed a 10 CFR Part 21 concern initially reported by the Hope 

Creek Generating Station of the Public Service Electric and Gas Company by 

telephone on July 7, 1993, involving a higher-than-analyzed ambient 

temperature for microswitches used in hydrogen analyzers. The concern was 

subsequently addressed in written Part 21 reports from Hope Creek on August 5, 

1993, and from Cegelec on March 2, 1994. The NRC log numbers for these 

reports are 93-294, 93-323, and 94-099. The reported concern was that 

Microswitch model PTW-5300 switches installed at Hope Creek were experiencing 

unspecified damage because their operating temperature was high. The licensee 

postulated that humidity could cause a short circuit that would disable a 

safety-related display, and stated that the qualified life of the switches was 

shortened. The Cegelec report stated that the Hope Creek switches were being 

operated more than 10% above their rated voltage, and noted that when the 

voltage was reduced to the rated 120 vdc, the temperature decreased by 44 F". 

During this inspection the inspector verified that the concern is not generic. 

9 

33 
r 

The Ametek/Dixson company submitted a 10 CFR Part 21 report dated April 18, 

1996, concerning bargraph indicators supplied to Cegelec and other customers. 

The inspector verified that Cegelec used the indicators only for a foreign 

customer; none were provided for domestic use. 

3.5 Industry GrouD Audit 

The inspector reviewed the reports of an industry group audit of Cegelec in 

January 1995, and a follow-up surveillance in May 1995. The surveillance 

report showed that the four audit findings had been corrected. 

noted that the industry group audit had thoroughly covered conformance to 

manufacturing procedures, which the NRC inspector did not have the opportunity 

to do. 

that Cegelec had developed and implemented appropriate procedures for the 

evaluation of Part 21 conditions, although that was not the case. The 

industry group audit otherwise appeared to be satisfactory. 

The inspector 

The inspector also noted that the industry group audit report stated 

4 PERSONNEL CONTACTED 

t * Beth A. Barbone, Operations Manager 

t * Eric Morales, Quality Assurance Manager 

t * Scott E. Crail, Engineering Manager 

Karen Cobb, Spare Parts 

Kevin Mullane, Receiving Inspector 

t Attended the entrance meeting on May 7, 1996 

* Attended the exit meeting on May 9, 1996 
***** ' 

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 2055!WO01 

May 28, 1996 

Mr. Arthur J. Spencer, Manager 

ASME Codes and Standards 

Factory Mutual Engineering Association 

1151 Boston-Providence Tpke. 

Norwood, MA 02062 

SUBJECT: NRC Inspection No. 99900603/96-01 

Dear Mr. Spencer: 

This letter addresses the U.S. Nuclear Regula,ary Commission (NRC) inspection 

of your offices in Bala-Cynwyd, PA on April 22 through 24 and in Norwood, MA 

on April 25 and 26 by Messrs. U. Potapovs, R. McIntyre, and S. Matthews and 

the discussions of their findings with Mr. S. Rudnickas, you, and members of 

your staff at the conclusion of the inspection. The inspection was conducted 

to evaluate the implementation of your quality assurance program in selected 

areas related to the performance of activities under the scope of American 

Society of Mechanical Engineers (ASME) Certificate of Accreditation issued to 

Arkwright Mutual Insurance Company. The scope of this inspection was focused 

on the Authorized Nuclear Inspector (ANI) and the Authorized Nuclear Inspector 

Supervisor (ANIS) services that your company provided to Amer Industrial 

Technologies, Inc. (AIT). Your program for compliance with the requirements 

of Part 21 of Title 10 of the Code of Federal Requlations (10 CFR Part 21) was 

also reviewed during this inspection. 

The inspection was accomplished through objective evaluation of selected 

procedures and records and discussions with appropriate staff. 

areas examined during the NRC inspection and the findings are discussed in the 

enclosed inspection report. 

The specific 

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

activities appeared to be in violation of NRC requirements. 

your procedure CP 110.00, "Title 10 Code of Federal Regulations Part 21", 

dated January 19, 1996, which defines the policy for 10 CFR Part 21 

compliance, did not provide for evaluation of deviations identified as a 

result of your activities at nuclear facilities and at other covered entities 

that manufacture basic components. 

Specifically, 

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

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

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

additional actions you plan to prevent recurrence. 

35 
A. Spencer -2- 

Additionally, during this inspection, we identified instances where the 

implementation of your qual i ty assurance program fai 1 ed to fully comply with 

the ASME Code requirements that are applicable to your activities under the 

scope of the Certificate of Accreditation issued to Arkwright Insurance 

Company dba Factory Mutual Engineering Association. The specific findings and 

references to the pertinent requirements are identified in the enclosures to 

this letter. 

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

accordance with the instructions specified in the enclosed Notice of 

Nonconformance. We will consider extending the response time if you can show 

good cause for us to do so. 

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

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

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

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

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

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

discuss them with you. 

Sincerely, 

Gregorfk.&wal ina, Acting Chief 

Special Inspection Branch 

Division of Inspection and Support Programs 

Off ice of Nuclear Reactor Regul at i on 

Docket No.: 99900603 

Enclosures : 1. Notice of Vi ol at i on 

2. Notice of Nonconformance 

3. Inspection Report 99900603/96-01 

36 
NOTICE OF VIOLATION 

Factory Mutual Engineering Association 

Norwood, MA 02062 

Docket No.: 99900603 

Report No.: 96-01 

During an NNC inspection conducted at your Bala-Cynwyd, PA and Norwood, MA 

facilities on April 22 through 26, 1996, a violation of NRC requirements was 

identified. 

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

In accordance with the "General Statement of Pol icy and Procedure 

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

regulations shall adopt appropriate procedures to ensure the evaluation of 

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

report if the evaluation can not be completed within 60 days, and the 

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

to a substantial safety hazard within 5 working days of completing the 

eval uat i on. 

Contrary to the above, Factory Mutual Engineering (FMEA) Contract Procedure CP 

110.00, "Title 10 Code of Federal Regulations Part 21," dated January 19, 

1996, which defines FMEA's policy for compliance with 10 CFR Part 21 did not: 

(1) provide for evaluation of deviations identified as a result of FMEA's 

activities at nuclear facilities and at other covered entities that 

manufacture basic components, (2) ensure that, if an evaluation of deviations 

can not be c:ompleted within 60 days, an interim report is submitted to the 

Commission, or (3) ensure that a director or responsible officer of FMEA is 

informed within 5 working days of the completion of the evaluation. 

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

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

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

ATTN: Document Control Desk, \!ashington, D.C. 20555 with a copy to the Chief, 

Special Inspection Branch, Division of Technical Support , Office of Nuclear 

Reactor Regulation, within 30 days of the date of the letter transmitting this 

Notice of Violation (Notice). This reply should be clearly marked as a "Reply 

to a Notice of Violation" and should include for each violation: (1) the 

reason for the violation, or, if contested, the basis for disputing the violation, (2) the corrective steps that have been taken and the results achieved, 

(3) the corrective steps that will be taken to avoid further violations, and 

(4) the date when full compliance will be achieved. Yo 

reference or include previ bus docketed correspondence , adequately addresses the required response. Where good 

consideration will be given to extending the response t 

Under the authority of Section 182 of the Act, 42 U.5.C 

shall be submitted under oath or affirmation. 

r response may 

f the correspondence 

cause is shown, 

me. 

2232, this response 

Enclosure 1 37 
Because your response will be placed in the NRC Pub1 ic Document Room (PDR) , to 

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

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

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

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

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

withholding the information from the pub1 ic. 

Dated at Rockvi 11 e, Mary1 and 

this 28th day of May, 1996 

2 

38 
NOTICE OF NONCONFORMANCE 

Factory Mutual Engineering Association, Inc. Docket No.: 99900603 

Norwood , Massachusetts Report No. : 96-01 

Based on the results of an NRC inspection conducted on April 22 through 26, 

1996, it appears that certain of your activities were not conducted in 

accordance with The American Society of Mechanical Engineers (ASME) Boiler and 

Pressure Vessel Code (ASME Code) requirements that are applicable to your 

activities under the scope of the Certificate of Accreditation issued to 

Arkwright Mutual Insurance Company dba Factory Mutual Engineering Association. 

A. Paragraph NCA-5220, "Categories of Inspector's Duties, I' of Subsection 

NCA, "General Requirements for Division 1 and Division 2," of Section 

111, "Rules for Construction of Nuclear Power Plant Components," 

(Section 111) of the ASME Code, requires, in part, that the Authorized 

Nuclear Inspector (ANI) perform all duties specifically required in ASME 

N626-1990, "Qualifications and iluties for Authorized Nuclear Inspection 

Agencies and Personnel," and addenda through N626b-1992. 

Section 0-3, "The Authorized Nuclear Inspector," Subsection 0-3.2, 

"Duties," Paragraph 0-3.2.18 of ASME N626b-1992, required, in part, that 

the ANI: shall keep a bound diary of activities and inspections made. 

The information to be recorded shall include a description of the item 

inspected, the type of observation made, the requirements that prompted 

the activity, and the results of inspection. 

Subsect 

Factory 

106.00, 

ion 0, "Documentation," paragraph 3, "Inspector's Diary," of 

Mutual Engineering Association (FMEA) Contract Procedures (CP) 

"ASME Code Activity - Authorized Inspector (AI/ANI) ,Ii Revision 

0, required, in part, that the information recorded include the 

materials that were checked for identification, documentation verified 

or reviewed, the fit-ups and dimensional checks made, procedures 

verified or reviewed, aual ifications verified or reviewed, the number of 

the QA/QC Program Monitoring Report, and a brief description of the 

monitoring activity and any deficiency. 

Contrary to the above requirements, the ANI'S diary entries for 

inspection activities performed at Amer Industrial Technologies 

during June 1993 through June 1995 did not, for the most part, 

the inspection results, the type of observations made, and the 

requirements that prompted the activity. (96-01-01) 

the 

(AW 

ncl ude 

B. Paragraph NCA-5290, "Certification of Data Reports and Construction 

Reports," of Subsection NCA of Section I11 of the ASME Code required, in 

part, that the appropriate Data Reports shall be certified by the ANI 

after satisfying himself that all requirements of Section 111 of the 

ASME Code have been met. 

Enclosure 2 

39 
Subsection 0, "Documentation," paragraph 2, "Manufacturer's Data 

Report," of FMEA CP 106.00, "ASME Code Activity - Authorized Inspector 

(AI/ANI)," Revision 0, required, in part, that the ANI shall verify that 

all other applicable ASME Code requirements have been satisfied before 

signing the Data Reports. 

Contrary to the above requirements, Section V, "Close Out of QA/QC 

Program Monitoring Report," (MR) of FMEA CP 106.00 failed to establish 

adequate measures that ensure that a1 1 MRs that document unsatisfactory 

conditions have been resolved by the Certificate Holder to the ANIS 

satisfaction before the Data Reports are certified by the ANI. 

For example, MR 93-10 written against AIT's job 331 was issued on June 

14, 1993, and identified that (1) the route sheet did not identify all 

the AIT procedures to be used and their revision level, and (2) AIT's 

purchase order for the tubes did not contain all the requirements of its 

QA manual. However, MR 93-10 was not closed by the ANI and AIT until 

May 31, 1995, even though the ANI certified the Data Reports for job 331 

in February, 1993. 

Additionally, FMEA CP 106.00 failed to establish adequate measures for 

the use, control, and distribution of the MRs. (96-01-02) 

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 Inspection 

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

the date of the letter transmitting this Notice of Nonconformance. This reply should be clearly marked as a "Reply to a Notice of Nonconformance" and should 

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

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

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

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

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

consideration will be given to extending the response time. 

Dated at Rockville, Maryland 

this 28th day of May, 1996 

2 

40 
REPORT NO.: 

ORGANIZATION: 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY : 

INSPECTION DATES: 

LEAD INSPECTOR: 

OTHER INSPECTORS: 

REVIEWED BY: 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF INSPECTION AND SUPPORT PROGRAMS 

99900603/96-01 

Factory Mutual Engineering Association (FMEA) 

1151 Boston-Providence Tpke 

Norwood, MA 02062 

Arthur J. Spencer, Manager 

ASME Codes and Standards 

FMEA is a wholly owned and operated subsidiary of 

A1 1 endal e, Arkwright , and Protect ion Mutual Insurance 

Companies and provides authorized inspection services 

under the scope of ASME Certificates of Accreditation 

held by these insurance companies. 

April 22 th,-c:gh 26, 1996 

Vendor Inspection Sectioh (VIS) 

Special Inspection Branch (PSIB) 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Richard P. McIntyre, VIS/PSIB 

Steven M. Matthews, VIS/PSIB 

Vendorqn~ection Section 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

APPROVED BY: 

Division of- Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Enclosure 3 

41 
1 SullllARY OF INSPECTION FINDINGS 

The scope of this inspection was limited to authorized nuclear inspection 

services provided by Factory Mutual Engineering Association (FMEA) under the 

scope of ASME Certificate of Accreditation issued to Arkwright Mutual 

Insurance Company (Arkwright) doing busine5.r as FMEA. More specifically, the 

inspection focused on the Authorized Nuclear Inspector (ANI) and the 

Authorized Nuclear Inspector Supervisor (ANIS) services provided to Amer 

Industrial Technologies, Inc. (AIT) and FMEA's activities related to 

administering and control 1 ing these services urlder their Qual i ty Assurance 

Manual (QAM). 

to AIT, review of indoctrination and training of personnel, internal and 

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

Reuulations, Part 21 (Part 21) requirements. 

The inspection included interviews of the ANI and ANIS assigned 

The inspection basis consisted of the following: 

FMEA QAM, Third Edition-1995, Revision 1. 

American Society of Mechanical Engineers (ASME) Boiler and Pressure 

ASME N626-1990, "Qualifications and Duties ef Authorized Nuclear 

Vessel Code (Code). 

Inspection Agencies and Personnel .'I 

10 CFR Part 21. 

1.1 Violations 

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

regulation must adopt appropriate procedures to ensure evaluation of 

deviations for reportability under this rule, FMEA procedure CP 110.00, "Title 

10 Code of Federal Regulations Part 21," did not provide for evaluation of 

deviations identified as a result of FMEA's activities at nuclear facilities 

and at other covered entities that manufacture basic comDonents. See Section 

0 

2 

3.1 of the report. (99900603/96-01-01) 

1.2 Nonconformances 

Nonconformance 99900603/96-01-02 W~S ident i f i ed and 

Section 3.3.1 of this report. 

Nonconformance 99900603/96-01-03 was identified and 

Section 3.3.1 of this report. 

STATUS OF PREVIOUS INSPECTION FINDINGS 

s discussed in 

s discussed in 

This was the first NRC inspection of FMEA activities performed under an ASME 

Certificate of Accreditation. 

2 

42 
3 INSPECTION FINDINGS AND OBSERVATIONS 

3.1 Entrance and Exit Meetinqs 

The entrance meeting was held on April 22, 1996 at FMEA Philadelphia district 

offices in Bala-Cynwyd, PA. A preliminary exit meeting was held at this 

location on April 24, 1996. The inspection was continued at FMEA Norwood, MA 

offices on April 25, and the inspection findings and observations were 

discussed with FMEA management during the final exit meeting on April 26, 

1996. 

3.2 10 CFR Part 21 Proqram 

The inspectors reviewed FMEA Contract Procedure (CP) 110.00, Revision 0, dated 

January 19, 1996, "Title 10 Code of Federal Regulations Part 21," which 

describes FHEA's policy for achieving compliance with this regulation. 

scope of this procedure addresses actions required to be taken by FMEA 

personnel performing activities at nuclear facilities when a condition 

"adverse" to quality could qualify as a reportable defect or deviation under 

10 CFR Part 21. 

plants and does not address the handling of potentially reportable conditions 

at manufacturing shops (ASME Certificate Holders) serviced by FMEA. 

The 

The procedure is limited to services performed at nuclear 

In summary, CP 110.00 requires FMEA personnel to notify appropriate management 

of potentially reportable conditions unless the FMEA personnel have been 

advised that the plant management (NRC licensee) is already aware of the 

condition and documented evidence exists that NRC has been notified, action 

has or will be implemented to resolve the condition, or the plant's evaluation 

indicates that the situation is not covered under 10 CFR Part 21. The 

procedure also notes that "Evaluations as to whether or not a defect or 

deviation will result in a substantial safety hazard, is beyond the scope of 

responsi bil ii ty of Factory Mutual personnel . " 

The inspectors determined that CP 110.00 did not adequately provide for the 

implementation 10 CFR Part 21. 

(a) The scope of CP 110.00 was limited to situations involving FMEA 

personnel performing activities at nuclear facilities. It did not 

recognize that deviations, as defined in the regulation, may also result 

from FMEA activities related to service contracts at manufacturing shops 

(Certificate Holder facilities) 

The scope of CP 110.00 did not require FMEA to evaluate deviations or 

failures to comply or. to inform the affected purchasers/l icensees when 

such deviations or failures to comply are the direct result of services 

provided by FMEA to Certificate Holders or licensees. For example, the 

procedure did not address situations where the ANI inspection activities 

are found to be in noncompliance with the technical requirements of the 

purchase documents after the performance of the contracted services. An 

example would be an after-the-fact identification that an ANI providing 

services to a nuclear plant or Certificate Holder lacked the required 

qual i fi cat i ons . 

(b) 

3 

43 
Failure to adopt appropriate procedures to evaluate deviations was identified 

as a violation of 10 CFR Part 21.21. (99900603/96-01-01) 

3.3 Qual itv Assurance Prosram ImDlementation 

The FMEA quality assurance program is described in their QAM for Contract 

Inspection Services, Third Edition-1995, Revision No. 1. 

endorsements and authorization statements by senior officeys of the three 

insurance companies that make up the Factory Mutual System and affirms 

compliance with ASME QAI-1, "Qualifications for Authorized Inspection" and 

with ASME Code Section I, 111, IVY VIII, X, and XI activities. The program 

implementation was examined by reviewing selected records and interviewing 

appropriate FMEA staff. The implementation review was focused on ANI and ANIS 

services provided to AIT during the 1993-1995 timeframe. 

The manual contains 

3.3.1 FMEA Activities at AIT 

During the NRC's inspection of AIT, conducted on January 29 through February 

2, 1996 (NRC Inspection Report 99901292/96-01, dated March 21, 1996), the 

inspection team reviewed documentation for four AIT jobs (331, 392, 442, and 

4102) that had been recently completed and shipped to NRC licensees. 

four jobs had been inspected by ANIs employed by FMEA. 

Although the ASME Data Reports for these jobs were certified by ANIs 

indicating that all ASME Code Section I11 requirements had been met by AIT, 

the NRC inspection determined that none of the four jobs completely met all 

ASME Code Section I11 requirements. 

inspection, the team reviewed FMEA's documentation supporting the 

determination that the certified items complied with the applicable ASME Code 

requirements. 

bound diaries that were maintained for all inspections conducted at AIT. 

These 

Therefore, during the course of this 

A significant portion of this review was focused on FMEA ANI 

Based on an evaluation of the bound diaries, the team concluded that the ANIs 

had performed extensive inspection and verification activities from prefabrication through final testing and Data Report certification. A 

significant amount of inspection activity was documented for all four jobs. 

However, the team concluded that the individual entries in the bound diary 

lacked sufficient documentation to substantiate the various and numerous ASME 

Code compliance issues identified and communicated to FMEA management by the 

ANIs regarding AIT's performance. The diaries were typically annotated in 

short, cryptic notes that were not always clear about the specific inspections 

or verifications performed nor documented the results of these inspections or 

verifications. 

review of specific issues with the ANIs during an interview, determined that 

the entries failed to document the inspection results, the type of 

observations made, and the requirements that prompted these activities. 

The team's evaluation of the bound diary entries and its 

Requirements to record the above information are imposed by Subsection 0, 

'Documentation," paragraph 3, "Inspector's Diary," of FMEA CP 106.00, "ASME 

Code Activity - Authorized Inspector (AI/ANI)." These requirements were also 

imposed on FMEA through its ASME accreditation as an Authorized Inspection 

Agency (AJA) in accordance ASME N626-1990, "Qualifications and Duties for 

4 

44 
Authorized Nuclear Inspection Agencies and Personnel, 'I and raragraph NCA-5220, 

"Categories of Inspector's Duties," of Subsection NCA, "General Requirements 

for Division 1 and Division 2," of Section I11 of the ASME Code. 

The team concluded that FMEA's ANI bound diary entries did not meet the 

requirements of its CP 106.00, ASME Code NCA-5220, and N626b-1992. 

demonstrate conformance with the applicable requirements and duties for ANIs 

was identified as Nonconformance 99900603/96-01-02. 

Failure to 

Additionally, during the team's evaluation of the bound diaries, the team 

noted that the ANIs had issued several QA/QC Program Monitoring Reports (MRs) 

in the course of performing inspection activities at AIT. As provided for in 

CP 106.00, the MRs should be completed by the ANIs for (1) any finding not 

completed before the ANI left the facility, or (2) on completing monitoring of 

one or more sections of the QA/QC manual. 

The bound diary contained an entry on June 14, 1993, that MR 93-10 was issued 

on AI1 job 331 which noted that (1) the route sheet did not identify all the 

AI1 procedures used to meet ASME Code requirements and their revision level, 

and (2) AIT's purchase order for the tubes did not contain all the 

requirements of its QA manual. 

The Data Reports for job 331 were certified by the ANI on February 4 and 8, 

1993, indicating that all ASME Code requirements were met by AIT. However, MR 

93-10 was not resolved by AIT or accepted by the ANI until May 31, 1995. 

The team's evaluation of CP 106.00 determined that it failed to require MRs to 

be satisfactorily closed before the ANI certifies the Data Report. Therefore, 

open deficiencies may exist at the time the ANI certifies the Data Report. 

The team also concluded that CP 106.00 failed to establish measures for the 

use, control, and distribution of the MRs, although the team did note that 

distribution was described on the backside of the MR. 

The team concluded that FMEA failed to establish measures to assure that all 

deficiencies identified on MRs were satisfactorily closed before the ANI 

certified the Data Reports. The team also concluded that FMEA had not 

adequately addressed the use, control, and distribution of MRs. Failure to 

establ ish these measures was identified as Nonconformance 99900603/96-01-03. 

3.3.2 Interfaces with Arkwriqht Mutual Insurance ComDanv 

The team determined that the inspection services contract for AIT was actually 

signed with Arkwright Mutual Insurance Company (Arkwright) . The ASME 

Certificate of Accreditation for Arkwright to perform these inspection 

services as an accredited AIA states the following: 

Arkwright Mutual Insurance Company 

Mutual Boiler Division 

225 Wyman Street 

Waltham, Massachusetts 02254-9198 

dba Factory Mutual Engineering Assnci at ion 

5 

45 
Review of mrrespondence files related to the AIT contract indicated that 

FMEA’s ANIs assigned to AIT, its District Office in Bala-Cynwyd, Pennsylvania, 

and the FMEA Home Office in Norwood, Massachusetts, had, since 1993, 

repeatedly advised Arkwright that the inspection services contract with AIT be 

dropped. According to FMEA, its request to drop AIT’s inspection services was 

based, in part, on AIT’s consistently poor performance in achieving ASME Code 

compliance. However, Arkwright was not receptive to FMEA’s repeated request 

and did not cancel the contract with AIT until 1995. 

The ANI’s responsibilities are considered by the ASME Code to be third party 

oversight of a Certificate Holder that is assumed to diligently comply with 

the ASME Code requirements. FMEA’s correspondence files reviewed during the 

inspection, however, identified that the normal assumptions and principles of 

third party oversight did not apply at the AIT fabrication facility. 

documentation and interviews with the ANI and ANIS indicated that during the 

period from 1993 through 1995, the ANIs had to exercise extreme diligence in 

the performance of inspections and verifications because the lack of ASME Code 

compliance was pervasive throughout AIT’s fabrication activities. 

September 2, 1995, correspondence to Arkwright, FMEA’s management noted that 

the services contract with AIT was to provide qualified inspectors to verify 

compliance rather than to provide shop quality control personnel and suggested 

that, if quality control services were to be provided, a separate agreement should be developed to cover such activity. 

The 

In 

After reviewing FMEA’s reports, records, ANI bound diaries and interviewing 

the ANIs the team concluded that the ANIs had very little confidence in AIT’s 

ability to comply with ASME Code requirements. Therefore, the basic 

assumption that their oversight activities were adequate to verify with 

reasonable assurance that every aspect of ASME Code compliance had been 

achieved by AIT was no longer valid. Although FMEA’s management reiterated to 

the ANIs not to certify Data Reports where they knew that ASME Code compliance 

was not met, the ANI’s could not be expected to verify every detail of the 

manufacturing process and, under the existing conditions, it was not 

reasonable for the ANI’s to assume that all details that were not verified 

compl i ed wi t h ASME Code. 

The previously discussed NRC findings during its inspection of AIT, described 

in NRC Inspection Report 99901292/96-01, fully support the teams conclusion 

that the ANIs could neither verify every detail of the manufacturing process 

nor assume that those details not verified complied with the ASME Code. 

During an interview with Arkwright’s management at FMEA’s Norwood, MA offices, 

the team briefly discussed Arkwright’s basis for maintaining the contract with 

AIT, and how its control of the AIT services contract met the ASME Certificate 

Factory Mutual Engineering Association when, in fact, FMEA did not have the 

authority to cancel the contract. 

personnel to adequately address these issues, the team determined that it 

would be appropriate to develop more complete information during a future 

follow up inspection of Arkwright. 

I of Accreditation provision that Arkwright was accredited doing business as 

Because of unavai 1 abi 1 i ty of key Arkwright 

6 
3.3.3 wits and InsDector Performance Evaluations 

The inspectors reviewed QAM Section 9, "Audits," and CP 108.00, "ASME National 

Board Audits and Evaluations," Revision 0, dated January 19, 1996, which 

define the audit and evaluation responsibilities for the ANIS and Authorized 

Nuclear Inservice Inspector Supervisor (ANIIS) at the FMEA District Offices 

and as well as the audit responsibilities for Home Office Supervisor. 

FMEA audits program is written to meet the requirements of ASME QAI-1. This 

program includes audit methods, ASME audit reports of nuclear Section I11 and 

XI work, authorized inspector performance evaluations and the audit schedule 

for each. 

Certificate holders facilities prior to the actual ASME survey to ensure that 

the Certificate Holder is prepared for the ASME survey team review. 

The 

The ANIS also conducts the ASME Pre-Survey audit at the ASME 

The inspectars reviewed the most recent Philadelphia and New Jersey District 

Office l i s t i n g o f ASME Section I11 shops and Section XI sites that also 

included information such as the ANI and ANII assignments, the ANIS semiannual 

shop and site audit schedule, and the ANIS schedule for ANI performance 

evaluations at the Section I11 nuclear shops and ANII evaluations at the 

Section XI nuclear reactor sites it; 3xh district. The ASME Section I11 shop 

audits and Section XI site audits are performed to review ASME Code compliance 

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

FMEA check lists (Form 2048 for Section 111 shops and Form 1453 for Section XI 

sites). 

both performed semiannually. 

The shop and site audits and inspector's performance evaluations are 

The inspectors also reviewed the results of the three most recent years of 

ANIS shop audits conducted at three ASME Section 111 Code shops; Amer 

Industrial Technology (AIT) , Valcor Engineering Corporation and Yarway 

Corporation. The audit reports included the appropriate documentation 

requi red by FMEA procedures i ncl udi ng audi t f i ndi ngs when appl i cab1 e. However 

the inspectors noted that the ANIS audits of AIT for 1993 to 1995 continually 

identified AIT's failure to effectively implement their QA program and to meet 

ASME code requirements, which resulted in unacceptable results in many of the 

areas reviewed. In fact, in the Comments and Recommendations section of the 

audit report checklist (Form 2048), the ANIS usually recommended to the home 

office supervisor that the contract with AIT be terminated. 

The inspectors a1 so reviewed the semi annual inspector's performance 

evaluations conducted by the ANIS during his shop audits at the above 

mentioned Section I11 shops. The ANI and ANII inspector's performance 

evaluations are also conducted utilizing standard FMEA checklists (Form 2051 

for ANIS at Section I11 shops and Form 2050 for ANIIs at Section XI sites). 

All inspector evaluations were performed and documented as per FMEA QA program 

requirements. 

3.3.4 Personnel Qualification and Training 

The inspectors reviewed QAM Section 8, "Indoctrination and Training," which 

describes the general requirements for the indoctrination and training of 

7 

47 
personnel and the more specific training requirements and responsibilities for 

the Home Office Supervisor and the District Supervisors. 

Code Activity - Authorized Supervisor (AIS/ANIS/ANIIS) ,'I CP 106.00, "ASME Code 

Activity - Authorized Inspector (AI/ANI) ,I' and CP 107.00, "ASME Code Activity 

- Authorized Inspector (ANII)," all Revision 0, dated January 19, 1996, 

describe the specific training responsibilities and required training courses 

for each of the identified positions. 

CP 105.00, "ASME 

The Philadelphia district ANIS maintains a training matrix of job functions 

versus training requ'rements. Th2 inspectors reviewed the indoctrination and 

training records for several employees currently working out of the 

Philadelphia district office as ANIs and ANIIs. 

maintained separately for ASME training and all other training. The training 

records reviewed by the inspectors included the appropriate documentation for 

the qualification and training that required for ANIs and ANIIs per ASME QAI-1 

and the FMEA quality program. The inspectors also verified through review of 

records that the Philadelphia district office ANIS had the appropriate 

qualifications and training as required by ASME QAI-1 and the FMEA quality 

program. 

Training records are 

3.3.4.1 Lead Auditor Oualifications 

The inspectors reviewed QAM Section 9.4, "Selection of Auditors," and CP 

109.00, "Lead Auditor - Authorized Nuclear supervisors (ANIS/ANIIS) ,I' Revision 

0, dated January 19, 1996, which outlines the minimum qualifications to 

certify and designate lead auditors for ASME and district audits. 

auditor qualifications are maintained at the FMEA home office in Norwood, 

Massachusetts. 

record files at the Norwood office for two ANIS/ANIISs, the Assistant Vice 

President, ATS, and the Manager of ASME Code and Standards, FMEA and verified 

that these individuals had attained and were maintaining the appropriate 

qualifications to be certified as a lead auditor. 

President, ATS, conducts the annual evaluation of the QA program at the FMEA 

home office and the Manager of ASME Code and Standards, FMEA, conducts the 

annual evaluation of the district ANIS/ANIIS's performance. 

The lead 

The inspectors reviewed the qualification and certification 

The Assistant Vice 

4 PERSONS CONTACTED 

The persons contacted during this inspection are listed below: 

Factory Hutual Engineering Associ at i on 

* 

* + A.J. Spencer, Manager, ASME Codes and Standards 

S. Rudnickas, Vice President & Manager B&M Engineering 

R.E. Montague, Asst. Vice President-District Manager 

+ W.R. Rogers 111, District Supervisor + R.G. Edl, District Chief Inspector 

D. Kinley, Authorized Nuclear Inspector 

8 

48 
Arkwri ght Technical Services, Inc. 

* C.M. D’Esopo, President 

+ Attended entrance meeting 

* Attended exit meeting 

9 

49 
I 

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20566-0001 

May 10, 1996 

Mr. James E. Quinn, Projects Manager 

LMR and SBWR Programs 

GE Nuclear Energy 

175 Curtner Avenue, M/C 165 

San Jose, CA 95125 

SUBJECT: NRC INSPECTION REPORT NO. 99900403/96-01 

Dear Mr. Quinn: 

This letter addresses the inspection conducted March 5 through 8, 1996, at the 

Paul Scherrer Institut (PSI) PANDA Test Facility in Wurenlingen, Switzerland, 

by Richard P. McIntyre of the Nuclear Regulatory Commission's (NRC's) Special 

Inspection Branch, Juan D. Peralta of the Quality Assurance and Maintenance 

Branch, John A. Kudrick and John D. Monninger of the Containment Systems and 

Severe Accident Branch, Alan E. Levin of the Reactor Systems Branch, and Dino 

C. Scaletti of the Standardization Project Directorate. The details of the 

inspection were discussed with you and the members of your staff during the 

inspection and at the exit meeting on March 8, 1996. 

The purpose of the inspection was to determine if testing activities performed 

at the PANDA test facility to support design certification of the GE Nuclear 

Energy (GE) simp1 ified boiling water reactor (SBWR) design were conducted 

under the appropriate provisions of the May 1990, GE NEDG-31831, "SBWR Design 

and Certification Program Quality Assurance Plan," as implemented by GE 

document PPCP-QA-01, "PANDA Project Control Plan" (PPCP) and the GE PANDA 

Quality Assurance Procedures (PQAPs). 

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 GE, in general, was adequately 

implementing the Project Control Plan and the Quality Assurance Procedures for 

testing activities performed at PANDA with the exception of two 

nonconformances. Also, the team identified three unresolved items related to 

PANDA and' SBWR design certification that will require response by GE and 

follow-up by the NRC during a future inspection at San Jose. 

the inspection team identified Nonconformances with program implementation 

with respect to (1) the preparation and issue of Apparent Test Results Reports 

and Data Transmittal Reports as xquired by the PANDA Test Specification and 

PANDA Test Plans, and (2) the failure to document abnormal occurrences 

detected during testing (subsequently causing matrix testing to be suspended 

and re-evaluated) using the existing nonconformance report process. 

Specifically, 

I 50 
3. Quinn -2- 

The unresolved items concerned (1) the appropriateness of GE's acceptance of 

engineering services activities performed by El ektrowatt Ingeni eurunternehmung 

AG (EJektrowatt) in October 1993 for the PANDA test facility as-built 

measurements, (2) the level of GE QA oversight for the engineering services 

work performed by the international technical associates (KEMA and Insti tuto 

de Investigaciones (IIE) of Mexico) for PANDA data analysis, and (3) the 

disposition for the recommendations and specific action items identified 

during the October 1991 PANDA Design Review regarding facility design, quality 

assurance programmatic aspects, and technical issues. 

The response requested by this letter is 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 , 

/&+a<+ Special nspection Branch 

fl Division of. Inspection and Support Programs 

Office of Nucl ear Reactor Regul at i on 

Docket No. : 52-004 

Enclosures: 1. Notice of Nonconformance 

2. Inspection Report No. 99900403/96-01 

cc w/encls: See Next Page 
Mr. James E. Quinn 

GE Nuclear Energy 

cc: Mr. Rob Wallace 

GE Nuclear Energy 

1299 Pennsylvania Avenue, N.W. 

Suite 1100 

Washington, DC 20004 

Director, Criteria & Standards Division 

Office of Radiation Programs 

U.S. Environmental Protection Agency 

401 M Street, S.W. 

Washington, DC 20460 

Mr. Sterling Franks 

U.S. Department of Energy 

Washington, DC 20585 

NE-42 

Mr. John E. Leatherman, Manager 

SBWR Design Certification 

GE Nuclear Energy 

175 Curtner Avenue, MC-781 

San Jose, CA 95125 

Mr. Steven A. Hucik 

GE Nuclear Energy 

175 Curtner Avenue, MC-780 

San Jose, CA 95125 

Docket No. 52-004 

Mr. Brian McIntyre 

West i nghouse Electric Corporation 

Energy Systems Business Unit 

Box 355 

Pittsburgh, PA 15222 

Mr. Tom J. Mulford, Manager 

SBWR Design Cert i f i cat i on 

Electric Power Research Institute 

3412 Hillview Avenue 

Palo Alto, CA 94304-1395 

52 
NOTICE OF NONCONFORMANCE 

GE Nuclear Energy 

San Jose, CA 95125 

Docket No.: 52-004 

99900403/96-01 

Based on the results of a Nuclear Regulatory Commission (NRC) inspection, 

conducted from March 5 through March 8, 1996, of the ;E's PANDA test program 

at the Paul Scherrer Institut (PSI) in Wurenlingen, Switzerland, related to 

the SBWR design certification activities, it appears that certain activities 

were not conducted in accordance with NRC requirements. 

A. Criterion XI, "Test Control," of Appendix B to 10 CFR 50, requires, in 

part, that test results be documented and evaluated to assure that test 

requirements have been sati sf i ed. 

Chapter 11, "Basic Requirements," Section 11, "Test Control ,I1 of 

ANSI/ASME NQA-1-1983, "Qual i ty Assurance Program Requirements for 

Nuclear Facilities," requires, in part, that test results be documented 

and that their conformance with acceptance criteria be evaluated. 

Paragraph 5.3.14 of PQAP-TC, "Test Control ,I' Revision 3, dated September 

18, 1995, requires that the PSI PANDA Project Manager (P-PM) prepare 

test reports per the Test Specification and Test Plan requirements. 

Section 11, "Reporting," of GE Document 25A5587, "PANDA Test 

Specification," Revision 1, dated January 26, 1995, requires (1) 

preparation of an Apparent Test Results report within approximately one 

week following performance of the test, and (2) preparation of Final 

Test Reports per the schedule specified in the Test Plan and Procedures 

Document . 

Section 10, "Reports," of GE Document 25A5764, "PANDA Test Plan - Tests 

M3, M3A, M3B, M4, M7," Revisions 1, 2, and 3 dated September 18, 1995, 

October 16, 1995, and November 15, 1995, respectively, requires (1) 

preparation of an Apparent Test Results report within approximately two 

weeks of completion of each transient integral system test, and (2) 

preparation of a Data Transmittal Report approximately two months after 

the last test is performed. 

Section 10, "Reports," of GE Document 25A5785, "PANDA Test Plan - Tests 

M2, MOA, MlOB," dated November 21, 1995, requires (1) preparation of an 

Apparent Test Results report within approximately two weeks of 

completion of each transient integral system test, and (2) preparation 

of a Data Transmittal Report approximately two months after the last 

test is performed. 

Section 10, "Reports," of GE Document 25A5788, "PANDA Test Plan - Tests 

M6/8," dated December 7, 1995, requires (1) preparation of an Apparent 

Enclosure 1 

53 
lest Results report within approximately two weeks of completion of each 

transient integral system test, and (2) preparation of a Data 

Transmittal Report approximately two months after the last test is 

performed. 

Section 10, "Reports," of GE Document 25A5824, "PANDA Test Plan - Test 

M9," dated December 12, 1995, requires (1) preparation of an Apparent 

Test Results report within approximately two weeks of completion of each 

transient integral system test, and (2) preparation of a Data 

Transmittal Report approximately two months after the last test is 

performed. 

Contrary to the above, Apparent Test Results reports and Data 

Transmittal Reports were not prepared and issued in accordance with the 

Test Specification or Test Plan and Procedures. (96-01-01) 

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

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

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

type appropriate to the circumstances and that such activities be 

accomplished in accordance with these instructions, procedures, or 

drawings . 

Criterion XV, "Nonconforming Materials, Parts, or Components," of 

Appendix B to 10 CFR 50, requires, in part, that nonconforming items be 

reviewed and accepted, rejected, repaired or reworked in accordance with 

documented procedures. 

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

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

conditions adverse to quality, such as failures, malfunctions, 

deficiencies, deviations, defective material and equipment, and 

nonconfomances are promptly identified and corrected. 

Chapter 11, "Basic Requirements," Section 11, "Test Control ,'I of 

ANSI/ASME NQA-1-1983, "Quality Assurance Program Requirements for 

Nuclear Facilities," requires, in part, that test results be documented 

and that their conformance with acceptance criteria be evaluated. 

Paragraph 5.3.11 of PQAP-TC, "Test Control," Revision 3, dated September 

18, 1995, requires that the PSI PANDA Project Manager (P-PM) (1) review 

and resolve all test anomalies identified during the test, and (2) 

document resolutions, conditions requiring correction, and corrective 

actions per PQAP-NC. 

Section 4, "Requirements," of PQAP-NC, "Nonconformance Control and 

Corrective Action," Revision 0, dated January 31, 1995, provides that 

any nonconforming item which can affect PANDA test results, or 

deviations from the test specification/procedure, or test conditions and 

resul ts showi ng abnormal occurrences shall be i dent i f i ed, treated as a 

nonconformance, and documented and reported for resolution (disposition) 

prior to continuation of subsequent phase testing. 

54 
Contrary to the above, (1) when abnormal occurrences (subsequently 

causing matrix testing to be suspended and re-eval uated) were detected 

during testing, no nonconformance reports were generated to document 

these events; (2) PSI Procedure "Data Base Modification" (issued in 

March 1996) was being used by PSI testing personnel to perform 

activities that introduced deviations from the test control process 

already specified by PQAP-TC, and from the nonconformance identification 

process established in PQAP-NC; and (3) PSI Procedure "Data Base 

Modification'' had not been identified or described as a Quality 

Assurance Procedure governed by PPCP-QA-01 , i .e. , as a procedure 

comprising the bases of the QA system implemented by PSI and GE in 

meeting the requirements of NEDG-31831, even though it was being used to 

perform quality related activities affecting PANDA test results. 

(96-01-02) 

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 Inspection 

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

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

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

include for each nonconformance: (1) a description of the steps that were or 

will be taken tu 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 Rockvi 11 e, Mary1 and 

This 10 th clay of my 1996 

55 
ORGAN I ZAT I ON : 

REPORT NO. : 

CORRESPONDENCE 

ADDRESS : 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY: 

INSPECTION 

CONDUCTED: 

TEAM LEADER: 

OTHER INSPECTORS: 

REV I EWED : 

APPROVED: 

GE Nuclear Energy 

San Jose, Cal i forni a 

99900403/96-01 

Mr. James E. Quinn, Projects Manager 

LMR and SBWR Programs 

GE Nuclear Energy 

175 Curtner Avenue 

San Jose, California 95125 

Mr. Kenneth W. Brayman, Manager 

Quality Assurance Systems 

(408) 925-6587 

GE Nuclear Energy (GE) is engaged in the supply of 

advanced boiling water reactor designs to utilities. 

GE also furnishes engineering services, nuclear 

replacement parts, and dedication services for 

commercial grade electrical and mechanical equipment. 

March 5 through 8, 1996 

5h [46 Richard P. McIntyre Date 

Vendor Inspection Section (VIS) 

Special Inspection Branch (PSIB) 

Juan D. Peralta, HQMB 

John A. Kudrick, SCSB 

John D. Monninger, SCSB 

Alan E. Levin, SRXB 

Oino C. Scaletti, PDST 

Grggory @a1 ina, Section Chief, VIS 

- 

INSPECTION BASES: 10 CFR Part 50, Appendix 6 and 10 CFR Part 21 

INSPECTION SCOPE: To determine if activities performed to support the 

design of the SBWR and, specifically, testing 

activities performed at the PANDA Test Facility at the 

Paul Scherrer Institut in Wurenl ingen, Switzerland 

were conducted under the appropriate provisions of the 

May 1990, GE NEDG-31831, "SBWR Design and 

Certification Program Quality Assurance Plan." 

PLANT SITE 

APPLICABILITY: None 

Enclosure 2 

56 
1 INSPECTION SUMMARY 

1.2 

e 

e 

1.2 

e 

e 

e 

Nonconformance 

Nonconformance 99900403/96-01-01 was ident 

Section 3.4.1 of this report. 

Nonconformance 99900404/96-01-02 was i dent 

Section 3.8 of this report. 

Unresolved Item 

fied and is discussed in 

fied and is discussed in 

Unresolved Item 99900403/96-01-03 was identified and is discussed in 

Section 3.2 of this report. 

Unresolved Item 99900403/96-01-04 was identified and is discussed in 

Section 3.3.2 of this report. 

Unresolved Item 99900403/96-01-05 was identified and is discussed in 

Section 3.4.2 of this report. 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

No previous inspections have been conducted at this test facility. 

3 INSPECTION FINDINGS AND OTHER COMMENTS 

3.1 GE SBWR Oualitv Assurance Proaram 

Chapter 17 of the SBWR standard safety analysis report (SSAR) describes the GE 

quality assurance (QA) program for the design phase of the SBWR program. 

QA program i s identified as "Nucl ear Energy Business Operations Qual i ty 

Assurance Program Description, I' NEDO-1 1209-04AY Revision 8, the 1 atest 

revision approved by the NRC. NEDO-11209-04A applies to all GE activities 

affecting quality of items and services supplied to nuclear power plants and 

establishes GE's compliance with the provisions of Appendix B to 10 CFR 50. 

NEDG-31831, "SBWR Design and Certification Program Quality Assurance Plan," 

dated May 1990, was developed by GE to fulfill the QA requirements of the SBWR 

reactor design and certification program. NEDG-31831 meets the requirements 

of ANSI/ASME NQA-1-1983 and its NQA-la-1983 addenda as endorsed by the NRC in 

Regulatory Guide 1.28, Revision 3. Additionally, NEDG-31831 provides that 

design and testing work performed by international technical associates will 

be performed to their internal QA programs acceptable to the regulatory 

authorities of their respective countries as evaluated by GE for compliance 

with the provisions of ANSI/ASME NQA-1-1983. 

The 

-2- 

57 
3.2 PANDA OA Program for SBWR Design Certification Testinq 

Under an agreement between the Swiss Confederation (represented by the Paul 

Scherrer Institut [PSI]), the Electric Power Research Institute (EPRI), and 

GE, PSI performed passive decay heat removal and fission product retention 

tests in the PANDA test facility. 

the performance and behavior of the SBWR passive containment cooling system 

(PCCS) operating in typical ?est-LOCA containmeqt environments. 

were primarily focused on simulating the response of the SBWR containment 

cooling systems in order to (1) obtain additional data to support the adequacy 

of TRACG in predicting the quasi-steady heat rejection rate of a PCC heat 

exchanger and identify the effects of scale on PCC performance, (2) provide a 

sufficient database to confirm the capability of TRACG to predict SBWR 

containment system performance, encompassing systems interaction effects, and 

(3) demonstrate startup and long-term operation of a passive containment 

cooling system (Concept Demonstration). 

The purpose of these tests was to evaluate 

These tests 

A GE readiness review conducted at PSI during October 19 through 21, 1994, 

concluded that PSI had not adequately implemented a QA program meeting the 

appropriate requirements of ANSI/ASME NQA-1-1983, and that the PANDA facility 

was not ready to initiate testing. In a letter to GE, dated Oecember 19, 

1994, the NRC staff requested that GE provide a discussion of the corrective 

actions taken by GE as a result of the readiness review findings, including 

the area of QA. 

In its response letter to the NRC, dated March 7, 1995, GE stated that as a 

result of a GE readiness review at PSI during October 1994, the PANDA quality 

assurance program would be restructured so that it would be conducted under 

direct GE supervision and governed by the provisions in NEDG-31831. To this 

effect, GE developed the PANDA Project Control Plan, PPCP-QA-01, Revision 1, 

dated May 1, 1995. PPCP-QA-01, in conjunction with nine other QA procedures, 

describe the organization, quality related activities, events and procedures 

necessary to ensure and verify that the PANDA project at PSI is conducted in 

accordance with the provisions of NEDG-31831. 

the PANDA test facility and test results is contained in the PANDA Test File 

(PTF) and organized accordingly. 

All documentation related to 

In accordance with the provisions of PPCY-QA-01 and PQAP-TC, "Test Control ,'I 

Revision 3, dated September 18, 1995, a test specification, GE Document No. 

25A5587, "PANDA Test Specification," Revision 1, was prepared and issued by GE 

as required by the provisions in Engineering Operating Procedure (EOP) 

35-3.00, "Engineering Tests." GE Document No. 25A5587 required that the PANDA 

tests be performed in conformance with PPCP-QA-01, which is based on the 

requirements of Appendix B to 10 CFR 50, NEDG-31831, and ANSI/ASME NQA-1-1983. 

During the inspection, the team reviewed a1 1 relevant documentation and 

available test data found in the PTF. Based on these reviews, the team 

concluded that, in general, GE had adequately restructured the PANDA QA 

program in accordance with the provisions in NEDG-31831. However, GE failed 

to adequately implement certain provisions of NEDG-31831, prior to January 

1995, related to the appropriateness of GE's acceptance of engineering 

-3- 

58 
services provided by a subcontractor to PSI. Specifically, the team 

determined that GE had not yet adequately addressed an issue related to 

activities performed at PSI by Elektrowatt Ingenieurunternehmung AG 

(Elektrowatt) in October 1993 and which had been identified during the October 

19 through 21, 1994, readiness review at PANDA. Elektrowatt was hired by PSI 

to perform the facility as-built measurements which was an activity having 

substantial impact on the quality of test results generated at PANDA. 

The team was concerned that GE concluded in the October 1994 Readiness Review 

Report (without providing any justification or taking any compensatory or 

corrective actions) that the PANDA facility as-built measurement activities 

performed by Elektrowatt were satisfactory, while at the same time 

acknowledged that Elektrowatt had not been audited by either GE or PSI as a 

supplier of services affecting quality. Unresolved Item 99900403/96-01-03. 

This issue was identified as 

3.3 Desi qn Control 

The purpose of the review of design control was: (1) to assure that applicable 

regulatory requirements, design bases, codes and standards, and GE test 

specification requirements were correctly trans1 ated into design drawings , procedures, and instructions per PQAP-DC, "Document Control, I' (2) to assure 

that changes or deviations from specified design requirements and quality 

standards were identified, documented, and controlled, (3) to verify final 

PANDA test facility as-built drawings and overall control of test facility 

configuration as described in PQAP-V, "Verification," and (4) to assure that 

computer data acquisition software and documentation was controlled as 

described in PQAP-DA, "Data Acquisition System Control." 

The team reviewed the following material related to design control for the 

PANDA test program: 

0 Design and as-built drawings 

0 PANDA scal ing analyses 

0 PANDA line loss calculatioris based on estimated SBWR line losses 

0 Record of GE Design Review (San Jose, October 1991) 

The PANDA test facility at PSI was designed to evaluate the performance of the 

SBWR passive containment cooling system operating in post-LOCA containment 

environments. The PANDA tests were to demonstrate the SBWR thermal-hydraulic 

performance, heat removal capability, and systems interactions and to provide 

data for confirmation of the TRACG computer models used to analyze the SBWR 

performance. 

GE prepared and issued document 25A5587, "PANDA Test Specification," Revision 

1, on January 26, 1995, for PANDA tests. The PANDA test specification 

specifies the top-level requirements for tests related to post-LOCA decay heat 

removal from the containment of the SBWR to be performed at the PANDA test 

facility at PSI. The test specification provides gmeral criteria for the 

PANDA test program including: purpose, objectives, facility description, test 

instrumentation, data acquisition system and recording, data processing and 

analysis, shakedown and plant characterization tests, steady-state performance 

-4- 

59 
tests, transient integral systems tests, pretest predictions, acceptance 

criteria, reporting requirements, record retention, and quality assurance. 

The actual experimental testing at PSI was to be performed in accordance with 

the Test Specification (25A5587) through the development of specific Test 

P1 ans and Procedures in accordance with GE PANDA Qual i ty Assurance Procedure 

PQAP-DC, Revision 1, "Document Control .'I PUAP-DC defines the requirements and 

process for issuing, revising, modifying, and distributing the Test Plans and 

Procedures. 

GE and PSI prepared the Test Plans and Procedures, which define the detailed 

or specific test requirements. 

set up and performed to meet the quality assurance requirements, any special 

conditions associated with the test, and the test requirements specified in 

the Test Specification. 

required to perform the test in accordance with test and quality assurance 

requirements. 

the course of the inspection are listed in the table below for the plant 

characterization, shakedown, steady-state, and integral systems tests. 

The test plans describe how the test is to be 

The test procedures describe the specific procedures 

The specific test plans and procedures reviewed by the team in 

TEST TEST PLAN TEST PROCEDURE 

n VESSEL HEAT LOSS ALPHA-510 ALPHA-5 10 

LINE PRESSURE DROP ALPHA-510 ALPHA-5 10 

Sl-S6 ALPHA-410-1 ALPHA-410-1 

s7-s9 ALPHA-410-2 ALPHA-410-2 

S 10-S 13 ALPHA-410-2 ALPHA-410-2 

M3 25A5764 R1 ALPHA-520-0 r ~~~ 

M3A 25A5764 R2 ALPHA-520-2 

M3B 25A5764 R2 ALPHA-520-2 

M7 25A5764 R3 ALPHA-521-0 

M2 25A5785 RO ALPHA-527-0 

MlOA 25A5785 RO ALPHA-527-0 

MlOB 25A5785 RO ALPHA-527-0 

25A5788 RO ALPHA-529-0 

25A5824 RO ALPHA-528-0 

These Test Plan and Procedures were issued and controlled in accordance with 

PQAP-DC. A PANDA Engineering Review Memorandum (P-ERM) was required for 

I review and approval o f a l l Test Plans and Procedures by PQAP-DC. In the 

-5- 
course of the inspection, the team reviewed several completed P-ERMs relating 

to revisions to the following Test Plan and Procedures: 

e 

e 

ALPHA-410 "PANDA Steady-State PCC Performance Tests Test Plan and Test 

Procedures" 

ALPHA-520 "PANDA Transient Tests M3A, M3B, & M4 Integral System Test 

Procedure" 

Based on a review of these P-ERMs, the team concluded that issues and comments 

identified by GE and PSI personnel as a result of the review and approval 

process of the Test Plan and Procedures were adequately identified, 

documented, resolved, and control 1 ed. 

GE PANDA Qual i ty Assurance Procedure PQAP-V , Revi si on 1, "Veri f i cation, It was 

developed to control the process for verification of the PANDA test facility 

configuration and testing activities. Verifications were to be performed for 

activities such as: calculations affecting test results, measurements 

appearing on as-built drawings, and test initial conditions. PQAP-V provided 

a "Verification Cover Sheet" to control and document the verification process. 

Extensive documentation was contained in the PTF on scaling of the PANDA 

facility and determination of line losses for the facility, based on design 

information for the SBWR. During the course of the inspection, the team 

reviewed an independent verification that required an a1 ternate calculation to 

be performed to verify the correctness of the original calculations. 

verification related to the establishment of the PANDA system line loss 

coefficient measurements. The line loss calculations were performed by 

several engineers using different methodologies; the results were then crosschecked and independently design verified and documentation of the results of 

this design study is extensive. 

performed in accordance with PQAP-V and utilized the Verification Cover Sheet 

for control and approval of the 'verification. 

This 

The team noted that the verification was 

3.3.1 Data Acquisition System (DAS) 

The team evaluated the information relating to the Data Acquisition System 

(DAS) contained in the PTF. 

volumes of the PTF. However, due to either incomplete or missing information 

or the use of German documentation within the PTF, this information needed to 

be supplemented by discussions with PSI personnel so that a thorough 

understanding of the scope of the DAS could be gained. Based on the 

information contained in the PTF and discussions with PSI personnel, the team 

concluded that the DAS was sufficient for meeting the instrumentation 

requirements of the PANDA test program and that it was controlled in 

accordance with PQAP-DA, "Data Acquisition System Control . I' 

The DAS information is contained in four separate 

3.3.2 1991 PANDA Design Review 

The original PANDA design was developed from the SBWR conceptual design as it 

existed in the late 1980's; the volumetric scale was derived from 

representation of the passive containment cooling system (PCCS) heat exchanger 

(HX). The PCCS HX design was changed about 1991, which necessitated a slight 

-6- 

61 
change in the volumetric scale of the facility (from about 1:18 to 1:25). 

design review was convened in San Jose in October 1991 to assess the facility 

design and to determine technical issues requiring GE or PSI to follow-up. 

Based on a review of documentation contained in the PTF, it appeared that the 

design review was independent and comprehensive. Numerous recommendations 

were made regarding facility design, quality assurance programmatic aspects, 

and technical issues, and specific action items were assigned to the 

participating organizations (primarily, GE and PSI). However, there was no 

record in the PTF of whether the action items and recommendations were ever 

dispositioned. 

review, the NRC team was informed that, since the design review memorandum was 

originated by GE/San Jose, the written record of disposition of the 

recommendations of the review group should be located in the design record 

file (DRF) in San Jose. Therefore, verification of the disposition of the 

design review action items was not possible at PANDA. The GE disposition of 

the recommendations and specific action items identified during the October 

1991 PANDA design review regarding facility design, quality assurance 

programmatic aspects, and technical issues was identified as Unresolved Item 

A 

When GE and PSI were asked about follow-up to the design 

99900403/96-01-04. 

3.4 Test Control 

The purpose of the review of test control was: (1) to determine whether a 

suitable test program was developed to assure that all testing required to 

demonstrate that systems and components would perform satisfactorily in 

service, (2) to determine that such a test program was identified and 

performed in accordance with written test procedures which incorporate or 

reference the requirements and acceptance limits contained in the applicable 

design documents, (3) to assure that test procedures include provisions for 

assuring that all prerequisites for the given test have been met, that 

adequate instrumentation is available and used, and that testing is performed 

under suitable environmental conditions, and (4) to assure that test results 

are documented and evaluated to assure that test requirements have been 

satisfied. 

To assess the level of control over the testing program, the team examined the 

adequacy, implementation, and documentation resulting from the development and 

performance of facility characterization tests, shakedown tests, steady state 

tests, and integral system tests. The PANDA matrix tests (steady state and 

integral system) were performed in accordance with GE Panda Qual i ty Assurance 

Procedure "Test Control," PQAP-TC. The purpose of PQAP-TC is to define the 

process for specifying, performing, eval uat i ng, and documenting the PANDA 

tests. The specific Test Plans and Procedure reviewed by the team along with 

resulting test file documentation are specified below. 

Section 8, "Shakedown and Plant Characterization," of the GE PANDA Test 

Specification (25A5587, Revision 1) required facility shakedown and plant 

characterization tests to be performed. 

in a manner which would expose the facility components and systems to 

conditions similar to those expected during the matrix tests. 

characterization tests were to consist of tests that quantify specific 

The shakedown tests where to be run 

The 

-7- 

62 

I 
characteristics of the facility such as vessel heat loss and line pressure 

drop tests. 

The facility characterization tests were completed in July 1995. 

reviewed Section 9 of the PTF, "Facility Characteristics," to assess whether 

adequate qual i ty assurance measures had been fol 1 owed in the preparation, 

conduct, and documentation of these tests. The facility characterization 

tests were performed in accordance with ALPHA-510 "PANDA Facility 

characterization Heat Loss and Selected System Lines Pressure Loss Test Plan 

and Procedures." The heat loss test is needed for calculation of energy 

balances which in turn would be used to assess system performance and to 

reliably model heat losses from the PANDA test facility in computer code 

analyses. 

The team 

With respect to the heat loss test, Section 9 of the PTF only included the 

test plan and procedure. 

test result reports were not available. 

reports were still under development and provided the team with a draft 

report, ALPHA-519-A, "PANDA Facility Characterization Vessel Heat Loss 

Measurements," dated August 11, 1995, for review. The draft report indicated 

that the heat loss calculations were preliminary and were intended to provide 

a first look at the vessel heat losses. 

calculations had not included all potential heat losses nor the vessel leakage 

rates. 

Results of the test, apparent test result and final 

GE and PSI stated that the test 

The draft report indicated that the 

The system 'line pressure loss test was performed to assure that system line 

pressure drop characteristics measured in the PANDA facility adequately 

simulated the pressure loss characteristics of the full scale SBWR system. 

This test was performed for loss measurements in the isolation condenser and 

primary containment cooling (PCC) system feed line, PCC vent line, gravity 

driven cooling system lines, equalization lines, and main steam lines. 

Section 9 of the PTF included a report, ALPHA-517-0, "PANDA Facility 

Characterization System Line Loss Coefficient Measurements," dated February 

14, 1996, which provided the results and an evaluation of the system line 

pressure loss tests. 

The team inquired as to whether additional facility characterization tests had 

been performed, in addition to the heat loss and system line pressure loss 

test. GE and PSI indicated that a leak test had been performed at the PANDA 

facility in accordance with ALPHA-511 "PANDA Facility Characterization Vessel 

Cold Leak Test Plan and Procedure." This test is important because the 

leakage rate from each PANDA vessel must be known to permit calculation of 

vessel heat losses from the heat loss test data. The leak rate is used to 

separate the components of pressure drop due to condensation and mass lost 

from the system. Furthermore, an estimate of the overall leakage rate is 

necessary to characterize the system for the transient tests. 

from the system must be quantified to properly interpret data from the 

transient tests. 

been developed, reviewed, and controlled through the use of the P-ERM and 

Verification Cover Sheet in accordance with PQAP-DC. 

The mass loss 

From the review of ALPHA-511, the team concluded that it had 

-8- 

63 
Section 9, "Test Matrix," of the GE PANDA Test Specification (25A5587, 

Revision 1) required a series of steady-state tests to be conducted using one 

of the PANDA PCC condensers. The objectives of the steady-state tests was to 

provide additional data to support the adequacy of TRACG to predict the quasisteady heat rejection rate of a PCC heat exchdnger and identify the effects of 

scaling on PCC performance by using one of the PANDA PCC condensers connected 

directly to the steam supply. 

accordance with ALPHA-410, "PANDA Steady-State PCC Performance Tests Test P an 

and Procedures. It 

The team reviewed Section 8 of the PTF "Steady-State Tests,'' including the 

test specification, test plan and procedures, shakedown test results, data 

reduction/reduced data records, apparent test results report, and analytica 

work. In addition, the PTF included: a copy of the control room procedures 

used, excerpts of the PANDA journal, instrumentation 1 ist, DAS channel 

allocatlon table, instrument checks, checklists per the test plan and 

procedures, valve status reports, re-zeroing charts, DAS monitor printout, and 

any non-conformance reports. 

for the test, instrumentation requirements, and test acceptance criteria. 

The steady-state tests were conducted in 

The test procedures specified the prerequisites 

Prior to performing the actual steady-state tests, PSI performed shakedown 

tests to expose the PANDA facility components to conditions similar to those 

expected during the matrix tests. During the first series of shakedown tests 

for the Sl-S6 steady-state tests, steady-state conditions could not be 

achieved as required by ALPHA-410. 

test acceptance criteria through use of a Nonconformance Report in accordance 

with GE PANDA Qual i ty Assurance Procedure PQAP-NC, "Nonconformance Control and 

Corrective Action," Revision 0, dated January 31, 1995. PQAP-NC establishes 

the requirements and procedures for the identification, documentation, 

resolution and control of nonconforming items. With respect to tests Sl-S6, three nonconformances were ident i f ied and documented in accordance with 

PQAP-NC and one nonconformance resulted from the S7-S9 tests. 

PSI documented the failure to meet the 

Section 9, "Test Matrix," of the GE PANDA Test Specification (25A5587, 

Revision 1) required a series of transient integral systems tests to be 

conducted. These tests were to provide an integral systems database for PCC 

system performance with conditions representative of the long-term post-LOCA 

SBWR containment. The objectives of the transient integral systems tests was 

to provide a sufficient database to confirm the capability of TRACG to predict 

SBWR containment system performance, including potential systems interaction 

effects. The transient integral systems tests were conducted in accordance 

with the various test plans and procedures identified in the table in Section 

3.3 above. 

The team reviewed the available documentation within Section 10 of the PTF 

relating to the transient integral systems tests. The PTF contained 

comparable information to that included for the steady-state tests such as 

copies of the control room procedures used, excerpts of the PANDA journal, 

instrumentation 1 ist, DAS channel allocation table, instrument checks, 

check1 ists per the test plan and procedures, valve status reports, re-zeroing charts, DAS monitor printout, trending charts, and any nonconformance reports. 

64 

-9- 
The team concluded that a suitable test program was developed by GE and PSI 

with applicable test plans and procedures in accordance with the Test 

Specification. 

prerequisites were met and that adequate instrumentation was available. 

The test procedures included provisions for assuring that 

3.4.1 Reporting of PANDA Test Results 

Section 11, "Reporting," of the GE PANDA Test Specification (25A5587, Revision 

1) specified preparation of Apparent Test Results (ATR) reports and Final Test 

Reports (FTW). 

preliminary results for each test or each test series that were to be issued 

within approximately one week following performance of the tests. The FTRs 

are considered to be verified reports which contain the data, analysis, and 

results of all tests and transmitted to GE per the schedule specified in the 

Test Plan and Procedures documents. The FTRs are identified as Data 

Transmittal Reports (DTRs) within the "Reports" section of the various test 

plan and Procedures. 

The ATR reports are considered to be unverified reports of 

For the facility characterization tests, Section 4.3, "Post-test/Apparent Test 

Results Report Inputs" of ALPHA-510 qcecifies that following completion of the 

tests, data reduction will be performed to support preparation of the Test 

Results reports (TR). 

all the required measurements covering the full test duration. 

will be reviewed and reported in the TR. Section 4.4, "Post-test/Final Test 

Report," specifies that the Final Test Report (FTR) will transmit all the data 

for the system line pressure loss and the heat loss tests. It will provide 

detailed information on the test instrumentation, test conditions, and the 

format for the data. In addition, samples of key data will be presented in 

plots along with simplified sketches of the test facility configurations 

during testing. 

This data reduction will include time history plots of 

These results 

GE and PSI provided the team with "draft" copies of ALPHA-519-A, "PANDA 

Facility Characterization Vessel Heat Loss Measurements," dated August 11, 

1995, and ALPHA-517, "PANDA Facility Characterization System Line Loss 

Coefficient Measurements," dated February 14, 1996. The team concluded that 

these draft reports do not meet the timeliness requirements of the Test 

Specification or Test Plan and Procedure for providing ATR reports and FTRs 

after completion of the tests. 

For the steady-state tests, Section 10, "Reports," of ALPHA-410, "PANDA 

Steady-State PCC Performance Tests Test P1 an and Test Procedures, I' specifies 

preparation of apparent test results reports within approximately 1 week of 

completion of the steady-state tests. In addition, ALPHA-410 specifies 

preparation of the DTR approximately 2 months after completion of the steadystate tests. 

"PANDA Steady-State Tests S1 through S6 PCC Performance Apparent Test 

Results," however, a DTR or FTR had not been issued. The team concluded that 

a DTR or FTR had not been prepared in accordance with the Test Specification 

or the Test Plan and Procedure. 

GE and PSI provided the team with various versions of ALPHA-509, 

For the transient integral systems tests, Section 10, "Reports," of GE 

Document 25A5764, "PANDA Test Plan - Tests M3, M3A, M3B, M4, M7," Revisions 1, 

-10- 

65 
2, and 3 dated September 18, 1995, October 16, 1995, and November 15, 1995, 

respectively, requires (1) preparation of an ATR report within approximately 

two weeks of completion of each transient integral system test, and (2) 

preparation of a DTR approximately two months after the last test is 

performed. GE and PSI had not prepared any ATR reports or DTRs for the M3, 

M3A, M3B, M4, or M7 transient integral system tests. 

Section 10, "Reports," of GE Document 25A5785, "PANDA Test Plan - Tests M2, 

MIOA, MlOB," dated November 21, 1995, requires (1) preparation of an ATR 

report within approximately two weeks of completion of each transient integral 

system test, and (2) preparation of a DTR approximately two months after the 

last test is performed. GE and PSI had not prepared any ATR reports or DTRs 

for the M2, MlOA, or MlOB transient integral system tests. 

Section 10, "Reports," of GE Document 25A5788, "PANDA Test Plan - Tests M6/8," 

dated December 7, 1995, requires (1) preparation of an ATR report within 

approximately two weeks of completion of each transient integral system test, 

and (2) preparation of a DTR approximately two months after the last test is 

performed. GE and PSI had not prepared an ATR report or DTR for the M6/8 

transient integral system test. 

Section 10, "Reports," of GE Document 25A5824, "PANDA Test Plan - Test M9," 

dated December 12, 1995, requires (1) preparation of an ATR report within 

approximately two weeks of completion of each transient integral system test, 

and (2) preparation of a DTR approximately two months after the last test is 

performed. 

transient integral system test. 

GE and PSI had not prepared an ATR report or DTR for the M9 

The failure to prepare ATR reports, FTRs and/or DTRs, on a time schedule 

consist with the applicable Test Specification and Test Plan and Procedures 

requirements is identified as Nonconformance 99900403/96-01-01. 

3.4.2 PANDA lest Analyses 

The team also examined the analytical efforts that support the PANDA testing 

program. Section 10, "Pretest Predictions/Acceptance Criteria, I' of GE PANDA 

Test Specification 25A5587, Revision 1, specifies that pretest calculations be 

performed for some of the matrix tests planned for SBWR certification. This 

activity was to include development of a TRACG input model for the PANDA 

facility, verification of the input model against as-built test facility data, 

design review of the input model, calibration of the input model using heat loss and pressure drop data from test facility characterization testing, 

selection of the test conditions for simulation, performance of the 

calculations, and documentation of the results. 

GE provided the NRC with the SBWR-Pretest Report for PANDA Test M9 in a letter 

dated December 12, 1995. 

for the TRACG code for application to the SBWR program. 

include both pre and post-test calculations for tests in the PANDA test 

program. 

This report was to support the validation efforts 

These calculations 

The analyses of the tests were being performed by an SBWR PANDA 

-11- 

66 
analysis team, with participation from PSI, KEMA and ECN in the Netherlands, 

Institute de Investigaciones (IIE) of Mexico, and GE. 

GE indicated that the TRACG modeling of the PANDA test facility was developed 

by PSI. GE then had an individual with IIE of Mexico verify the TRACG 

modeling and nodalization of the PANDA facility. 

to the level of quality assurance oversight that GE had performed of IIE or 

the individual performing the work. GE presented conflicting stories with 

respect to whether the agreement for review of the TRACG modeling was with IIE 

as a subcontractor or with an individual of IIE as a subcontractor. The 

extent and level of effectiveness of quality assurance oversight by GE over 

its SBWR program international technical associates was identified as 

Unreso 1 ved It em 99900403/96-01-05. 

The team inquired to GE as 

3.5 As-Built Drawings and Confiquration Control 

PQAP-V, "Verification," Revision 1, dated May 1, 1995, implements the 

applicable requirements of GE EOP 42-6.00, "Independent Design Verification," 

and EOP 40-7.00, "Design Reviews," for verification of the PANDA test facility 

configuration and testing activities. 

as-built drawings are stored in Section 2.1, "Faciiity Drawings," and in 

Section 3.1, "As-Built Drawings," of the PTF. 

Records of PANDA'S facility and 

In October 1993, PSI contracted with Elektrowatt (see Section 3.2) to generate 

as-built drawings for PANDA. 

depicted in a Giovanola (the facility builder) design drawing No. 164-A3526- 

IC (PSI Drawing No. 1-290111~). 

develop an as-built of the main configuration and was subsequently given the 

designation of PSI Drawing No. 1-290300. A1 1 subsequent as-buil t measurements 

taken by PSI, including instrument and valve locations, were based on the 

Elektrowatt measurements. The Passive Containment Cooling System (PCC) and 

Isolation Containment System (IC) units (manufactured by Jaggi , AG) were 

measured by PSI personnel after their arrival on-site to establish their 

as-built dimensions. 

The facility configuration was originally 

This drawing was used by Elektrowatt to 

As-built tolerances for the PANDA facility were established by GE in Document 

No. 25A5764, "PANDA Test Plan - Tests M3, M4, M7," Revision 1, dated September 

18, 1995. The team found evidence that PSI had performed a review to verify 

that all as-built dimensions identified in drawings generated by Elektrowatt 

met the tolerance criteria specified in GE Document No. 25A5764. 

Except for the unresolved item identified in Section 3.2, above, and based on 

the reviews of pertinent documents in Sections 2.1 and 3.1 of the PANDA PTF, 

the team concluded that activities performed by PSI after January 1995 were 

consistent with the provisions in PQAP-V. 

3.6 Procurement Control 

PQAP-PC, "Procurement Control," Revision 0, dated January 31, 1995, defines 

the requirements for procurement initiated by PSI in support of the PANDA test 

program, after test facility commissioning, for equipment and services. This 

-12- 

67 
procedure implements the applicable requirements of GE Engineering Operating 

Procedures, EOP 45-1.00, "Procurement Initiation and Control ,'I EOP 45-2.00, 

"Procurement of Engineering Services," and in part, EOP 35-3.20, "Calibration 

Control . 'I 

In this area, the team was primarily interested in examining the 

implementation of PSI procurement provisions with respect to calibration 

services. 

evidence that after January 1995, PSI had adequately implemented the 

applicable provisions in PQAP-PC. 

As discussed below in Section 3.7, the team found objective 

3.7 Control of Measurinq and Test EauiDment 

PQAP-CC, "Control of Measuring and Test Equipment," Revision 0, dated January 

31, 1995 defines and establishes all requirements related to the processes and 

procedures used for calibration of PANDA instrumentation. Section 4, 

"Instrumentation," of the PTF contained all documentation related to the 

procurement and calibration of PANDA instrumentation, including calibration 

certificates furnished by companies accredited by the Swiss Federal Office of 

Metrology (Eidgenossisches Amt fur Messwesen) in Bern. 

In Section 5.4, "Instrument Calibration," of PSI Document No. ALPHA-410, 

"PANDA Steady-State Tests - PCC Performance Test Plan and Procedures," 

Revision 2, dated May 16, 1995, PSI describes in detail its approach for 

ensuring that calibration of the various PANDA instruments was adequately 

performed and documented. Except for pressure and differential pressure 

sensors, all instruments were individually, or on a sampling basis, sent to 

the Swiss Federal Office of Metrology in Bern for calibration. 

All pressure and differential pressure sensors used in PANDA were manufactured 

by Rosemount, Inc. Except for the Model 2088 and SMART, all Rosemount 

pressure sensors were calibrated by PSI prior to installation in the facility 

using a reference or standard traceable to the Swiss Federal Office of 

Metrology and in accordance with the requirements in PSI Document No. ALPHA408, "PANDA Instrumentation and Control - PANDA Pressure Transmitter 

Calibration," Revision 1. For the Model 2088 and SMART sensors, the Rosemount 

factory calibration data was used. 

The team inquired as to why PSI was relying solely on the manufacturer's 

calibration data for the Model 2088 and SMART sensors. PSI stated that these 

instruments were software-controlled and PSI lacked the necessary hardware and/or software to test them properly. 

Switzerland, where these instruments would be re-cal ibrated after completion 

of testing, is a metrology laboratory accredited by the Swiss Federal Office 

of Metro1 ogy. 

Based on the above information, the team concluded that PSI had adequately 

implemented the provisions in PQAP-CC and PQAP-PC and this area was identified 

as a strength in the PANDA QA program. 

PSI also stated that Rosemount of 

-13- 

68 
3.8 Nonconformance Control and Corrective Action 

PQAP-NC, "Nonconformance Control and Corrective Action, I' Revision 0, dated 

January 31, 1995, establishes the requirements and describe the procedure for 

the identification, documentation, resol uti on and control of nonconforming 

items for the PANDA program. 

related activities that can affect PANDA test results and it implements the 

applicable requirements of GE EOP 75-4.10, "Control of Nonconforming 

Material," and EOP 75-3.00, "Corrective Action and hdits." 

This procedure applies to all PANDA quality 

Based on reviews of nonconformance reports found in the PTF, and based on 

conversations with PSI test personnel, the team learned that although abnormal 

occurrences, subsequently causing matrix testing to be suspended and 

re-evaluated, had been detected during testing, no nonconformance reports had 

been generated by PSI to document these events. The team also learned of the 

existence of a new PSI procedure ("Data Base Modification," issued in early 

March 1996) used extensively by PSI testing personnel to evaluate, and when 

necessary, modify, i .e. , revise or delete, actual test results data. A1 though 

this "Data Base Modification'' procedure was clearly being used to perform an 

activity affecting quality as well as an activity that introduced deviations 

from the test control process specified by PQAP-TC, and from the 

nonconformance identification process identified in PQAP-NC, the "Data Base 

Modification" procedure had not been identified or described as a Quality 

Assurance Procedure governed by PPCP-QA-01. This i ssue was identified as 

Nonconformance 99900403/96-01-02. 

3.9 Personnel Trainincl and Qualification 

PQAP-PT, "Personnel Training and Qualification," Revision 0, dated January 31, 

1995, establishes the personnel training and qualification requirements to be 

implemented on the PANDA Project for test facility personnel. 

implements the applicable requirements of the appropriate GE EOPs and states 

that individuals who perform activities affecting the quality of the PANDA 

project must be proficient in the appropriate technical discipline and the 

procedural systems. 

PQAP-PT 

Technical qualifications specify a minimum education, experience, and/or 

special technical training requirements. Procedurally, each individual shall 

be indoctrinated or instructed in the applicable quality assurance procedures. 

Indoctrination and training shall be attained and maintained by methods such 

as procedure reading, class training and/or on the job training. 

The team verified through review of PTF training records that a71 PANDA 

personnel had received training in all sections of the PQAP and test 

procedures. 

Personnel qualifications were done in accordance to level of education and 

years of experience in the desired fields. All personnel running and 

supervising the tests were appropriately trained and qualified in accordance 

with PQAP-PT. 

Manager both qualified to the highest qualification level required. 

The PSI ALPHA Project Manager and the and the PSI PANDA Project 

-14- 

69 
There were no records of subcontractor training and qualification in the PTF. 

However, GE stated that such documentation existed in the PANDA DRF maintained 

in San Jose for the training of certain international technical associates 

performing PANDA data analysis. 

Based on the above review of personnel training and qualification records in 

the PANDA PTF, the team concluded that GE/PSI had adequately implemented the 

provisions and requirements of PQAP-PT for PSI personnel. 

3.10 Quality Assurance Records 

PQAP-R, "Quality Assurance Records," Revision 0, dated January 31, 1995, 

defines the requirements for identification, accumulation, review, 

maintenance, and retention of the quality assurance records in the PTF. 

PQAP-R implements the applicable requirements of GE EOP 42-10.00, "Design 

Record File," and EOP 75-6.00, "Qual ity Assurance Records. 

PQAP-R requires that a central file of legible, accurate and complete QA 

records, the PANDA Test File, shall be established with an index and table of 

contents. 

duration of the testing and at completion of testing, the PTF will be 

transferred to GE for inclusion in the PANDA Design Record File. 

PQAP-R also requires that the PTF be stored in an archive for the 

The NRC team was informed of two pertinent facts about the test files at PSI: 

first, since the program is still active, the PSI files have not been closed 

and, in fact, detailed information on most of the tests had not yet been 

included in the PTF; and, second, the PSI files do not, and are not intended 

to, comprise the complete DRF for the PANDA program. Important supporting 

information is contained in the DRF at GE's offices in San Jose. When the PSI 

test files have been completed and closed, the PTF will be provided to GE, and 

the combined set of files will comprise the complete PANDA DRF. 

Test result records were identifiable and retrievable to the extent they were 

included in the PTF. 

included in the PTF until it has completed the PSI Project Manager's review 

process. 

data that was not yet stored in the PTF. 

However, the team was told that test data is not 

This process resulted in the team having to request completed test 

Overall, the documentation in the PTF reflected evidence of appropriate 

implementation of the PPCP and the QA procedures. 

these reviews, the team concluded that the QA records control process was 

adequately establ ished and implemented for the PANDA test program. 

Based on the results of 

3.11 Audits 

Sections 4.0, "Project Assessment," and 4.1, "Audits," of PPCP-QA-01, define 

the project implementation requirements for internal audit activities at PANDA 

though GE establ i shed procedures P&P 70-1 1, GE Qual i ty System Requirements, 

and Administrative Guide AG-017. A11 internal audits and oversight of the 

PANDA test program were conducted by GE certified auditor(s). 

confirmed that audits were performed by appropriately trained QA personnel 

with GE-certified lead auditors. 

The team 

Audits included the following: 

-15- 

70 
Audit plan and schedule 

Audit check list 

Preaudit orientation and training 

Assessment 

Audit report issued to management 

Corrective action requests (CARs) 

Resolution of audit findings 

Follow-up to closure of CARs 

( if required) 

The team reviewed results of a GE audit of the PANDA Test Facility conducted 

January 31 through February 2, 1995. The team also reviewed the results of 

the readiness assessment conducted in October 1994, including the open items 

and recommendations that were documented in the Readiness Assessment Report. 

During the January 1995 GE audit, it was determined that 11 of the 15 open 

items and 5 of the 9 recommendations identified during the readiness review 

were still unresolved. GE performed a second readiness review in September 

1995 and the results were documented in a report identifying several findings. 

Corrective act ion requests (CARs) were issued and the appropriate foll ow-up to 

assure closure of the CARs was documented. 

The team determined that no external supplier audits were performed by either 

PSI or GE for PANDA suppliers. 

team concluded that GE was imp1 ement i ng an appropri ate internal audit program 

at the PANDA test facility. 

Based on the results of these reviews, the 

4 PERSONNEL CONTACTED 

GE Nuclear Enercty 

a James E. Quinn, Projects Manager, LMR and SBWR Programs a John Torbeck, Project Manager, SBWR Test Operations a 

a G. Wingate 

Norman Barcl ay, Manager, Audit Programs 

a J. Fitch 

Paul Scherrer Insti tut 

a Jorg Dreir a George Varadi a Max Huggenberger a Otto Fischer a George Yadigaroglu a Paul Coddington a J. Healzer 

-16- 

71 
0 Richard P. McIntyre, Team Leader, Special Inspection Branch a Juan 0. Peralta, Quality Assurance and Maintenance Branch 

0 John A. Kudrick, Containment Systems and Severe Accident Branch (SCSB) 

e John Monni nger, SCSB 0 Alan E. Levin, Reactor Systems Branch 0 Dino C. Scaletti, Standardization Project Directorate 

-17- 

72 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 205659801 

March 5, 1996 

Mr. Craig P. Kipp 

Plant Manager General Electric Nuclear Energy 

Nuclear Energy Production 

P.O. Box 780, Mail Code A20 

Wilmingtan, NC 28402-0780 

SUBJECT: NONPROPRIETARY VERSION OF NRC INSPECTION REPORT NO. 99900003/95-01 

Dear Mr. Kipp: 

This letter transmits the nonproprietary version of the U.S. Nuclear 

Regulatory Comnission's (NRC's) Inspection Report 999000003/95-01. 

to you dated January 4, 1996, transmitted the original (proprietary) version 

of the report. On the basis of cur A4scussions on February 12, 1996, and 

review of the information in your letters of January 24, 1996, (RJR-96-013), 

and its enclosures (Affidavit, dated January 23, 1996, and Proprietary 

Information Summary Sheet) and February 27, 1996, and its enclosures 

(Affidavit, dated February 27, 1996, and Proprietary Information Summary 

Sheet), we have concluded that certain specific information identified in your 

letters could be regarded as proprietary and, as such, were removed from the 

inspection report. 

report, the NRC has briefly summarized the deleted text. 

Our letter 

In the revised nonproprietary (pub1 ic) version of the 

Your response to either this letter or our letter dated January 4, 1996, and 

their enclosure 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 (a) of the NRC "Rules of Practice," a copy of 

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

If you have any questions concerning this matter, please contact Steven M. 

Matthews at (301) 415-3191. 

Sincerely , 

1 Djvision of' Inspection 

and Support Programs 

Office of Nuclear Reactor Regulation 

Docket No.: 99900003 

Enclosure: Report No. 99900003/95-01 

73 
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-0001 

Mr. Craig P. Kipp 

P1 ant Manager 

General Electric Nuclear Energy 

Nucl ear Energy Production 

P.O. Box 780, Mail Code AZO 

Hi 1 mi ngton, NC 28402-0780 

SUBJECT: NRC INSPECTION REPORT NO. 99900003/95-01 

Dear Mr. Kipp: 

From August 14 through September 1, 1995, the U.S. Nuclear Regulatory 

Commission (NRC) conducted an inspection of the General Electric Nuclear 

Energy (GENE) activities at the Nuclear Eneryy Production (NEP) facilities in 

Wilmington, North Carolina. This letter trclnsmits the report of that 

inspection. 

Steven M. Matthews of NRC's Special Inspection Branch led the inspection team, 

which included the other inspectors named in the report. The team conducted a 

performance-based evaluation of the NEP management, staff, and quality 

programs. They also assessed NEP's implementation of those programs related 

to boiling-water reactor reload core design, safety analysis and 1 icensing 

processes, fuel assembl ies, fuel-re1 ated core components, and fuel -re1 ated 

inspection services supplied to the U.S. nuclear industry. In the course of 

this evaluation, the team examined technical documentation, procedures, and 

representative records. They a1 so interviewed and held discussions with NEP 

personnel. In addition, the team listened to presentations by NEP personnel 

.and observed work activities in progress. 

On the basis of this inspection, the team determined that, for the most part, 

the "GE Nuclear Energy, Quality Assurance Program Description," NEDO-11209- 

04A, Revision 8, Class 1 (approved by the NRC on March 31, 1989), was 

implemented in an appropriate manner for the areas evaluated. Therefore, the 

team did not cite any violations or nonconiormances. The enclosed inspection 

report identifies the areas examined and presents a detailed discussion of the 

team's conclusions. 

The report describes the team's conclusion that NEP's nonsafety-re1 ated 

classification of the constituent fuel bundle parts was not based on a 

functional evaluation of the parts. 

fuel bundle operates in a unique BWR environment, a basis for its nonsafetyre1 ated cl assi f i cat i on of parts could be devel oped. Therefore, NEP i s 

requested to submit its functional evaluation and technical basis for the 

nonsafety-re1 ated parts classification. 

NEP asserted that , because the channeled 

During this inspection, the team identified areas and activities within NEP 

that it considered strengths of your organization, as described in the 

inspection report. 

activities that affect quality. The most significant of these concerns the 

However, the team also observed weaknesses in certain 

74 
C. Kipp -2- 

accuracy of design methods used to ensure that the reactor can safely be 

operated. 

reactivity for the recent reloads with the newer fuel designs and long cycle 

lengths was a weakness of the NEP reload design process. 

The team concluded that the erru, A in czlculated hot and cold core 

The team noted, however, that NEP was addressing this weakness by implementing 

improvements in its steady-state nuclear methods. These improvements include 

a formal review process, increasing benchmarking of certain calculations for 

an extensive number of bundle design conditions, implementing revised computer 

codes, and developing a new lattice code. However, because the issue of 

accuracy in eigewalue calculation is related to the issue of the accuracy of 

design methods used to provide assurance that the reactor can safely be 

operated, it is appropriate that the proposed design improvements be 

thoroughly documented, peer reviewed, and monitored over a period of time to 

ensure that the new design method. 3-e indeed meeting the requirements that 

have been placed on them. 

The team concluded that the near-term use of the revised computer codes in 

combination with the eigenvalue selection process should help reduce 

uncertainties in the cold critical and shutdown margins. The team also 

concluded that the assurance of adequate shutdown margin can be strengthened 

by joint GENEllicensee actions consistent with the plant startup safety 

analys i s . 

During this inspection, the team also reviewed the new GE12 and GE13 fuel 

designs as well as followed up on issues resulting from an earlier audit of 

the GEll fuel design. The team raised concerns regarding the upper limit 

subcooling range and the altered definition for the bundle R-factor 

distribution. GENE responded in timely manner to these concerns with 

submittals described in the inspection report. GENE’S submittals, in response 

to the team’s concerns, are under evaluztion by the NRC and will be addressed 

separately. 

The team’s review of the new fuel designs described in Section 3.5 of the 

enclosed report identified instances where fuel design commitments to the NRC 

were not fully met, as prescribed in Amendment 22 of the ”General Electric 

Standard Application for Reload (GESTAR) 11” topical report documented in 

NEO€-24011-P-A, “General Electric Standard Application For Reactor Fuel” 

(approved by the NRC on July 23, 1990). 

Neither the weaknesses nor the observations summarized in Appendix A of the 

enclosed inspection report require a written response from NEP. The report 

also includes several open items that could not be resolved during the 

inspection, or for which the team needs additional information to reach its 

conclusions. 

items, as listed in Appendix B of the enclosed report. 

NEP is requested to submit a written response to these open 

75 
C. Kipp -3- 

In accordance with Section 2.790(a) of Tit? 10 of the Code of Federal 

Requlations, a copy of this letter and its enclosure will be placed in the NRC 

Public Document Room and made available to the public unless you notify this 

office by telephone within 10 days of the aate of this letter and submit a 

written application to withhold the information contained therein. Such 

application must be consistent with the requirements of 10 CFR 2.790(b)(l). 

Your response to this letter and its enclosure is 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 inspection, we will be pleased 

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

Robert M. Gal 14 Chief 

Special Inspection Branch 

Division of Inspection 

and Support Programs 

Office of Nuclear Reactor Regulation 

Docket No.: 99900003 

Enclosure: Report No. 99900003/95-01 

76 
REPORT NO. : 

ORGANIZATION: 

ORGANIZATIONAL 

CONTACT: 

NUCLEAR INDUSTRY 

ACTIVITY: 

INSPECTION DATES: 

LEAD INSPECTOR: 

OTHER INSPECTORS: 

REVIEWED BY: 

APPROVED BY: 

REVISED NONPROPRIETARY VERSION 

8, S. NUCLEAR REGULATORY CO~ISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF INSPECTION AND SUPf9RT PROGRAHS 

99900003/95-0 1 

General Electric Nuclear Energy 

Nuclear Energy Prduction 

Wilmington, North Carolina 

James F. K1 approth , Manager 

Fuel and Fac i 1 i ty Licensing 

Environmental Health & Safety and 

Nuclear Quality Assurance 

General Electric Nuclear Energy (GENE) , Nuclear 

Energy Production (NEP) provides boi 1 i ng-water reactor 

reload core designs, safety analysis, and 1 icensing, 

fuel assemblies, fuel-related core components, and 

fuel-related inspection services to the U.S. nuclear 

industry. 

August 14 through September 1, 1995 

Steven M. Matthews 

Vendor Inspection Section (VIS) 

Special Inspection Branch (PSIB) 

- Date 

David H. Brewer, PSIB/VIS 

Daniel W. Brite, Battelle Pacific Northwest Laboratory 

Carl E. Beyer, Battelle Pacific Northwest Laboratory 

Dr. John F. Carew, Brookhaven National Laboratory 

Carl 3. Czajkowski , Brookhaven National Laboratory 

Rodney L. Grow, Parameter, Inc. 

Edward D. Kendrick, DSSA/SRXB 

Patrick S. Lacy, Parameter, Inc. 

Dr. Partha Neogy, Brookhaven National Laboratory 

Larry E. Phillips, Chief, DSSA/SRXB 

Alexander C. Schafer, Parameter, Inc. 

John W. Thompson, AEOD/DEIIB 

3/sjpc 

Date 

/ 

obert u@ lo, Chief, NRR/DISP 

77 

3/%66 

Date 

ENCLOSURE 
CONTENTS ..... 

Paae 

ABBREVIATIONS AND ACRONYMS ....................... vi 

1 SCOPE AND SUMMARY OF INSPECTION FINDINGS ............. 1 

1.1 Violations 1 

1.2 Nonconformances 1 

1.3 Weaknesses. Strengths. and Observations ........... 2 

1.4 Open Items ......................... 2 

......................... ....................... 

2 STATUS OF PREVIOUS INSPECTION FINDINGS .............. 2 

3 ZNSPECTION FINDINGS AND OTHER COMMENTS .............. 2 

3.1 Background ......................... 2 

3.2 Entrance Meetings and Final Exit Meeting .......... 2 

3.3 Nuclear Quality Assurance .................. 3 

3.4 Reload Core Design. Safety Analysis. and Licensing 

Processes ......................... 4 

3.4.1 Licensee Requirements ................ 4 

3.4.2 Reload Core Design Procpss .............. 5 

3.4.3 Reload Safety Analysis Process ............ 14 

3.4.4 Reload Licensing Process ............... 19 

3.4.5 Fuel Assembly Mechanical Design ........... 27 

3.4.6 

Programs ...................... 30 

BWR Engineering Database and Engineering Computer 

3.5 Fuel Designs ........................ 34 

3.5.1 Lead Use Assemblies. 2.1(2) ............. 35 

3.5.2 Thermal-Mechanical. 2.2 ............... 35 

3.5.3 Safety Limit MCPR. 2.6 ............... 39 

3.5.4 Stability Licensing Acceptance Criteria. 2.9 .... 40 

3.5.5 Refueling Accident. 2.13 .............. 41 

3.6 Nuclear Fuel Fabrication .................. 42 

3.6.1 

3.6.2 

3.6.3 

3.6.4 

3.6.5 

3.6.6 

3.6.7 

3.6.8 

3.6.9 

Procurement ..................... 42 

Safety Classification ................ 43 

Chemical and Ceramic Operations ........... 46 

Fuel Clad Tubing ................... 51 

Fuel Bundle Components ................ 52 

Fuel Rod Fabrication ................. 57 

Fuel Bundle Assembly ................. 59 

Fuel Certification .................. 59 

Laboratories ..................... 60 

. ii .. 

78 
3.7 Corrective Actions ..................... 66 

3.8 Training .......................... 68 

3.8.1 Reload Core Engineering ............... 68 

3.8.2 Fuel Production ................... 70 

3.9 10 CFR Part 21 ........................ 72 

APPENDIX A WEAKNESSES. STRENGTHS. AND OBSERVATIONS .......... A-1 

APPENDIX B OPEN ITEMS ........................ B- 1 

APPENDIX C PERSONS CONTACTED ..................... c- 1 
ABBREVIATIONS AND ACRONYMS 

ABM 

AOU 

A00 

ANS I 

APRM 

ASME 

ASTM 

ASNT 

ATLAS 

AVL 

BOC 

BWR 

BWREDB 

CAR 

cc I 

c EC 

CFR 

CILC 

CMOB 

CM&S 

COLR 

CP&L 

C PR 

CSR 

DB 

DBE 

A 

DIP 

DISP 

DNBR 

DRF 

DSSA 

DWS 

ECCS 

ECP 

EOC 

EOP 

ET 

€UP 

FCO 

FFWTR 

FIV 

FMO 

FRED 

FSAR 

GDC 

GENE 

GESTAR 

GESTR 

Automated Bundle Assembly Machine 

Ammon i um D i ur an at e 

Anticipated Operat i onal Occurrence 

American National Standards Institute 

Average Power Range Monitor 

The American Society of Mechanical Engineers 

American Society for Testing and Materials 

American Society for Nondestructive Testing 

GENE’S 8.6-megawatt heat-transfer loop 

Approved Vendors List 

Beginning Of Cycle 

Boiling-Water Reactor 

Boi 1 ing-water Reactor Engineering Data Bank 

Corrective Action Request 

Control Component Individual 

Commonwealth Edison Company 

Code of Federal Regulation 

Crud-Induced-Localized Corrosion 

Core Monitoring Data Bank 

Chemical, Metal 1 urgical and Spectral 

Core Operating Limits Report 

Carolina Power & light Company 

Critical Power Ratio 

Contractile Strain Ratio 

Design Bases 

Design-Basis Event 

Delta (Differential) 

Design Interface Procedure 

Division of Inspection and Support Programs (NRR) 

Departure from Nucleate Boi 1 i ng Ratio 

Design Record File 

Division of Systems Safety and Analysis (NRR) 

Daily Weld Sample 

Emergency Core Cooling System 

Engineering Computer Program 

End Of Cycle 

Engineering Operating Procedure 

Eddy Current Examination 

Energy Uti 1 i zat i on P1 an 

Fuel Components Operations 

Final Feedwater Temperature Reduction 

Flow-Induced Vibration 

Fuel Manufacturing Operations 

Fuel Release Engineering Data 

Final Safety Analysis Report 

General Design Criterion 

General Electric Nuclear Energy 

GENE Standard Application for Reload 

GENE Stress and Thermal Analysis of Fuel Rods 

- iv - 

80 
GRF 

GWd/MTU 

HFP 

ICF 

I PHT 

IMRS 

LFWH 

LHGR 

LMCS 

LOCA 

LUA 

MAPS 

MAPLHGR 

MCNP 

MCPR 

MFB 

MICS 

MSG 

M&TE 

MSIV 

NEP 

NDE 

NOR 

NFUF 

NIST 

NPI 

NQA 

NRC 

NRR 

00 

ID 

OLMC PR 

OPL-3 

OPS 

o/u 

PCMI 

PLR 

PMQC 

PO 

P&P 

PRC 

PSC 

PSIB 

PWR 

QA 

QATS 

QC 

QCI I 

RBM 

RIP 

ABBREVIATIONS and ACRONYMS Continued 

Grid-To-Rod Fretting 

Gigawatt-Days Per Metric Tonne of Initial Uranium Metal 

Hot Full -Power 

Increased Core Flow 

In-Process Heat Treatment 

In-Process Material Release System 

Loss-of -Feedwate r Heat i ng 

Linear Heat Generation Rate 

Laboratory Material Control System 

Loss-of-Coolant Accident 

Lead Use Assembly 

Magnetic and Passive Scanners 

Maximum Average P1 anar Linear Heat Generat ion Rate 

Monte-Carlo Neutron Particle computer code 

Minimum Critical Power Ratio 

Mi sl ocated Fuel Bund1 e 

Material Inventory Computer System 

Mi 11 -S1 ug-Granul it 

Measuring and Test Equipment 

Main Steam Isolation Valve 

Nuclear Energy Product ion 

Nondestructive Examination 

Nucl ear Design Report 

Nuclear Fuel Users Forum 

National Institirte of Standards and Technology 

New Product Introduction 

Nuclear Qual i ty Assurance 

U.S. Nuclear Regulatory Commission 

Office of Nuclear Reactor Regulation (NRC) Outside Diameter 

One-Dimensional 

Operating Limit Minimum Critical Power Ratio 

Operating Plant Licensing Parameters 

Operation Parameter Sheet 

OxygenlUrani um Ratio 

Pellet/Cl adding Mechanical Interaction 

Part-Length Rod 

Purchased Materi a1 Qual i ty Control 

Purchase Order 

Policy and Procedure 

Potent i a1 1 y Reportabl e Condition 

Potential Safety Concern 

Special Inspection Branch (NRR/DISP) 

Pres sur i zed- Wat er React or 

Quality Assurance 

Quality At The Source 

Qual i ty Control 

Quality Control Inspector Instructions 

Rod Block Monitor 

Rod Internal Pressure 

81 
RLQR 

RMS 

RUE 

sco 

SER 

SIL 

SLMCPR 

SRLR 

SUP 

SRXB 

TDP 

T/H 

30 

TIP 

TPE 

TPM 

TREX 

UT 

WA 

ABBREVIATIONS and ACRONYHS Continued 

Reload Licensing Quality Review 

Root Mean Squared 

Rod Withdrawal Error 

Service Components Operations 

Safety Eva1 uat i on Report 

Service Information Letter 

Safety Limit Minimum Critical Pwer Ratio 

Supplemental Reload Licensing Report 

Standard Review P1 an 

Reactor Systems Branch (NRR/DSSA) 

Technical Design Procedure 

Thermal -Hydraul i c (s) 

Three-Dimensional 

Traversing In-Core Probe 

Technical Program Engineer 

Technical Project Manager 

Tube Reduced Extrusion 

U1 trasoni c Testing 

Work Authorization 

- vi - 

82 
1 SCOPE AND SUHUARY OF INSPECTION FINDINGS: 

From August 14 through September 1, 1995, the U.S. Nuclear Regulatory 

Commission (NRC) conducted an inspection of the General Electric Nuclear 

Energy (GENE), activities at the Nuclear Energy Production (NEP) facilities in 

Wilmington, North Carolina. During this fiiapection, the NRC inspection team 

conducted a performance-based evaluation of the NEP management , staff , and 

quality programs. They a1 so assessed the NEP implementation of those programs 

related to boiling-water reactor (BWR) reload core design, reload safety 

analysis and 1 icensing processes, fuel assemblies, fuel-related core 

components , and fuel -re1 ated inspect ion services suppl i ed to the U. S. nuclear 

industry . 

This inspection was conducted to establish a basis for confidence that NEP 

products and services supplied to the U.S. nuclear industry would perform 

their safety function. The following guide1 ines, standards, and regulations 

constitute the basis for this inspection: 

General Design Criterion (GDC) 10. "Reactor Design," and GDC 12, 

"Suppression of Reactor Power Oscil lation3," of Appendix A, "General Design 

Criteria for Nuclear Power Plants," to Part 50, "Licensing of Production and 

Utilization Facilities," of Title 10 of the Code of Federal Requlations 

(10 CFR Part 50). 

Appendix By "Quality Assurance Criteria for Nuclear Power Plants and 

Fuel Reprocessing Plants," to 10 CFR Part 50. 

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

Defect. It 

Section 4.2, "Fuel System Design," of NRC NUREG-0800, "Standard 

Review Plan," Revision 2, dated July 1981, and its Appendix A, "Evaluation of 

Fuel Assembly Structural Response to Externally Appl ied Forces, 'I Revision 0. 

"GE Nuclear Energy, Qual i ty Assurance Program Description, 'I NEDO11209-04AY Revision 8, Class 1 (approved by the NRC on March 31, 1989, as 

meeting the requirements of Appendix B to 10 CFR Part 50), hereafter referred 

to as the 'QA topical report." 

Amendment 22 of the "General Electric Standard Application for Reload 

(GESTAR) 11" topical report documented in NEDE-24011-P-AY "General Electric 

Standard Application for Reactor Fuel" (approved by the NRC on July 23, 1990). 

1.1 Viol otions 

No violations were identified during this inspection. 

1.r Nonconformances 

No nonconformances were identified during this inspection. 

-1- 

83 
1.3 Weaknesses, Strenclths . and Observations 

During this inspection, the team noted weaknesses and observations concerning 

NEP activities that affect quality. 

strengths, and observations are summarized in Appendix A to this report. 

Neither the weaknesses nor the observatioYr described in this report require 

any specific action by or written response from NEP. 

The more significant weaknesses, 

1.4 ODen Items 

A written response is requested for the open items described in this report 

and summarized in Appendix B. 

2 STATUS OF PREVIOUS INSPECTION FINDINGS 

No previous fuel inspection findings required follow-up during this 

inspection. 

3 INSPECTION FINDINGS AND OTHER COHHENTS 

3.1 Backoround 

GENE serves the majority of its fuel and component customers from the NEP 

manufacturing facility in Wilmington, North Carolina, where production of BWR 

nuclear fuel and reactor equipment products began in 1968. 

has been recognized for its manufacturing technology, advanced automation, and 

integrated information systems. GENE recently moved its core engineering 

staff to the NEP facility from its facility in San Jose, California. 

The NEP facility 

GENE has amassed an extensive fuel performance experience base. According to 

GENE, as of December 31, 1993, over 4.8 million GENE 8x8 fuel-type production 

Zircaloy-clad UO fuel rods were in, or had completed, operation in commercial 

BWRs. Of the toial, nearly 1.5 million GENE fuel rods were in operation, and 

GENE had loaded more than 2 million pellet/cladding mechanical interaction 

(PCMI) resistant barrier fuel rods in commercial BWRs. 

3.2 Entrance Heetincls and Final Exit Heeting 

The entrance meetings took place on August 14, 1995, for the reload core 

design, safety analysis, and licensing process portions of the inspection, and 

on Aug,ust 21, 1995, for the fuel design and fuel production portions of the 

inspection. 

management and staff, and discussed the scope of the inspection. 

also reviewed its responsibilities for handling proprietary information, as 

well as those of NEP. 

the management and staff of the applicable NEP organizations. 

During these meetings, the team met with members of NEP 

The team 

In addition the team established contact persons within 

During the inspection, the team conducted a performance-based evaluation of 

NEP through technically directed observations aM assessments of processes, 

activities, and documentation. The team examined technical documentation, 

procedures, and representat i ve records. They a1 so interviewed and he1 d 

discussions with NEP personnel. In addition, the team listened to 

-2- 

84 
presentations by NEP personnel , and observed work activities in progress. 

This report describes the specific areas examined, the documentation reviewed, 

and the team’s findings. Appendix C to this report, list the persons who 

participated in and were contacted during this inspection. 

During its final exit meeting with NEP management and staff, on September 1, 

1995, the team summarized the open items, as well as NEP’s weaknesses and 

strengths, and the team’s observations. 

3.3 Nuclear Qual i tv Assurance 

The team found that NEP’s QA process was governed by the QA topical report 

(NEDO-11209-04A). GENE’S stated policy was to obtain quality leadership, and 

to achieve and maintain high quality in products and services through timely 

and effective compliance with all quality requirements. The QA topical report 

described the QA program used by NEP to fulfill the regulatory aspects of this 

pol icy. The team observed that a1 1 managers within GENE with qual i ty-re1 ated 

responsibility had full authority to implement the applicable elements of the 

program within their respective areas of responsibility. 

the GENE QA program, a basic responsibility of each organization within GENE, 

had the unqualified endorsement and support of GENE’S Vice President and 

General Manager. 

Implementation of 

The team noted that Nuclear Quality Assurance (NQA) reports to the General 

Manager of Nucl ear Operations , who reported to GENE’ s Vi ce President and 

General Manager. NQA was a staff organization assigned responsi bil i ty for 

establishing the GENE level qual ity-related policies and procedures (P&Ps) . The GENE line organizations, including NEP, had the responsibility for 

ensuring conformance with applicable design and QA requirements. Effective 

August 25, 1994, the NEP Manager, NQA, began reporting to a newly established 

Manager of Environmental Health & Safety and Nuclear Quality Assurance with 

direct access to the NEP General Manager. The team found the following NQA 

subsections were in effect as of August 16, 1994: 

Fuel QA, located in Fuel Manufacturing Operations (FMO) building 

Components QA, located in Fuel Components Operations (FCO) building 

Services QA, located in Service Components Operations (SCO) building 

Purchased Material Quality Control (QC) 

Quality Audits & Programs 

Fuel Engineering QA 

GENE Systems 

GENE Audits 

Quality at the Source and New Product Introduction (NPI) 

QA Programs Japan 

QA Programs Europe 

Software QA & NPI 

The team concluded that the QA topical report and the NQA organization were, 

for the most part, implemented in an acceptable manner for the areas evaluated 

during this inspection. 

-3- 

85 
3.4 Reload Core Desian, Safety Analysis. and licensing Processes 

To evaluate NEP’s reload core design, safety analysis, and licensing 

processes, the team reviewed the performance, interfaces, and documeritation of 

the reload analysis processes. The reload analysis activities were found to 

consist of determining the 1 icensee requirements, bundle design and core 

loading pattern; performing the steady-state and transient neutronic and 

thermal hydraulic analysis; and updating the cycle-specific reload licensing 

analysis and the process computer databank. 

performed these processes based on the technical design procedures (TDPs), and 

documents the results in a series of design record files (DRFs). 

Specifically, NEP prepared a DRF for the bundle design, reload licensing, core 

management, and process computer update. Technical Project Manager (TPM) has the overall responsibility for the reload 

analysis, and was assisted by other engineers as needed. 

paragraphs describe the areas evaluated by the team and summarizes the team’s 

f i ndi ngs . 

3.4.1 Licensee Requirements 

The team observed that NEP 

The team also noted that the 

The fol.lowing 

The licensee requirements concerning a fuel reload for a given reactor cycle 

were specified and documented in the fuel release engineering data (FRED) and 

the operating plant licensing parameters (OPL-3). 

theoretically contained all of the basic information to perform the analyses 

for the reload licensing submittal. This database had a very structured 

methodology for verifying (between NEP and the licensee) that each OPL-3 

information entry was correct and agreed to by both parties. There were three 

signoffs for each information entry in the OPL-3 database: one each for NEP 

and the licensee, and one for resolution of any differences between NEP’s data 

entries and those of the licensee. When the OPL-3 database was agreed to by 

both parties (when a17 differences are resolved), the licensee sent an 

agreement letter to NEP and NEP entered the OPL-3 database into the BoilingWater Reactor Engineering Data Bank (BWREDB) for specific design analyses. 

The OPL-3 database 

FRED consists of the licensee’s energy utilization plan (EUP) requirements and 

the domain of the operating cycle, including a list of equipment out of 

service during the cycle. The team determined that NEP does not verify FRED 

with the licensee in the same manner as the OPL-3 data. 

simply provides FRED to NEP to enter into the BWREDB. During the review of 

the reload core design, safety analysis, and licensing processes for Georgia 

Power Company, Edwin I. Hatch Nuclear Plant Unit 1 (Hatch 1) Cycle 16, the 

team questioned why NEP and the licensee had not formally reviewed the FRED 

database as they had for the OPL-3 data, particularly since some of the data 

from each document was the same and shared by some calculations. The team 

concluded that not verifying FRED with the licensee in the same manner as the 

OPL-3 data was a weakness that is further discussed in Section 3.4.4.2(2) of 

this report. 

Rather, the licensee 

-4- 

86 
3.4.2 Reload Core Design Process 

The team reviewed the reload core design process as follows and identified 

that the bundle design process for the next fuel cycle (cycle n) began with 

the cycle n-1 core inventory and the licensee's EUP for cycle n. 

performed the bundle physics calculations using the TGBLA computer pr;ogram 

and three-dimensional (3D) core simulation calculations using PANACEA . NEP 

evaluated the preliminary bundle designs and core loading patterns -via an 

iterative design process to determine an acceptable bundle performance. NEP 

then produced the bundle design report and completed and closed the bundle 

design DRF. 

NEP 

The determination of an acceptable reload design began with the work 

authorization (WA), which (via the Product and Performance Specification) 

identified Items such as the applicable documents, procedures, design bases, 

and inputs from two design inte iface documents (FRED and OPL-3). 

NEP reviewed the nuclear design 

on the choice of hot and cold e 

then determined the reload core 

pattern that satisfied the EUP, 

bases for the reload, with specific emphasis 

genvalues used for the 3D simulations. 

hydraulics data as well as a reference loading 

any special licensee requirements, and the 

The team noted that the reference loading pattern was the basis for 

NEP 

nuclear design bases (e.g., hot excess reactivity, thermal limits, shutdown 

margins). 

the reload core licensing. 

3.4.2.1 Eigenvalue Selection and Uncertainties 

The team was aware of NEP's recent problems with the accuracy of the PANACEA 

hot and cold reactivity (eigenvalue) predictions. 

recent trend toward longer fuel cycles and new bundle designs, which utilize 

higher U enrichments and gadolinium (Gd) loadings, had contributed to 

increaseg scatter in the predicted hot and cold eigenvalues. 

The team observed that the 

NEP told the team that the reload core design process was modified in 1994 to 

add a technical review of the design-basis eigenvalues for a specific reload 

core. "Determination of Critical Eigenvalues," TDP-0012, Revision 2, M.E. 

Harding, May 1995, Volume 4: Nuclear 1, was revised to reflect this 

requirement. 

The team observed that for the Philadelphia Electric Company, Limerick 

Generating Station Unit 1 (Limerick 1) Cyclf 6, the cold eigenvalue was 

mispredicted in 1994 by 0.8 percent delta K (961yo. Similarly, in 1995, for 

'GENE Fuel Bundle Lattice Nuclear Design Model (TGBLA), a lattice physics 

code that provides lattice-averaged diffusion cross sections and relative rod 

power peaking for the BWR simulator core calculations 

program representing the BWR core, exclusive of the external flow loop 

'A three-dimensional (3D), coupled nuclear/thermal lhydraul ic computer 

3Percent of Reactivity Addition 

-5- 

87 
the Boston Edison Company, Pilgrim Nuclear Power Station Unit 1 (Pilgrim) Cycle 11, the cold eigenvalue was mispredicted by 0.9W. NEP presented to 

the team data from other plants showing significhnt variation and trends in 

the cold and hot eigenvalues within a cycle and from cycle to cycle. 

pursued an in-depth review of the TGBLA and PANACEA benchmarking, the 

eigenvalue selection process, and NEP nuclear methods development activities. 

The following paragraphs summarize the team's findings. 

NEP told the team that the computer code versions currently in use were 

TGBLAO4 and PANAC09, and that the new versions which will be put into 

production by February 1996 are TGBLAO6 and PANACll. The team reviewed the 

benchmarking documentation for TGBLA04, and compared it with the earlier 

version, TGBLA03. 

identical results for the older 8x8 lattice designs. The TGBLA04 Software 

Test Report documented comparisons between MERIT, NEP's reference Monte-Carlo 

benchmark code at that time, and both TGBLA04 and TGBLA03. These results 

showed that TGBLA04 agreed with MERIT better than TGBLA03 fo.r the newer 9x9 

and 10x10 designs; however, the team noted that the eigenvalues calculated 

using TGBLA04 ranged from 0.015 AK sma: ;er to 0.003 IK larger than MERIT. 

The team 

This comparison revealed that both versions produced nearly 

In Section 2.1(1), "NRC-Approved Models," of NEDE-31917P, "GEll Compliance 

With Amendment 22 of NEDE-24011-P-A (GESTAR II)," dated April 1991, GENE 

stated that "extensive comparisons have been made with Monte-Carlo results to 

ensure that the 9x9 simulations are as accurate as the previous 8x8 

experience.'* However, the team concluded that this statement was not well 

supported. The team observed that '3 very limited number of benchmark 

comparisons were made (three lattices at 40% and 70% void) for the GEll 

design. In addition, the team expressed the concern that the comparisons were 

not consistent between the 8x8 and 9x9 designs, and that the relative accuracy of TGBLA04 for the GEll design had not been established. 

In discussions concerning the development history and benchmark program for 

the new code versions (TGBLA06 and PANAC11) NEP personnel told the team that 

the lattice physics code, TGBLA06, contains various enhancements. These 

enhancements reportedly include features that will improve the accuracy for 

new designs with higher U 

predictions for Gd and wa%- rods, as well as various types of fuel and Gd 

rods nfar water rods. NEP also told the team that development of qtransport 

theory code is in progress to replace the current diffusion theory code. 

In addition, NEP told the team that the 30 core simulation code, PANAC11, 

contains enhancements to its diffusion theory model , spectral and control rod 

history models, and rod power model. 

enrichments and Gd loadings by improving 

4A theory for the treatment of diffusion in a medium of neutrons or gamma 

'An approximate theory for the diffusion of particles, especially 

rays, based on the Boltzmann transport equation 

neutrons, based on the assumption that in a homogeneous medium, the current 

density is proportional to the gradient of the particle flux density 

-6- 

88 
The team examined the DRF documenting the TGBLA06 benchmark, software testing, 

and design review. The team observed that Monte-Csrlo Neutron Particle (MCNP) 

code, with an NEP-enhanced nuclear data library, was used for the new 

benchmark, and that a comprehensive test matrlx was run. The DRF contained 

comparisons between MCNP, TGBLA04, and TGBLA06 for 8x8, 9x9, and 10x10 

lattices with various enrichment, Gd loading, and water rod configurations. 

Calculations were performed at O%, 40%, and 70% void, cold, controlleduncontrolled, and bora+?d conditions. The team observed that the eigenvalue 

root mean squared (RMS) , [deleted pursuant to 10 CFR 2.790 - document 

described specific values], for [deleted pursuant to 10 CFR 2.790 - document 

described a specific value] cases was [deleted pursuant to 10 CFR 2.790 - 

document described a specific value]%AK. The team also observed that the 

TGBLA06 results were significantly better than the TGBLA04 results. 

addition, the scatter in K-infinity (Ka) differences compared with MCNP was 

significantly less for TGBLA06 than for TGBLA04. 

In 

The team also reviewed the PANACll benchmarking, testing, and design review 

documentation, and examined TGBLA06/PANAC11 benchmark comparisons to operating 

data from nine different reactors with three to five cycles per reactor. 

These results showed that, comparea c~ the earlier versions, TGBLA06IPANACll 

s igni f i cantly reduced the cycl e-to-cyc? e d i scont i nui ty i n K-effect i ve (Keff) 

(beginning of cycle (BOC) n - end of cycle (EOC) n-1 = [deleted pursuant to 

10 CFR 2.790 - document described specific values]), as well as the downward 

drift of Keff within the cycle. The team also noted small improvement in 

power distribution accuracy, as indicated by traversing in-core probe (TIP) 

comparisons ([deleted pursuant to 10 CFR 2.790 - document described a specific 

value]% RMS vs [deleted pursuant to 10 CFR 2.790 - document described a 

specific value]%). 

The team reviewed the hot and cold eigenvalue selection and review processes, 

and discussed its findings with NEP management and the TPMs. 

team reviewed TDP-0012 for determining hot and cold eigenvalues, and compared 

the procedure to recent plant DRFs. The team observed that the empirical 

methodology in TDP-0012 for adjusting the hot and cold eigenvalues did not 

necessarily improve predictions, as indicated by the Limerick 1 Cycle 6 and 

Pilgrim Cycle 11 results described below. The team also observed that the 

TPMs were considering the prior cycle eigenvalue errors, critical value vs 

prediction, for both the hot and cold eigenvalues. In addition, the team 

noted that TDP-0012 required a formal review of the eigenvalues to be used 

before each reload safety evaluation was initiated. 

documentation from this process for several reloads. 

In addition, the 

The team evaluated 

The team concluded that the 9x9 lattice benchmark did not support GENE’S 

statement in NEDE-31917P pertaining to lattice simulation accuracy. 

Specifically, the team expressed the following concerns: 

Carl o benchmark code. 

Limiited comparisons were made between the 9x9 lattices and the MonteVery limited inter-comparisons of these 9x9 benchmarks were made 

against the comparable 8x8 lattices comgared to the Monte-Carlo benchmark 

code. 

-7- 

89 
The comparisons that were made did not show consistent results, 

The re1 ative accuracy was not establ i shed. 

A considerable number of comparisons between TGBLAOI and TGBLA03 

results were made for 9x9 lattice conditions; however, the team did not 

conzider these to be benchmarks in conformance with the statement in 

Amendment 22 of GESTAR 11. 

The team concluded that the errors in calculated hot and cold core reactivity 

for the recent reloads with the newer fuel designs and long cycle lengths was 

a weakness of the NEP reload design process. The team noted, however, that 

NEP has addressed this weakness by implementing steady-state nuclear methods 

improvements (discussed in Section 3.4.2.3 of this report). 

3.4.2.2 Cold Shutdown Margin 

At the start of each hew operating cycle, each licensee preforms a cold 

critical shutdown margin demonstration test. 

technical specifications require a cold shutdown margin ranging from 0.25 to 

0.38%. The team was told that NEP typically designs a reload core to have a 

shutdown margin of at least 1W. The team observed, however, that should the 

recent NEP mispredictions of cold eigenvalues continue, the potential exists 

to have less shutdown margin than required. For example, several plants have 

missed the cold shutdown K by approximately 0.9W. 

determined that had the cores been designed to a 1% shutdown margin, the 

misprediction would have resulted in a shutdown margin cf 0.1 to 0.2W 

compared to the required 0.25 to 0.38W. The team postulated that the 

differences between the predicted shutdown margin and the true shutdown margin 

could be attributed to inaccuracies in the TGBLA/PANACEA model, inaccuracies 

in modeling the true reactor conditions during the test, or the test procedure 

i tsel f. 

The team noted that 1 icensee 

The team’s evaluation 

To evaluate these concerns, the team reviewed indepth three recent cold 

critical shutdown tests (Duane Arnold Cycle 14, Pilgrim Cycle 2, and 

Limerick 1 Cycle 6), as discussed in the following paragraphs. 

In each case, the calculated cold shutdown margin for each reload core is 

reported in the supplemental reload 1 icensing report (SRLR). 

shutdown margin is reported in terms of the cold, BOC Keff for the core 

condition of all inserted control rods, except the strongest worth rod. In 

addition, the reload R-value, maximum increase in cold core reactivity with 

exposure expressed in AK, is reported. The following paragraphs summarize the 

team’s findings: 

The cold 

(1) Duane Arnold Cycle 14 

During the Cycle 14 start-up of the Duane Arnold Energy Center plant, 

the calculated limiting eigenvalue was found ta have a margin of only 

.07% in excess of the technical specification requirement. Preliminary 

assessment had indicated that the calculated shutdown margin in 

reactivity was actually less than the technical specification value, 

-8- 

90 
causing a ten-hour delay in plant start-up. NEP characterized the 

incident as a "near-miss," and prompted a root cause evaluation by GENE 

(memorandum from G.E. Dix to J.S. Armijo on "Root Cause Evaluation of 

Duane Arnold Shutdown Margin, I' dated May 16, 1995). The calctll ated 

shutdown margin was found to be less than the nominal design value by 

about a factor of 2. The team deterirt',led that such large differences 

between the calculated shutdown margin and the nominal design value 

raise concerns regarding the potential for reduced reactivity margins in 

the safe shutdown capability of the core, as well as the accuracy of the 

design methods (and/or their implementation) used to ensure that the 

reactor core can safely be operated. 

The relatively large deviation of the calculated core shutdown margin 

from the nominal design value for Duane Arnold Cycle 14 is not an 

isolated incident, and GENE had been aware of the potential issues 

mentioned above. For example, an 0.8% deviation of the Limerick 1 

Cycle 6 calculated cold critical eigenvalue from the design target value 

prompted a memorandum from 3. E. Wood, Manager, Design Process 

Improvement to the TPM, regardjng "Cc'ld Critical Eigenvalue," dated 

April 11, 1994. The memorandum addttessed the regulatory and operational 

implications of the inconsistency between the calculated and selected 

target eigenvalues. It also provided the following specific directions, 

among others: 

Select cold critical eigenvalues on a best estimate basis. 

4' Provide sufficient cold shutdown margin in the design to meet 

Provide a special review of the cold critical eigenvalue 

technical specification requirements. 

select ion process. 

A later memorandum regarding "Critical Eigenvalue Review," dated 

August 10, 1994, addressed the issue at greater length. The memorandum 

characterized the difference between the calculated critical eigenvalue 

and 1.0000 as separable into two components: 

(a) a predictable bias that can be specified at the time of core 

design for a specific operating cycle 

(b) an uncertainty that may cause the calculated critical 

eigenvalue to move up or down, with respect to the bias, during the 

operating cycl e 

The team noted that the objective of the design-basis eigenvalue is to 

negate the impact of the predictable bias. The memorandum of August 10, 

1994, also specified target RMS differences between the design-basis and 

calculated critical eigenvalues for the BWRs in operation. These target 

RMS differences had been specified for BOC co'c! and EOC hot conditions, 

as well as for the hot eigenvalue trajectory difference. The memorandum 

noted that the BOC cold critical eigenvalue predictive capability did 

not meet the accuracy target values. However, the memorandum noted that 

-9- 

91 
the EOC hot eigenvalue predictive capability met the accuracy target 

values, while the hot critical trajectory predictive capability met the 

accuracy target values only for the low cycle energy cores. 

memorandum a1 so recommended design process improvements involving both 

eigenvalue selection and the nuclear models used in the design 

cal cul at ions. 

The 

The team’s review of the root cause evaluation concerning the erroneous 

Duane Arnold Cycle 14 eigenvalue caiculation revealed that a little less 

than one half of the difference between the calculated and design-basi s 

eigenvalues was attributable to an error in applying the design process. 

Specifically, the results of an improved method for eigenvalue 

temperature correction had not been properly accounted for. This human 

error aspect of the incident was discussed extensively in the root cause 

analysis. The analysis also identified a number of remedial actions, 

including improved training, more explicitly written procedures, and 

better implementation of the design verification process. 

Even after accounting for the contribution of the error in temperature 

correction, the root cause evaluation revealed that a recent trend in 

the shift in eigenvalues in the BWRs in operation had not been factored 

into the design process. (This last observation is consistent with the 

observations reported in the memoranda written by J.E. Wood.) 

Specifically, the team noted that the eigenvalue shift noted in the root 

cause evaluation occurred at Duane Arnold despite one of the 

recommendations made in Wood’s memorandum of August 10, 1994. (That 

recomnendation involved an improvement in the eigenvalue selection 

process, being incorporated in the TDP-0012, and implemented in the 

Duane Arnold Cycle 14 design calculations.) 

in the Wood’s memorandum, improvements in the nuclear models, had not 

been addressed in the root cause evaluation; however, such improvements 

were discussed in some detail at a presentation to the team by NEP 

during the course of this inspection. Significant aspects of the 

improvements in nuclear models are discussed below in Section 3.4.2.3 of 

this report. 

The second recommendation 

(2) Pilgrim Cycle 11 

The team’s review noted that for Pilgrim Cycle 11, the SRLR specified a 

cold Keff of 0.986 to yield a BOC cold shutdown margin of 1.4M. The 

specified R-value was 0.003 to yield a minimum shutdown margin of 1.1% 

sometime later in the cycle. 

the determination of the cold and hot target eigenvalues were performe.’ 

according to the required process and relevant TDPs. 

by TDP-0012, Revision 1, were used to adjust the hot critical 

eigenvalues for the amount and location of the Gd rods and cycle 

exposure, and to adjust the cold eigenvalues for exposure. 

noted that NEP compared predicted and measured results from Pilgrim 

Cycle 10 to other units. 

The calculations of these values including 

Methods prescribed 

The team 

- 10 - 

92 
I The NEP evaluation team then conducted and documented a technical review 

of the hot and cold eigenvalue selections report, in accordance with 

'TDP-0012. The team's review noted that the results of the technical 

review were a1 1 based on the TGBLAO4 and PANACEA09/10 model s; TGBLA06 

and PANACEA11 checks were not made. 

I The actual measurements showed a 80C cold shutdown margin of 

approximately 0.496111(. Thus, in the BOC cold shutdown with the most 

reactive rod out, the core was approximately 1.OW more reactive than 

predicted by the accepted procedures and computer codes. Pilgrim 

conducted the actual measurements using local critical measurements 

(i.e., withdrawing the most reactive rod (object rod) with an adjacent 

rod (margin rod) set at the required shutdown margin) rather than insequence critical measurements (i .e., control rods are removed by bank 

in-sequence so that the core remains approximately symmetrical over the 

rods). 

the core was approaching criticality. Subsequently, with revised data 

from NEP, Pilgrim repeated the measurement, and made two additional 

shutdown margin measurements using two other strong rods with adjacent 

rods as the margin rod. 

Pilgrim terminated the initial measurement when it appeared that 

NEP performed a substant i a1 analysis to understand and isolate possible 

contributors to the 1.OW misprediction. These calculations identified 

two such contributors. The more significant contributor was the lack of 

burnup mesh points in the nuclear data available to PANACEA. (The mesh 

points in burnup space did not closely match the bundle peak Ka values.) 

The less significant contributor resulted from contro? rod depletion. 

These two effects contributed to a total of 0.4%AK, leaving 0.6W 

unexpl ai ned quanti tat i vel y . 

Pilgrim a1 so performed i n-sequence measurements, and observed some 

misprediction in the cold BOC critical K. However, this misprediction 

at 0.3MK was not large. The significance of this value is the 

comparison against the local critical misprediction of 1.0961y(. That is 

there is a 0.7W prediction bias between the two types of cold 

measurements. The team noted that the NEP shutdown margin measurement 

using in-sequence critical values under-estimated the Pilgrim Cycle 11 

BOC cold shutdown margin compared to a direct measurement using local 

critical values as conducted by Pilgrim. At the time of the inspection, 

the Pilgrim- specific cause or more likely causes of this 0.7M were 

not identified for the team. 

The team determined that with a BOC value of 0.4'161y(, and a SRLR R-value 

of 0.3%Lu(, it is possib7e that the required shutdown margin at the most 

rqactive burnup point will not be met. 

the stated values are rounded, and the R-value may have decreased. 

Therefore, the team identified the need to determine the current 

shutdown margin value at the most reactive time in Cycle 11. 

considers this an open item, and requested that NEP notify the NRC upon 

identifying this value. 

"Possible" is used here since 

The team 

(Open Item 95-01-01) 

- 11 - 

93 
(3) Limrick 1 Cycle 6 

The team’s review indicated that for Limerick 1 Cycle 6, the SRLR 

specified a cold Keff of 0.980, thus implying a BOC cold shutdown margin 

of 2W [(1.00 - 0.980) x 1001. The specified R-value was 0.004, 

indicating that the shutdown margin was at a minimum sometime later in 

the cycle, and that the margin was 1.6W. 

values including the determination of the target cold eigenvalue was 

reviewed by the team and compared to the actual results for Cycle 6 

operation. 

The calculations of these 

The Limerick reload 1 icensing DRF documented the process of selecting 

the eigenvalues. 

Cycle 4 value was 0.998. An earlier version of the TDP-0012 empirical 

AK correction methodology produced a correction of +O.O047AK. 

both the prior cycle results and the correction methodology indicated 

that the BOC cold eigenvalue should be about 1.004 to 1.005, 

cold eigenvalue for BOC Cycle 6 was set at 1.004 via the eigenvalue 

selection process. 

The Limerick 1 Cycle 5 BOC Keff was 1.004, and the 

Thus, 

The target 

The calculated PANACEA eigenvalue at the actual plant cold critical 

condition was 0.996 which resulted in a reactivity misprediction of 

0.8W. The its review, the team determined that this implies that the 

BOC cold shutdown margin may have been 1.2%, instead of the calculated 

2%. 

large shutdown margin, the technical specification shutdown margin was 

satisfied. 

The team concluded that because this reload was designed with a 

From its evaluation of these cold critical shutdown tests as well as the 

evidence from Pilgrim Cycle 11 predictions and measurements led the team to 

conclude that NEP’s current methods are not adequate to predict the technical 

specification requirements for the most reactive rod-out shutdown margin based 

on in-sequence critical measurements. The team determined that NEP’s present 

errors in target RMS deviations in eigenvalues, and to adequately quantify 

biases and uncertainties in eigenvalue calculations especially for high-energy 

cycle, heterogeneous cores was a weakness. 

eigenvalue calculations relates to the issue concerning the accuracy of design 

methods used to ensure that the reactor can be safely operated, as discussed 

in Section 3.4.2.3 below. 

The issue of accuracy in 

3.4.2.3 Steady-State Nuclear Hethods Improvements 

NEP explained to the team that recent trends in BWR operation and fuel 

management are characterized by 1 onger cycles, higher fuel exposures, fuel 

enrichments, burnable poison loading, and lower design margins to reduce fuel 

cycle costs. 

traditional GENE nuclear predictive capabilities. 

TO enhance their predictive capabi:ity, NEP has initiated a phased program 

designed to improve the lattice physics and core physics models used in the 

design calculations. 

improved cross-sections, better neutron moderation and resonance shielding 

NEP stated that these trends have posed a challenge to 

The team found that the lattice physics changes included 

- 12 - 

94 
treatment, more detailed modeling of burnable poison depletion, improved 

spectral weighting of diffusion coefficients, and ap, improved subchannel void 

distribution model. These efforts were based on a significant number of 

lattice conditions, and thus also provide a basis for statistical 

quantification of reactivity differences for the design process. 

Core physics enhancements included improved neutron diffusion models, improved 

cold temperature and rod power models, and better accounting of spectral 

history and control rod history. NEP reported that the updated nuclear models 

and codes were being benchmarked, and will be used as design tools after the 

first quarter of 1996. NEP also reported that application of the new design 

mode7s shows that the new models significantly improve performance for 

approximately 30% of the plants with high-energy cycles and highly 

heterogeneous cores, where traditional design methods do not provide the 

target predictive accuracy. 

performance for approximately 70% of the plants with lower-energy cycles and 

more homogeneous cores, where traditional methods already provide the desired 

predictive accuracy. 

The team concluded that the new design methods, therefore, have the potential 

to resolve at least some of the presently observed shortcomings in eigenvalue 

predictions. Additionally, NEP had undertaken measures to reduce the 

incidence of human errors by improving procedures and training engineers in 

the proper use of the procedures. 

The new models also provide slightly improved 

NEP management recognized the team’s concerns. 

and plans, and participated in several discussions with the team on the 

subject of shutdown margin. 

following plans to address these concerns: 

NEP presented its evaluation 

In addition, NEP informed the team of the 

GENE prontised to release a service information letter (SXL) to its 

customers bay February 1996, addressing the eigenvalue and shutdown margin 

concerns. 

NEP will rely on its review process to ensure safe shutdown margins. 

Parallel predictions will be performed using both the current 

(TGBLA04/PANAC9) methodol ogi es and the new (TGBLA06/PANACll) methodologies and 

shutdown margins will be checked with the new methodologies to ensure the 1% 

design-basis margin. 

Future cold PANACEA models wi 11 have mu1 ti-temperature neutronic data 

from TGBLA. 

GENE will revise the appropriate TDPs to reflect the new methodology. 

NEP will review the cold critical test procedure with licensees to 

GENE will accelerate the development of new methods, including the 

determine if a more appropriate test for shutdown margin should be performed. 

transport theory lattice code. 

- 13 - 

95 
The team concluded that the near-term use of TGBLA06/PANAC11 in combination 

with the eigenvalue selection process should help reduce uncertainties in the 

cold critical and shutdown margins. The team also concluded that the 

assurance of adequate shutdown margin can be strengthened by joint GENE/licensee actions consistent with the plant startup safety analysis. The 

team also discussed with NEP other near-term actions that may include 

analytical changes, such as increasing the design shutdown margin from 1 to 

2%, ciA implementing test procedure changes (such as performing a strongest rod 

out cold critical measure instead of an in-sequence critical measure). 

In reaching its conclusions concerning this issue, the team considered the 

information provided in Wood’s memoranda, as well as the root cause evaluation 

of the Duane Arnold Cycle 14 eigenvalue calculation, and the presentations 

made on the improved steady-state nuclear methods. 

calculations, specifically, these requirements are the ability to quantify the 

predictable biases and uncertainties, and to demonstrate that the target RMS 

deviations in eigenvalues are being met. 

eigenvalue calculation have fallen short of its own expectations in the recent 

past. However, the team also concluded that GENE has a process in place, 

involving improved nuclear models arid Detter procedures and training, that has 

the potential to resolve this problem. However, the team observed that to 

ensure that NEP’s effort is meeting its intent, NQA review and monitoring would be appropriate. 

For shutdown margin 

GENE’s predictive capabilities in 

The team concluded that, given that the introduction of the new nuclear models 

and codes constitutes a major upgrade of GENE’s nuclear design methods, the 

use of the new models as design tools after the first quarter of 1996 is not 

an unreasonable schedule. 

3.4.3 Reload Safety Analysis Process 

During its review of the reload safety analysis process, the team found that 

NEP performs the reload safety analysis for approximately 20 reload cores each 

year. The team noted that the reload safety analysis process was highly 

automated and computerized, with minimum user involvement in the code-to-code 

and code-to-database interfaces. The reload safety analysis is initiated by 

the final release design basis work authorization (WA), the fuel application 

design basis WA, and/or the Licensing WA. 

with coordinated with the licensee and within NEP by the Fuel Project Manager. 

The responsible engineer performs and documents the reload safety analyses and 

the TPM oversees the technical details of the analysis. 

The reload licensing activities 

During its review of the process, the team found that the OPL-3 and the FRED 

documents, together with the WAS, provide the plant operating and design data 

required for the reload analysis. The OPL-3 is an extensive document, which 

includes initial operating conditions (e.g., power level, flows, and 

temperatures) , scram parameters (e.g. , setpoints, response times and delays) , and plant equipment parameters (e.g., volumes, lengths, and capacities). 

After resolving all comments and finalizing the data, the licensee provides NEP with a final signed and verified version of the OPL-3. 

- 14 - 

96 
The FRED document provides fuel bundle identification and exposure data, cycle 

energy requirements, selected cycle margin and operational flexibility 

improvement options, 1 imiting transient analysis statepoints, technical 

specification requirements, and the plant equipment status. 

The team found that the NEP reload safety analysis process is documented and 

controlled using an extensive set of TDPs, nuclear engineering technical 

procedl;;es documents (e.g., the GETAB design procedure document) , design bases 

(Des) documents, and engineering operating procedures (EOPs) given in NEDE21109. The TDPs and EOPs describe the overall methodology, as well as the 

task-specific methods, assumptions, and interfaces. The EOPs provide the 

general requirements concerning computer codes, design review/veri f icat ion, 

DRFs, WAS, engineering records, and reload licensing. In addition, NEP is 

currently establishing a system of "analysis block guides" that will provide 

additional detail for performing and verifying the reload analyses. 

The found that the results of each reload analysis are documented in detail in 

a system of DRFs, according to the requirements of EOP 42-10.00, "Design 

Record File," dated March 27, 1995. Tbe detailed results of each of the major 

reload tasks (e.g., reload 1 icensing, core monitoring, and bundle design) are 

documented in a separate DRF. The responsible enqineer compiles and maintains 

the DRF. 

In order to ensure the quality and traceability of the engineering data, NEP 

employs the computerized BWREDB system. 

retrieve engineering input/output data, and to facilitate automation of the 

reload design and safety analysis process. 

previous cycles, as well as the data necessary for the current reload cycle 

analysis. 

3.4.6.1 of this report. 

The BWREDB is used to store and 

The BWREDB includes data from 

The team's review of the BWREDB system is described in Section 

The team determined that the NEP's reload safety analysis provides the 

technical basis to support cycle operation and ensure adequate margin to the 

safety 1 imi ts during normal operation and anticipated operational occurrences 

(AOOs). The analysis consisted of several major evaluations, including 

nuclear design-basis analysis, core hydraulics analysis, determination of the 

reference core loading pattern, transient analysis, loss-of-feedwater heating 

(LFWH) analysis, rod withdrawal error (RWE) analysis, mislocated fuel bundle 

(MFB) analysis, loss-of-coolant accident (LOCA) analysis, and (i) the 

stabi 7 i ty analysis (where required). 

The team noted that the reload analysis also includes a fuel-dependent 

evaluation of the LOCA and a determination cf the peak clad temperature, clad 

oxidation fraction, and maximum average planar linear heat generation rate 

(MAPLHGR) limit versus fuel exposure. SAFE/REFLOOD/CHASTE code, using gap conductance and fission gas release data 

detemi ned by GEGAP/GESTR6, or SAFER/GESTR-LOCA. 

This analysis is performed using the 

6GE-NE Stress and Thermal Analysis of Fuel Rods (GESTR), a computer 

program to provide best-estimate predictions of the thermal-mechanical 

performance of nuclear fuel rods experiencing variable power histories 

- 15 - 

97 
Currently, U.S. BWR nuclear plants are not required to submit a reload 

stability analysis. Some 1 icensees, however, perform a cycle-specific 

stabil i ty analysis to evaluate the core and channe i hydrodynamic stabi 1 i ty 

margins for selected points on the power-flow operating map. The core and 

channel decay ratios are determined with the ODYSY (frequency-domain) 

stability code using ISCOR thermal-hydraulic input and cycle-specific dynamic 

parameters (e.g., reactivity coefficients and neutron 1 ifetime) 

The team also noted that in addition to the safety analysis, the cycle reload 

evaluation includes the fuel bundle design, fuel cycle analysis, core design 

and management, and core monitoring analysis. The fuel bundle design analysis 

is performed to determine the bundle neutronics characteristics including 

reactivity coefficients, pin-wise power distribution, and nuclear data for 

input to PANACEA. The fuel cycle analysis evaluates the reload cycle core 

performance, and the core management analysis provides the uti1 ity information 

concerning reactor startup, reactivity curves, and operating strategy. The 

core monitoring analysis is performed to define the site process computer data 

bank. 

The team evaluated NEP’s task-wise verifiLation of the individual analyses 

(e.g., verification of the rod withdrawal accident analysis), and found that 

it consisted of a committee review of critical steps in the reload process 

(e.9. , the eigenvalue selection and transient selection reviews), and an 

overall re1 oad 1 i cens i ng qual i ty review (RLQR) , as descri bed i n Section 

3.4.4.2 of this report. Many of these verifications were performed with 

detailed checklists that identified the critical issues in the reload 

eva7uation. 

members of the i4EP reload review committees. Documentation of NEP’s 

verification was included in the DRF. 

The team also noted that utility staff frequently participated as 

After NEP verified the analyses and results, NEP entered the reload data to 

the BWREDB. 

in the SRLR and the NOR. 

NEP provided the results of the reload evaluation to the utility 

The team concluded that these evaluations were generally performed using the 

methods described in Amendment 22 of GESTAR I1 and that NEP’s verification of 

the reload evaluation was extensive. 

3.4.3.1 Cycle-Dependent Safety Limit Uinimum Critical Power Ratio 

In discussions with the team, and also at a meeting with the NRC staff on 

March 16, 1995, NEP indicated that it intends to submit a cycle-dependent 

SLMCPR methodology for application to reload cores. 

cycl e-dependent method01 ogy wi 11 use the pl ant cycl e-speci f i c power 

distribution rather than the bounding power distribution used in current 

SLMCPR determination. 

SLMCPR, as well as the conservatism in the CPR thermal margin for some plants. 

While the final documentation of the methodology has not been completed, the 

cycle-specific SLMCPR was considered to be of subslmtial importance and was 

included in the team’s discussions with NEP during this inspection. 

following paragraphs summarize the team’s findings. 

According to NEP, the new 

The team determined that this change will reduce the 

The 

- 16 - 

98 
NEP reported that the present SLMCPR method was developed several years ago. 

Since that time, the method had become overly conservative for many reload 

cores because of the improved thermal margin of recent fuel bundle designs, 

and because of the increased number of core designs that must be bounded. 

present method calculated the SLMCPR necessary to ensure that 99.9% of the 

fuel rods avoid boiling transition. The ex;drLed number of rods experiencing 

boiling transition was determined with GESAM using a Monte-Carlo approach 

accounting for uncertainties in the GEXL correlation and the bundle power. 

The core statepoint was selected under the following conservative assumptions: 

The 

a bounding equilibrium core 

a core radial power distribution selected to maximize the number of 

fuel bundles at or near thermal limits 

a local power distribution selected to maximize the number of rods 

near boiling transition 

The team determined that the cycle-specific methodology relaxes the 

conservatism inherent in each of these assumptions. 

team’s discussions with NEP about the planned cycle-dependent SLMCPR 

methodology, the team made the following observations: 

On the basis of the 

Before el iminating the present conservatism in the SLMCPR 

determination, the team questioned whether this conservatism was required in 

the original methodology to accommodate other methods approximation? or nonconservatism (e.g., lack of CPR data, differences between the ATLAS tests 

and actual core operating conditions, statistical assumptions, and rod 

bowing). 

The present SLMCPR statistical method was a bounding core statepoint, 

and no allowance was required to account for uncertainty in the assumed 

statepoint. However, the cycle-specific approach involves, an analysis to 

determine the limiting core statepoint, and an allowance for uncertainty may be required in this determination. 

The MCPR determination currently incorporates channel bow effects by 

modifying the fuel bundle R-factor to account for the increased local peaking 

caused by channel bow. 

channel bow effects by increasing the R-factor uncertainty in the Monte-Carlo 

process. However, the Monte-Carlo analysis only accounts for the random 

component of the channel bow effect. 

component of the local power increase resulting from channel bow should be 

determined and included separately as an adjustment to the bundle R-factor. 

NEP indicated that the new methodology may include the 

Consequently, the systematic (or bias) 

7GENE’s 8.6-megawatt heat-transfer loop 

- 17 - 

99 
3.4.3.2 Rod Withdrawal Error Analysis 

The rod withdrawal error analysis determines the rod block monitor (RBM) 

setpoints, which ensure the necessary critical power and thermal mechanical 

linear heat generation rate (LHGR) margin. The team's review found that the 

core statepoint analysis and response to the control rod withdrawal are 

detfrmined with the PANACEA 3D simulator. In the analysis, the highest worth 

rod , with a fresh bundle immediately adjacent, is selected and withdrawn 

from the statepoint of peak hot excess reactivity, determined in the reference 

loading pattern analysis. 

RBM setpoints for plants equipped with the average power range monitor (APRM), 

RBM, and Technical Specification upgrade. 

3.4.3.3 Reload Licensing Transient Analysis 

A statistical approach was used to determine the 

The reload licensing transient analysis determines the transient 

overpressurization and reduction in margin to fuel thermal 1 imits and was 

performed using ODYN or REDY, 

transient analyses included the 1 imi ting pressure and power increase events, 

feedwater controller failure event, and the main steam iJolation valve (MSIV) 

closure (flux scram) overpressurization event. The ODYN one-dimensional 

(ID) computer code and the REDY point-kinetics transient code are used to 

analyze these events and evaluate the reduction in critical power ratio (CPR) 

and/or system overpressurization. 

LFWH, RWE, and MFB events. 

analysis of the LFWH transient). The GETAB code is used to analyze the fuel 

type-dependent operating 1 imit minimum CPR (OLMCPR) for these events, using 

the ODYN or PANACEA calculaftd heat flux, the ISCOR thermal-hydraulics data, 

and the fuel-dependent GEXL correlation. 

The team's review found that the reload 

The PANACEA code is used to analyze the 

(The NRC has approved the steady-state PANACEA 

The team found that the standard reload transient analysis uses precalculated 

core nuclear characteristics (neutron 1 ifetime; delayed neutron fraction; and 

doppler, void, and scram reactivity). NEP's transient analysis acceptance 

criteria were based on the requirements that (a) the number of fuel rods in 

b o i l i n g t r a n s i t i o n is limited to a maximum of 0.1%, (b) the cladding plastic 

strain is less than 1% and fuel centerline melt is precluded, and (c) both NRC 

and operational system and vessel pressure limits were not violated. 

'For plants equipped with the rod worth limiter, a four-rod gang is 

'Computer code that simulates the dynamic behavior of BWRs; ODYNM is 

withdrawn. 

applicable to plant analysis beginning with BWR/1 through BWR/4; ODYNV 

simulates dynamic behavior of valve flow control in 6&s, and is applicable to 

plant analysis beginning with BWR/S. 

"GENE Critical Quality Boiling Length Correlation (GEXL) computer code 

- 18 - 

100 
The team determined that since many licensee final safety analysis report 

(FSAR) transients are not limiting and/or have minimal sensitivity to reload 

dependent parameters, typical NEP reload analyses only included a selected set 

of potentially limiting transients. Typical limiting transients include the 

turbine trip, load rejection, pressure regulator failure, MSIV closure, LFWH, 

feedwater controller failure, and feedwater/high--pressure cool ant injection 

events. 

var! 3us assumptions concerning pl ant equipment performance in these analyses. 

lhe team observed that transient-specific TOPS and EOPs specify 

3.4.3.4 Quadrant-Synmnetric Fuel Failures 

During the team’s review, NEP provided a discussion concerning an incident in 

which fuel rod failures occurred in quadrant-symetric fuel bundles. 

bundles containing the failed rods were located in quadrant-symmetric 

peripheral locations with the reactor during the previous two cycles, and were 

moved into the central high-powered region of the core during the last cycle. 

When the first fuel rod failed, the control rods in the failed bundle location 

were inserted, as were the control rods in the symmetric locations. Although 

the power was reduced in the symmetric bundles with no fuel failures, fuel 

rods in these bundles failed later in the cycle (presumably because of 

operation before control rod insertion). 

The 

While NEP reported that the fuel failure mechanism is presently unknown, the 

team determined that the incident suggests a quadrant-symmetric mechanism that 

occurs at symmetric locations. The team noted that the fai?ure could be the 

result of an analytical error (associated with 3D-MONICOREY PANACEA, rod block 

setpoints, or RWE analysis, for example), since these tend to be quadrantsymmetric. Therefore, the team identified this issue as an open item, and 

requested that NEP notify the NRC upon identifying the failure mechanism. 

(Open Item 95-01-02) 

3.4.4 Reload Licensing Process 

To evaluate the reload licensing process, the team began its evaluation with 

the documents supplied to the licensee as the end result of the reload 

1 icensing process. 

The nuclear design report (NOR) is one of two NEP deliverables to the licensee 

resulting from the reload 1 icensing process. This report documents the reload 

core loading pattern and fuel bundle design, as we17 as the key assumptions 

and methodology for the analysis, and the anticipated nuclear and T/H core 

performance. In addition, the report documents the following topics: 

The following paragraphs summarize the team’s findings. 

analytical methods used in the analysis 

design objectives for the reload 

description of the core and fuel 

the loading pattern the thermal performance and reactivity behavior 

results from steady state ceutronic safety ciijculations 

- 19 - 

101 
The team reviewed the NDRs as part of the reload design evaluation, and 

concluded that the required information was adequately documented with 

traceable references. 

The supplemental reload licensing report (SRLR) is the second of the two NEP 

deliverables to the licensee resulting from the reload licensing process. 

Thi s report supplements GESTAR I I Amendment 22 (the GENE 1 i censi ng document) 

for the specific reload being evaluated. The SRLR summarizes the fuel in the 

reloaa core, the steady state and transient safety results, and the resulting 

cycle-specific SLMCPR. In addition, the SRLR docbments the following topics: 

the shutdown margins the anticipated operational occurrences (AOOs) analysis 

cycle specific margin improvements and operating flexibility 

the cycle SLMCPR and GLMCPR 

the results from the pressurization and na-pressurization events 

discussions concerning events or accidents not analyzed on a cyclespecific basis because they were bounded by the GESTAR I1 analysis. 

The team reviewed the SRLR reports as part of the reload design evaluation, 

and concluded that the required information was adequately documented with 

traceable references. 

3.4.4.1 Reload Core Monitoring Data Bank 

The site computer performs an online core performance evaluation, using the 

Core Monitoring Data Bank (CMDB) together with the reactor instrumentation 

(e.g., neutron flux, core flow, and temperatures), to determine the core power 

distribution and evaluate the fuel thermal limits. The team found that the 

input used to generate the CMOB includes a set of full-core PANACEA 

calculations at selected cycle statepoints from the current and previous 

cycl es . limits, and core loading data taken from the BWREDB. 

resulting CMDB includes the following engineering data: 

The input a1 so i ncl udes cycl e-speci f i c operat i ng opti ons , thermal 

Based on that input, the 

channel-type core maps 

nuclear cross-sections 

T/H constants 

pressure-loss coefficients 

reactivity coefficients 

heat balance data 

gamma and thermal TIP factors 

ARTS coeff i ci ent s 

R-factor data. 

The team concluded that generation of the CMDB is highly automated. The CMDB 

undergoes an extensive verification including comparisons of the site computer 

predictions (using PANACEA "measurement" data) with the corresponding PANACEA 

predictions. These predictions must satisfy specified design acceptance 

criteria concerning power distribution and thermal 1 imits. When the 

generation of the CMDB is complete, the data is stored in the BWREDB and 

transmitted to the licensee by the Fuel Project Manager. 

- 20 - 

102 
3.4.4.2 Reload Licensing Quality Review 

Reload Licensing Quality Review (RLQR) was the last step in the quality 

assurance and verification process to which the SRLR and the NDR are subjected 

before being released to the licensee. The team determined that tne RLQR 

primarily concerns the integration and consistency of the SRLR. 

constituted a higher-level verification of the engineering analysis and data 

utilized in generating the SRLR and the NDR. A summary of the RLQR was 

maintained in a DRF by the chairman of the Review Board. The team concluded 

that together with the Eigenvalue Selection Review Committee and the Transient 

Selection Review Committee (which have a narrower focus), the RLQR brings to 

bear a broader range of expertise on the engineering analysis process. 

It therefore 

The team found that the RLQR was formatted into two parts. 

the review ,involves an overview of the scope and key results of the reload 

analysis. For this part of the review, the Fuel Projects Manager and the 

Licensing Engineer were required to be present. The second part of the RLQR 

involved a more detailed review of the technical content of the reload 

analysis. This part of the review required the participation of an 

engineering review team with expertise in the five process areas (bundle 

design, fuel cycle, core design and core management, T/H and transient 

analysis, and core monitoring). The technical package contained the following 

key i nforma t i on : 

The first part of 

the nuclear design bases 

the reference loading pattern 

RWE, LFWH, and MFB analyses 

nuclear design parameters 

reload transient analyses 

the GENE thermal analysis bases 

reload core hydraulics 

the emergency core cooling system 

the stability analysis 

analysis 

The team determined that the RLQR process is generally serving its intended 

function of providing a fairly broad range of QA and design verification. 

Nonetheless, the team concluded that the errors in eigenvalue calculations for 

Duane Arnold 1 Cycle 14 indicate that the RLQR process is not a substitute for 

meticulous verification of each element of the design calculation. 

To evaluate NEP's reload licensing practices, the team conducted an indepth 

review of four recent reload packages (La Salle 1 Cycle 7, Hatch 1 Cycle 16, 

Limerick 1 Cycle 6, and Pilgrim Cycle 11). The following paragraphs summar;- 

the team) s f i nd i ng s : 

(1) La Salle 1 Cycle 7 

During this inspection, the team reviewed the RLQR for Commonwealth 

Edison Company (CEC), La Salle County Nuclear Vwer Station Unit 1 

(La Salle 1) Cycle 7. The RLQR contained the signatures of the Fuel 

Project Manager, the Licensing Engineer, and the Engineering Review 

Team, indicating their acceptance of the reasonable accuracy and 

- 21 - 

103 
adequacy of the reload licensing analysis. The Engineering Review Team 

had identified six open items related to details of the analyses 

performed. Within 7 days of the RLQR meeting, the responsible engineer 

had resolved the concerns that had been raised to the satisfaction of 

the Engineering Review Team. 

La Salle 1 is one of four CEC plants for which CEC is currently 

performing all or some of the fuel management, safety and licensing, and 

plant support services. 

performing the balance of these services as well as fuel rod assembly, 

design, and fabrication services. 

fuel management, safety and 1 icensing and plant support are currently 

being transferred from GENE to CEC in a phased manner, distinct for each 

plant, over several fuel cycles. 

The team found that GENE is currently 

However, the responsibilities for 

The team noted that the input engineering data sources (such as the 

final release design basis WA, licensing NAY FRED, and OPL-3) were 

ordinarily sufficient to define the scope of the fuel management and 

safety and licensing work at NEP. In this case, however, the complex, 

phased transfer of responsibi;i:ies from GENE to CEC has necessitated 

the development of a design interfacing procedure (DIP). CEC and NEP 

jointly developed the DIP in conpiiance with a specific fuel contract 

provision, which required establishment of interfacing procedures 

required by the purchaser’s assumption of fuel management services. 

team’s review determined that the DIP provides guidelines for all 

nuclear design-related technical interactions between CEC and GENE 

during the transfer of responsibilities from GENE to CEC. 

The 

In accordance with the split scope responsibility, the licensing WA only 

specified the calculation of stability, emergency core cooling system (ECCS), LOCA, and MAPLHGRs by GENE for La Salle 1 Cycle 7 core 

operation. 

performed to meet the WA specifications, including responsibilities for 

primary and backup analyses. 

The DIP specified the detailed scope of the analyses to be 

As part of this inspection, the team reviewed the Nuclear Fuel 

Independent Design Verification Check1 ist, the Reload Transient Ana7ysis Design Verification Guide, the Reload Licensing Transient Analysis 

Selection Review Summary, LOCA Analysis Results, Stability Analysis 

Results, and the RLQR Report for La Salle 1 Cycle 7. 

this review, the team found that each document had the requisite 

signatures of the responsible engineer, verifier, and review members 

(when appropriate). 

licensing analysis, the DRFs were much less extensive than for a fullscope reload licensing analysis. 

results of all analyses required by the DIP for La Salle 1 Cycle 7 were 

documented. 

required for reload licensing of La Salle 1 Cycle 7, in accordance with 

existing procedures and guides. 

served its intended function of clearly specifyiqg the division of 

responsibilities for primary and backup analyses between GENE and CEC. 

On the basis of 

Because of GENE’S 1 imited scope in this reload 

None the less, the team found that the 

The inspection indicated NEP had performed the analyses 

The team also found that the DIP had 

- 22 - 

104 
The team concluded that the development of a comprehensive and workable 

DIP, in cooperation with the licensee, to respond to a unique situation 

(e.g. , phased transfer of fuel management and 1 icensing analysis 

responsibilities) was a strength of NEP. 

(2) Hatch 1 Cycle 16 

The team found that NEP conducted the Hatch 1 Cycle 16 reload evaluation 

using the methodology described in Amendment 22 of GESTAR 11. The Cycle 

16 reload consisted of four GE13 Lead Use Assemblies (LUAs), 180 fresh 

GESB bundles, and 376 irradiated fuel bundles. To provide improved 

operating flexibility and cycle extension, NEP performed an expanded 

operating domain analyses for increased core flow (ICF) and final 

feedwater temperature reduction (FFWTR). 

The team determined that the Cycle 16 reload was a split-scope analysis. 

That is, the licensee performed the initial fuel bundle design, fuel 

cycle analysis, and application fuel cycle analyses, while NEP performed 

and verifi?,d the fical fuel bundle design, 1 icensing, core 

management , and core monitwint) analyses. Each of these analyses 

were performed and documented by a different responsible engineer. 

Each of the NEP analyses is documented in a separate DRF, which details 

the formal record of the engineering analysis supporting the reload 

evaluation. The team reviewed each of the Cycle 16 DRFs; however, the 

primary focus was on the Reload Licensing DRF (311-02346) and Core 

Monitoring DRF (511-02415). In addition, since the Reload Licensing DRF 

consisted of 19 volumes, only portions were reviewed in detail 

(specifically, Volume 1, "OPL-3/FRED"; Volume 6, "PANACEA Collapse"; 

Volume 7, "ODYN Transient Calculation"; and Volume 8, "GETAB ACPR 

Calculation"). 

The team's review of DRF 311-02346, Volumes 1 and 6-8, indicated that 

the core and operating conditions were as specified in the EWA, and that 

the domains, options, and exposure were consistent with the 

specification. In addition, the review indicated that the data 

transfers to the BWREDB were documented and verified. The review of DRF 

311-02415, Volumes 1 and 2, indicated that the Cycle 16 core monitoring 

data generation was performed using the extensive 3D-Monicore QA 

procedure. 

the team identified a concern that the specified 3D-MONICORE/PANACEA 

accuracy acceptance criteria (i.e., standard deviations) did not appear 

as stringent as assumed in the SLMCPR determination. 

However, during the review of this verification procedure, 

In reviewing the Reload Licensing DKF (311-02346), the team noted that 

the scheduled DRF closure date had expired and an extension had not been 

obtained as required by EOP 42-10.00. 

the team noted that, for several volumes of DRFs 511-02346 and DRF 

During the review of the DRFs, 

- 23 - 

The licensee performed certain sections of the core management 11 

anal y s i s . 

105 
311-02415, the table of contents was missing or the forms required by 

EOP 42-10.00 were not used. 

the DRF documentation did not comply with EOP 42-10.00. As a result, 

the team identified a potential nonconformance during this part of the 

inspection. 

The team concluded that these omissions in 

In response to the team's determinat!m: that the documentation of DRF 

511-02346 did not conform to EOP 42-10.00, NEP explained that the DRF 

had not yet been completed. 

would be documented as required by EOP 42-10.00, and that the 

responsible engineer would be further trained on the documentation 

requirements of EOP 42-10.00. 

taken by NEP, the team determined that its concern regarding compliance 

with EOP 42-10.00 had been satisfied, and the potential nonconformance 

was closed. 

In addition, NEP explained that these items 

As a result of the corrective actions 

During the review of Volume 1 of DRF 311-02346, the team noted that the 

FRED document provided safety-related input to the reload 1 icensing 

safety analysis. EOP 42-6.00, "Independent Design Verification," dated 

March 27, 1995, required that this input data be verified before use in 

the licensing analysis. 

not covered by an established verification process. 

this concern to the NEP staff, the team was provided with a GENE 

internal Corrective Action Request (CAR) 95-3 (CAR No. 8), which 

identified the same concern. The team concluded that the lack of 

verification of the reload licensing FRED data was a weakness in the NEP 

verification process. 

However, certain FRED data provided by NEP is 

After expressing 

The team's review determined that selection of the 1 imiting transients 

for a particular reload depends on the plant configuration, selected 

margin-reduction and operating flexibility options, and the reload core 

design. NEP has recognized the importance and complexity of this step 

in the reload analysis, and has established the Transient Selection 

Review Committee to review this selection process. The team considered 

the formation of the Transient Selection Review Committee a strength in 

the NEP approach to reload licensing. The team noted, however, that the 

reload process does not require that the Transient Selection Review 

Committee document the basis of its selection of limiting transients. 

The team also noted the recent loss of several highly experienced senior 

engineers (in the transfer from San Jose) and the commensurate loss in 

corporate knowledge. Consequently, the team concluded that it was a 

weakness not to require that the Transient Selection Review Committee 

document the basis of its findings in order to allow new staff members 

to receive training in the transient selection in the context of an 

actual licensing analysis. 

- 24 - 

106 
(3) Limerick 1 Cycle 6 

NEP documented the reload licensing for Limerick 1 Cycle 6 in DRF 

511-02141. Bundle design, core management, and core monitoring were 

documented in additional DRFs. Split engineering responsibility exists 

for Limerick 1, and the division of responsibility between NEP and the 

licensee was reviewed by the team with the TPM and Fuel Project Manager. 

The team determined that in general , the 1 icensee has responsibil ity for 

the bundle design, reload loading pattern, and core management, while 

NEP has responsibility for the reload licensing analysis and the review 

of 1 icensee activities. 

The team discussed the Limerick 1 Cycle 6 core design and plant 

operations with NEP to determine the changes from Cycle 5 and the 

potential reload 1 icensing imp1 ications. In addition, the team 

inspected the reload licensing process for Limerick by reviewing the DRF 

relative to the requirements of TDP-0023, "Reload Licensing Analysis 

Procedure,'' Revision 0, G.N. Marrotte, dated April 1994, and the 

supporting TDPs. 

interfaces for the initial steps of the reload licensing process (WAY 

product & performance specification, FRED, and the OPL-3). 

The team's review included the key documents and 

The team found that all steps in the reload licensing process were 

documented in the DRF. In addition, the analysis methods, rational, and 

results for various analyses were discussed with the TPMs. 

selected two items for indepth review and discussion. 

involved the cold shutdown margin and corresponding eigenvalue selection 

as discussed in Sections 3.4.2.2 and 3.4.2.1, respectively. 

item was the reload transient analysis for the turbine trip without 

bypass. The team reviewed this analysis relative to the requirements in 

TDP-0839, "Automated Transient Analysis," Revision 0, D.C. Sere11 , dated 

April 1995, and TDP-0042, "Turbine Trip," Revision 0, F.T. Bolger, dated 

April 1995. The team found that the documentation and traceability were 

of a good quality. The team also reviewed the end product of the reload 

process, the SRLR, and found that the report was traceable to the DRF 

and consistent with the RLQR. 

The team 

The first 

The second 

The team concluded that the overall quality of the Limerick 1 Cycle 6 

DRFs was very good, and that a strength of the process was the technical 

discussions in the DRF by the TPM or the responsible engineer. The team 

also found that the Limerick 1 Cycle 6 reload team members were 

experlenced engineers. (The team noted, however, that the reload effort 

was largely completed in GENE'S facility in San Jose, California before 

the move to Wilmington, North Carolina.) 

the high level of automation in the analysis process contributed to the 

overall qual i ty . 

The team also concluded that 

- 25 - 

107 
(4) Pilgrim Cycle 11 

NEP documented the reload licensing for Pilgrim Cycle 11 in DRF 

511-02368. 

511-02335 and DRF 511-02528, respectively, 

DRFs determined that Pilgrim Reload 

bundles of a total core loading of 580 bundles. 

GElO bundles. 

weight percent (w/o) U relative to Cycle 10. 

that other than the pofhial significance of introducing GEll bundles, 

there were no cycle-specific changes from Cycle 10 to Cycle 11 of 

significance to the reload 1 icensing process. 

contained the WAY EUP, FRED, and OPL-3 documents. 

The team began its evaluation with an overview discussion with the 

Cycle 11 TPM and a process manager. This discussion included a review 

of the purpose and use of the EUP, FRED, and OPL-3, and the applicable 

EOPs and TDPs. The Pilgrim Cycle 11 design did not present unusual 

requirements. Except for cold chutdown margin measurements (discussed 

in Section 3.4.2.2(2)), at the time of this inspection, there had been 

no unusual events in the startup and operation of Cycle 11. 

Bundle design and core management were documented in DRF 

The team’s review of these 

Cycle 10 included 140 

(Cycle 11) involved 136 GEIl 

Cycle 11 bundle average enrichment increased by 0.23 

The review also noted 

The reload 1 icensing DRF 

The team determined that MA, FRED, and OPL-3 documents were exchanged 

between NEP and the licensee with NEP signatures per interface 

requirements of the applicable TDP, and verification per the 

requirements of EOP 42-6.00. 

clear as to cycle-specific changes (such as improved scram time, use of 

a power coastdown, and a preliminary design change from 140 to 136 

bundles to meet the €UP requirements). 

also addressed the need for additional testing for Cycle 11 because of 

the introduction of GEll fuel. 

The team concluded that the documents were 

The FRED and OPL-3 documents 

Volumes 3, 5, and 6 of DRF 511-02368 contained the RWE analysis, reload 

transient analysis, and GETAB mini review, respectively. The team 

examined these volumes and found that they complied with the relevant 

TDPs. 

mini-review, comparing the analysis process to the relevant TDPs and 

confirming that the Pilgrim Cycle 11 SRLR contained the XPR data. 

this investigation, the team concluded that the Pilgrim Cycle 11 reload 

transient analysis was satisfactorily completed per EOP 42-6.00 and the 

relevant TDPs. The team also found that the end product results in the 

SRLR were traceable to the DRF. Furthermore, although much of the 

analysis is automated, the DRF volumes contained discussion of the 

analysis results, comparisons to other results, and verification 

discussion. Both the responsible engineers and the verification 

engineers exhibited knowledge of the subject matter. This discussion Of 

results, combined with computer code automation as demonstrated in the 

Pilgrim Cycle 11 reload licensing analysis, is considered a strength for 

the organi zat i on. 

The team primarily used this effort to investigate the GETAB 

From 

- 26 - 

108 
Based on the team's examination of the appropriate EOPs, TDPs, computer code 

documentation, topical reports, and selected DRFs, as well as discussions with 

the engineers involved, the team determined that, with the exception of the 

weaknesses noted, the reload core design, re1 oad safety analysis, and re1 oad 

licensing analysis activities were excellently performed. 

3.4.5 Fuel Assembly Mechanical Design 

TO ovaluate NEP's fuel assembly mechanic&? design process, the team conducted 

an indepth review of the design process with certain emphasis on fuel failures 

mechanisms. The following paragraphs summarize the team's findings. 

The fuel assembly mechanical design process consisted of the design analysis 

and supporting mechanical, seismic, and flow-induced vibration (FIV) testing 

of the fuel assembly components, channels, and control blades. 

aJso includes preparation of the 1 icensing documentation and following the 

design through the manufacturing process. Typically, the fuel assembly 

mechanical design is perfwmed generically as new fuel designs are introduced, 

and the cycie-specific design analysis is minimal. NEP's generic analysis 

typically includes the following assessments: 

The process 

evaluation of fuel rod stress with the FURST code 

seismic spacer tests 

mechanical analysis for transient overpower conditions using GESTR-M 

stress analysis of the various assembly components (e.g., tie plates 

and water rods) using the ANSYS code 

FIV testing for the evaluation of fuel rod fretting 

a finite-element cladding collapse analysis. 

As part of the evaluation of the fuel mechanical design analysis, the team 

selected and reviewed DRF 311-01652 (the GE12 lower tie plate), DRF 511-02363 

(the evaluation of a GElO lower tie plate manufacturing deviation), and DRF 

511-02223 (the Hatch 1 Cycle 16 reload design). 

DRFs included the proper WAS, analysis documentation, and verification, and 

conformed to the requirements of EOP 42-10.00. 

component analyses in these DRFs, and found them to be acceptable. 

The team concluded that these 

The team also reviewed several 

3.4.5.1 Rod Bowing 

The NRC's acceptance of the Amendment 22 of GESTAR I1 rod bowing evaluation 

was based, in part, on the observation that the effect of fuel rod bowing for 

GENE BWR fuel was small, as well as the requirement that the rod bowing will 

be reported when the rod-to-rod gap closure is greater than 50%. In a letter 

from GENE to the NRC, dated January 30, 1995, "Friol Rod Bow in Excess of 50% 

Gap Closure," GENE reported gap closure greater than GO% for two GE6B fuel 

assembl i es . 

- 27 - 

109 
In order to c'iminate the requirement in Amendment 22 of GESTAR I1 to report 

future fuel rod bowing, GENE performed a set of full-scale fuel bundle ATLAS 

tests to determine the CPR effect of bowing the most limiting rod to contact 

with the adjacent rods. GENE documented the results of these tests in %E11 

Critical Power Test with Rod Bow to Contact," NEDE-31829-P, dated April 1990. 

In its letter to the NRC dated January 30, 1995, GENE submitted the results of 

the tests (NEOE-31829-P), which indicatea no CPR margin reduction as a result 

of rod bowing, and requested elimination of the reporting requirement from 

Amendment 22 of GESTAR 11. 

During its review of the tests, the team noted that the ATLAS measurements 

differed from the tests originally proposed to the NRC for assessing rod bow 

(letter from GENE to NRC dated March 14, 1995, "Proposed Rod Bow Test") in the 

following respects: 

The case of 85% gap clcsure was not included; this prevented the 

measurement of a CPR penalty at intermediate gap closures. 

The bowed rod was between spacers 2 and 3, rather than between 

spacers 1 and 2, as proposed. 

Consequently, the observed boiling transition occurred above the PLRs, and the 

team questioned the applicab lity of the measurements to elevations below the 

top of the PLRs. Therefore, the team determined that the correlation of the 

measurements between spacers 2 and 3 to the evaluation of rod bow below the 

top of the PLRs was an open tem, and requested that NEP notify the NRC upon 

developing the correlation. (Open Item 95-01-03) 

3.4.5.2 Fuel Failures 

NEP categorizes fuel failures observed in GENE BWR fuel into the following 

five failure mechanisms (as discussed in the paragraphs that follow): 

(1) crud-induced-local ized corrosion (CILC) 

(2) debris fretting 

(3) undetected manufacturing defects 

(4) pel 7 et /cl add i ng mechan i cal i nteract i on ( PCMI ) 

(5) unknown causes. 

NEP explained that the unknown category is somewhat misleading, because these 

failures most likely result from one of the other four listed causes; however, 

they have not been examined in enough detail to determine which of the four 

caused the failure. NEP reported that for fuel manufactured since 1989, there 

have been 13 failures that have been dispositioned, with 7 resulting from 

debris fretting, 5 from undetected manufacturing defects, and 1 from unknown 

causes. 

- 28 - 

110 
Crud-Induced-Local ized Corrosion 

According to NEP, CILC failures have not been observed in the last 

6 years. 

change from copper condensers and the use of deep bed dimineralizer 

cleanup systems along with better COG; ant chemistry control s. 

NEP believes that these failures have been eliminated with the 

Debris Fretting 

Debris fretting results when foreign iaterial becomes entrapped adjacent 

to the claddi.ng, and subsequent FIV causes the foreign material to wear 

through the cladding. The foreign material is either inadvertently 

introduced by the licensee into the core during a maintenance outage, or 

it is introduced into the fuel assembly during the manufacturing process 

(in the form of metal turnings or other debris created during the 

manufacturing process). NEP has recently implemented several steps to 

eliminate the introduction of debris in the fuel assembly manufacturing 

process. These improvements are discussed in detail in Section 3.6.5.1 

of this report. 

Undetected Manufacturing Defects 

NEP reports that manufacturing defects usually involve end plug welding 

defects or defects in the cladding or end plugs. 

defects NEP has introduced various changes in their manufacturing 

process, including performing 100% inspections of the various fabricated 

components and steps when possible. These improvements are discussed in 

detail in Section 3.6.5.1 of this report. 

To eliminate these 

Pel 1 et/Cl adding Mechanical Interaction 

PCMI failures result from significant changes in the local power level 

of a fuel rod, usually caused by control blade movements adjacent to the 

rod. 

of fission products leading to stress-corrosion assisted cracking of the 

cladding. According to NEP, PCMI failures were significantly reduced in 

the mid-1970s with the introduction of preconditioning interim operating 

management recommendations that reduced the magnitude of power changes 

during short time periods. 

barrier cladding that eliminated PCMI failures until recently. 

The power changes result in fuel pellet expansion and the release 

In the late-l970s, NEP introduced the 

NEP examined a recent fuel failure, and found that it was caused by the 

combination of a chip missing in the fuel pellet surface, and a power 

increase resulting in PCMI failure in a barrier clad rod. As a result 

of this discovery, NEP tightened the acceptance requirements for fuel 

pellet chip sizes in their fuel manufacturing process to eliminate the 

possibility of similar failures in the future. 

In addition, a licensee recently reported three fuel failures in barrier 

fuel. 

for greater than two cycles of operation, and the failures occurred when 

control blades were withdrawn adjacent to these rods. (These are 

Of these, two occurred jn rods with initial low-power operation 

- 29 - 

111 
classic operational symptoms of PCMI failures.) 

failures may have resulted from existing cladding defects assisted by 

the PCMI mechanism, or there may be other reasons for these failures 

(rather than PCMI being the principle cause for failure). 

cause is currently listed as unknown because these rods have not been 

examined . 

NEP indicated that the 

However, the 

As NEP noted, PCMI has nearly been eliminated as a primary failure 

cause, but it can result in extensive secondary damage to barrier clad 

rods and, to a lesser extent, to non-barrier clad rods once a primary 

rod failure occurs as a result of other causes. 

damage results from cladding hydride assisted embri ttlement caused by 

water ingression. If a significant local power increase occurs in this 

failed and embrittled rod, a long longitudinal crack usually results, 

extending from several inches to several feet, and often results in fuel 

loss and very high coolant activities. 

szcondary damage, NEP introduced a solutlon that involves leaving a 

control blade in near the failed rod to keep its power level low until 

it can be discharged from the core. 

significantly reduce the likelihood of secondary failure damage. 

The extensive secondary 

To prevent the extensive 

This has been shown to 

On the basis of its review, the team concluded that NEP has an aggressive 

program to monitor and review fuel performance. However, the team noted that 

post-irradiation examinations of failed fuel were not routinely performed to 

determine actual root cause of the fuel’s failure. 

3.4.6 BWR Engineering Database and Engineering Computer Programs 

To evaluate NEP’s BWR Engineering Database (BWREDB) and engineering computer 

programs (ECPs), the team conducted an indepth review of the data collection 

processes, and the configuration, testing, and verification of the computer 

programs. 

NEP uses the BWREDB as the QA-qualified input and output files for individual 

reload core design and safety analyses. 

specifications, the BWREDB also includes generic design information for each 

fuel design, as well as design and operating conditions for each plant/reactor 

in which NEP designs are applied. The BWREDB was applied in the reload core 

design and safety analyses and in the manufacturing of the fuel bundle. The 

engineering computer programs (ECPs) use the input data and generate output 

data for the reload analyses that demonstrate that the specific reload meets 

safety and design operating limits. The following paragraphs summarize the 

team’s findings. 

In addition to the reload 

3.4.6.1 BUR Engineering Database 

The team evaluated the BWREDB system in two areas of NEP’s operations (reload 

core design and safety analyses, and fuel production). The following 

paragraphs summarize the team’s findings. 

- 30 - 

112 
(1) Reload Core Design and Safety Analyses 

The team determined that once the FRED and OPL-3 data were entered into 

the BWREDB, the reload core design process began. The process used the 

BWREDB (includes input and output data from previous cycles) and the 

ECPs to perform the analyses necessary to demonstrate that safety and 

design limits and EUP requirements met for each reload cycle. 

Before data can be transferred to the BWREDB, the independent design 

verification ar.c! responsible manager who must indicate that all issues 

have been resolved and that the verifier is qualified to perform the 

verification. When the reload design analyses are completed and 

transferred to the BWREDB and the DRF is closed, a bundle design report 

and a fuel management summary report are created from the BWREDB. 

licensee provides an authorization for release and NEP issues a WA to 

fabricate the fuel. 

The 

(2) Fuel Fabrication 

The team determined that the rr;,,iufacturing release activity began with 

advancing inputs to manufacturing, which include enriched uranium and 

zircaloy hardware requirements. The BWREDB system is then used to 

define the fuel design dimensions and material, including enrichment and 

gadolinia (Gd,O,) levels, in each fuel rod and bundle for fabrication. 

The BWREDB provides a paperless system that automatically prepares 

engineering instructions and bundle drawings, as well as digitized information used to drive the Automated Bundle Assembly Machine (ABAM). 

This BWREDB information is automatically transferred to the appropriate 

fabrication areas, where nearly all of the fabrication is machine 

automated (including bundle loading of the fuel rods) so that the data 

does not need to be input. 

The team concluded that The BWREDB is a strength in the NEP fuel design and 

fabrication process, because it helps to reduce the probability of human 

error. The team also noted, however, that as with any automated system, human 

error can be introduced early in the process and will then be propagated 

through the design and fabrication process. Therefore, it is essential to 

ensure that the inputs to the BWREDB are adequately QA qualified and without 

error. 

into the BWREDB have adequate review procedures defined in 06-0002. 

After reviewing the BWREDB, the team determined that the data input 

3.4.6.2 Engineering Computer Programs 

The team’s review of the ECPs determined that once the ECP development is 

completed, validated, and verified, a review team (level 2) performed an 

independent review to determine that specifications have been met. 

2 review team also ensured that the code has been validated against proper 

data or other codes, and that the code has been properly tested, and gives 

3ppropriate calculational results. The ECP is not transferred to level 2 

until the Level 2 review is completed and ECP problems resolved. 

The level 

- 31 - 

113 
I 

NEP explained that completion of the level 2 review does not mean the code is 

approved for production applications for a particular plant and fuel design. 

Approval is not provided until an application design review is completed to 

determine whether the fo11 owing cri teri a have been met : 

Implementation will measurably improve the process. 

Implementation problems have been identified an? resolved. 

The new ECP is adequately controlled to ensure that its 

implementation is consistent across NEP. 

The team noted an exception to the level 2 requirements. These ECPs are 

designated as Level 2R. The NEP level 2R definition in "Engineering Computer 

Programs," EOP 40-3.00, Revision 17, dated March 27, 1995, explicitly states 

that restricted approved production programs are ECPs that do not satisfy all 

requirements for level 2, but may be applied to design tasks for a limited 

time with control component individual (CCI) approval. The team concluded 

that no other requirements are imposed for these ECPs. Therefore, the team 

questioned why EOP 40-3.00 did not include minimum requirements for level 2R 

ECP. The CCI advised the team that, according to EOP 40-3.00, any software 

components of the ECP that do not comply with level 2 requirements and that 

can affect the calculational results should be excluded in a level 2R ECP. 

(That is, an unresolved issue on the independent design verification of a 

software component of the ECP would exclude the use of this software component 

in level 2R applications of the ECP.) Even though the CCI's interpretation of 

the level 2R requirements in EOP 40-3.OC were satisfactory, the team noted 

that other intwpretations of the EOP requirements could be significantly less 

restrictive. Therefore, the team concluded that the minimum requirements for 

level 2R designated ECPs were not adequately prescribed by EOP 40-3.00. 

The team identified concerns described below that level 2 ECPs were used 

outside of the NRC-approved range of applications, and that NEP changed the 

ECPs without notifying the NRC of the changes. The two ECPs examined were the 

GESTRM versions (GESTRM06V, GESTRMO'IV and GESTRXOlV) and the ODYN code 

versions (ODYNM09V, ODYNV09V and ODYNMlOV), as discussed in the following 

paragraphs. 

(1) GESTRU 

The GESTRM ECP versions are codes used to provide best-estimate 

predictions of the thermal -mechanical performance of nuclear fuel rods. 

These codes set the maximum LHGRs versus burnup for different fuel 

designs. The team examined GESTRM ECP versions along with their stated 

range of applications, d~ well as the ECP software coding changes 

between versions. 

range for Gd,03. additions was [deleted pursuant to 10 CFR 2.790 - 

document described a specific value] w/o maximum. 

to 10 CFR 2.790 - document described a specifir value] w/o Gd,O, limit 

allowed by GESTRM was contrary to the NRC agreement with GENE and the 

subsequent NRC approval that GENE would not exceed [deleted pursuant to 

10 CFR 2.790 - document described a specific value] w/o Gd,O, until NEP 

The NEP documentation for GESTRM gave the application 

The [deleted pursuant 

- 32 - 

114 
obtained confirmatory fission gas release data from four segmented rods 

with [deleted pursuant to 10 CFR 2.790 - document described a specific 

value] w/o Gd 0,-UO, (per letter MFN-193-83, dated October 18, 1983, 

from GENE to hRC). Once this confirmatory data demonstrated 

satisfactory results with their ECP predictions, NEP would use up to 

[deleted pursuant to 10 CFR 2.790 - document described a specific value] 

W/O Gd,O,. 

This agreement was implemented because of NRC concerns about fission gas 

release and thermal differences between Gd 0,-UO fuel and UO, fuel. 

When the team questioned NEP about the status of the confirmatory 

fission gas release data from [deleted pursuant to 10 CFR 2.790 - 

document described a specific value] w/o Gd 0,-UO, segmented rods, NEP 

stated that the irradiations are complete, kut the rods have not been 

punctured yet to measure fission gas release. NEP further noted that 

Gd,O, content in the fuel is controlled and limited to [deleted pursuant 

to 10 CFR 2.790 - document described a specific value] w/o for all of 

their current fuel designs, and not by the GESTRM ECP application. 

NEP's DB-0009.02, "Design Basis Document - Standard Pellet List (GEll) ," 

Revision 0, dated May 1995, and DB-Xl10.02, "Design Basis Document - Standard Fuel Rod List (GEll)," Rebizion 2, dated May 1995, were 

reviewed by the team and found to limit Gd?O, to [deleted pursuant to 

10 CFR 2.790 - document described a specific value] w/o for the GEll 

designs. The team therefore considered this issue closed. 

The team's review found that the application range for fuel burnup in 

GESTRM06V was [deleted pursuant to 10 CFR 2.790 - document described a 

speci f i c value J GWd/MTU (peak rod average), whi 1 e in the GESTRM07V 

version it was increased to [deleted pursuant to 10 CFR 2.790 - document 

descri bed a speci f i c val ue] GWd/MTU (peak rod average). NRC NUREG-1503, 

"Final Safety Analysis Report (FSAR) ABWR," Volume 1, documenting NRC's 

approval of the ABWR reactor design, specified a burnup limit of 60 

GWd/MTU (peak rod average), and further stated that any extensions of 

this burnup limit would be submitted to the NRC for review and approval. 

Therefore, the team concluded thdt the application limit of [deleted 

pursuant to 10 CFR 2.790 - document described a specific value] GWd/MTU 

in GESTRM07V exceeds the NRC staff burnup limit of 60 GWd/MTU for BWR 

fuel designs. 

NEP stated that it disagreed with the NRC staff's position on the burnup 

limit off 60 GWd/MTU, and asserted that burnup extension does not 

constitute a safety issue. However, for the reasons stated in 

NUREG-1503, the NRC staff considers fuel burnup extensions a safety 

issue and, therefore, disagrees with NEP's position. 

burnup extension beyond 60 GWd/MTU was not resolved during this 

inspection, and, therefore, the team requested that GENE respond to this 

issue as an open item from this inspection. 

The team's review found that thz GESTRXOlV version was created for use 

by external users with only minor changes to the originally approved 

GESTRMOCi version. 

version was limited by NRC to 60 GWd/MTU rod average. 

The issue of a 

(Open Item 95-01-04) 

The application range for fuel burnup for this 

- 33 - 

115 
(2) ODYN 

The OOYN code versions simulate the dynamic behavior of BWRs, and are 

used by NEP to perform ID pressurization transient analyses. 

ODYNM09V version is approved by the NRC for plants with motor generator 

regulated flow pumps, and the ODYNVVr version is approved for plants 

with control valves on recirculation flow. The teanl determined that a 

new version ODYNMlOV has recently been given Level 2 status, out had not 

received application design review approval by the NRC. The ODYNMlOV 

version includes the mcdel ing for motor generator plants from ODYNM09V, 

as well as the modeling for control valve plants from ODYNVOSV. A third 

model had been added to simulate variable speed flow pump mtors, and 

was being offered to licensees as an option for GENE BWRs. The team 

determined that the variable speed motor simulation required several 

lines of software coding additions to ODYN, and was needed for plants 

that intend to use the new variable speed motors in order to calculate 

the correct timing of the pump trip during pressurization transients. 

The team asked NEP whether tbe CIDYNMlOV version will be submitted to the 

NRC for review because the team concluded that the simulation of the 

variable speed motor was not a trivial modeling task. 

responsible engineer for ODYN versions stated that ODYNMlOV was not to 

be sent to the NRC for review and approval becauss it did not result in 

a significant change to the calculational results for those plants using 

ODYNM09V and ODYNV09V. The team reviewed NEP’s analysis results, and 

confirmed that the changes in the results were indeed small and 

considered this issue closed. 

The 

The NEP ECP 

3.5 Fuel Desicns 

As part of its inspection, the team evaluated NEP’s new fuel designs, GE12 and 

GE13 (as described in NEDE-32417-P and NEDE-32198-P, respectively) for 

compliance with Amendment 22 of GESTAR I1 (NEDE-24011-P-A). The design 

acceptance criteria described in Amendment 22 of GESTAR I1 were approved by 

the NRC staff in a safety evaluation report (SER) dated July 23, 1990. 

Amendment 22 of GESTAR I1 also established a set of fuel licensing acceptance 

criteria for evaluating new fuel designs, and established the critical power 

correlation bases and SLMCPR criteria for new designs. In addition, Amendment 

22 of GESTAR I1 established the applicability of previous GENE generic 

analyses to new fuel designs. 

The NRC staff previously evaluated, and generally approved, the GEll fuel 

design, as documented in an SER dated March 25, 1992, with regard to the 

approved design acceptance criteria in Amendment 22 of GESTAR 11. 

the NRC staff also identified deficiencies in NEP s application of the design 

acceptance criteria to the GEll fuel design. The NRC staff discussed certain 

improvements for future fuel design submittals to NRC. The team’s evaluation 

of the GE12 and GE13 fuel designs was therefore, n part, a followup of the 

previous evaluation of the GEll fuel design. The following paragraphs 

sumnarize the team’s findings. 

However, 

- 34 - 

116 
The GE13 fuel design is essentia ly a variation of the GEl! fuel design, in 

which the GE13 has one more grid spacer (8 spacers in total) and longer partlength rods (PLRs) than the GEll design. The v3riation improves critical 

power performance for GE13. The team considered that GE13 is basically a 

product evolved from GEll, and that there are no fundamental differences 

between these two fuel designs. The team’s review of the GE13 fuel design for 

reload licensing applications resulted ho*pio-*er in two open issues relating to 

MCPR and R-factor approved methodologies (as discussed later in this section). 

Compliance of the GE12 fuel design with Amendment 22 of GESTAR I1 was 

therefore the main focus of this portion of the inspection. The team’s 

evaluation also included a detailed review of the NEP evaluation for 

compliance with each of the Amendment 22 of GESTAR I1 design criteria 

addressed in NEDE-32417-P. The team also reviewed NEP’s corrective actions 

taken to close out the deficiencies identified by the NRC SRXB during its 

audit of the GEll fuel design (as reported in an SER dated March 25, 1992). 

From its review, the team determined that the GE12 fuel has some design 

features that differ significantly from those of GE11/13. 

GE12 fuel design is a 10x10 array, while the GE11/13 fuel designs are in a 9x9 

array. The GE12 fuel consists of 92 fuel rods, 2 central water rods, with 

8 Inconel grid spacers encased in an interactive fuel channel. There are more 

PLRs in GE12 than in GE11/13. The fuel rods can have UO, rods or U0,-Gd20, 

rods with a natural uranium blanket at the top and bottom ends. 

For example, the 

The following topics are organized and numbered to correspond to NEP’s 

documentation of the GE12 (NEDE-32417-P) and GE13 (NEDE-32198-P) evaluations 

for compliance to Amendment 22 of GESTAR 11. The following paragraphs 

sumnarizes the team’s findings: 

3.5.1 Lead Use Assemblies, 2.1(2) 

For GE12 fuel design, NEP has four demonstration lead use assemblies (LUAs) 

currently irradiated in Pennsylvania Power and Light Company, Susquehanna 

Steam Electric Station Unit 2 core. The team noted that NEP plans to have one 

or more domestic reactors to receive GE12 fuel for irradiation program. In 

addition, NEP reported loaded another 12 fuel assemblies of various GE12 

design features in German and Swedish reactors in 1993 and 1994, respectively. 

NEP will use these irradiated fuel assemblies to verify the GE12 fuel design 

features. 

collecting corrosion data (Section 2.3 of NEDE-32417-P) , the team concluded 

that NEP has satisfied the intent of the LUA requirement in the acceptance 

cri ter i a. 

Based on these LUA programs and NEP’s commitment to continue 

3.5.2 Them1 -Mechani cal , 2.2 

(1) Stressg Strain, and Fatigue, 2.2(1) 

The stress, strain, and fatigue criteria for NEP BWR designs have been 

previously reviewed and approved and applied to previous designs (NEDE22148-P-A and NEDE-31917-P), 

remain applicable to the GE12 fuel design. 

NEP maintains +,hat these same criteria 

Of particular concern in the 

- 35 - 

117 
GE12 design is the high burnup levels of [deleted pursuant to 10 CFR 

2.790 - document described a specific value] Gwd/MTU (peak rod averages) 

intended for this design. This concern is due to the accelerated 

corrosion at high burnups (= 60 Gwd/MTU) observed in pressurized-water 

reactor (PWR) fuel designs. 

reduced the strain capability to below the 1% uniform strain limit 

specified in the NRC Standard Review Plan (SRP) Section 4.2, "Fuel 

System Design." 

conservative than the SRP Section 4.2 strain criterion. 

total strain criterion, NEP estimated the effective uniform strain for 

their high burnup cladding to be below the SRP strain criterion. 

In addition, NEP presented both rod-average and maximum corrosion values 

up to a rod-average burnup of [deleted pursuant to 10 CFR 2.790 - 

document described a specific value] Gwd/MTU, demonstrating that the 

maximum corrosion in NEP designs is less than those observed in PWR fuel 

designs. This level of corrosion was found to be acceptable for the 

strain and fatigue criteria used for the GE12 design. 

The high corrosion in PWR cladding has 

NEP uses a total strain criterion of 1% that is more 

Based on the 1% 

NEP has performed cladding stress, strain, and strain fatigue analyses 

for the GE12 design using the GESTR-MECHANICAL code NEDE-24011-P-A-6, 

Amendment 10, and approved methods that include consideration of normal 

operation, AOOs, and uncertainties in operation and fabrication. These 

analyses demonstrate that the GE12 design meets the previously approved 

design criteria for those plant applications defined. Therefore, the 

team concluded that stress, strain, and strain fatigue are acceptable 

for the GE12 design up to a rod average burnup limit by NRC of 60 

Gwd/MTU. 

(2) Fretting, 2.2(2) 

The GE12 fuel has a unit cell spacer design made of heat-treated Inconel 

alloy. All eight low-pressure drop Inconel spacers are identical. The 

GE12 fuel pressure drop is controlled by part length rods, low-pressure 

drop Inconel spacers, a low-pressure drop upper tie plate, and a highpressure drop lower tie plate. 

The team asked NEP whether the GE12 fuel design including new spacers 

will affect assembly rod vibration and fretting wear characteristics. 

NEP responded that the GE12 fuel was tested to ensure that the design 

features do not significantly increase FIV response, and thereby do not 

increase the potential for fretting wear. The GE12 fuel was tested for 

vibration response in a flow device against the P8x8R flow data. The 

results showed that the two vibration responses of GE12 and P8x8R fuel 

designs were very similar. Since the P8x8R fuel presented no 

significant vibration problem in the past experience, GENE concluded 

that the GE12 fuel also should have no vibration problem. 

Another type of fretting wear is caused by foreign material (debris) 

entrapped and vibrating adjacent to the fuel rods. The failure is 

usually located near the vicinity of lower tie plate and lower end 

plugs. The debris fretting is characterized by a smooth abraded area 

- 36 - 

118 
(3) 

with secondary hydriding away from the debris perforation. NEP has a 

new design lower tie plate with smaller flow holes that significantly 

reduce the size of passable debri s. 

Based on the similarity of the two vibration responses of GE12 and P8x8R 

fuel and the new design of the lower Lie plate, the team concluded that 

the GE12 fuel design has minimized the tendency for fretting wear caused 

by flow-induced vibration or debris-induced fretting. Thus, the team 

concluded that the fretting wear design performance is acceptable and 

meets the acceptance criteria. 

Hetall Thinning, 2.2(3) 

NEP does not have a cladding corrosion or crud limit on the cladding 

design, other than the criterion that the effects of cladding corrosion 

and crud buildup are to be included in its thermal and mechanical 

analyses. In the past, the BWR fuel rods have been troubled by a 

part icul ar type of nodul ar corrosion call ed crud-i nduced 1 oca1 i zed 

corrosion (CILC). The CILC failure mechanism was attributed to 

environmental condition, operational history, and material 

susceptibility. 

copper a1 1 oy condenser tubes and f i 1 ter demi neral i zer condensate cleanup 

systems. 

The CILC failuras were usually limited to plants with 

NEP and BWR owners have taken mitigating actions, including replacement 

of copper bearing condensers with titanium or stainless steel, water 

chemistry control , and improved tubing fabrication and testing methods. 

With these improvements, GENE has effectively reduced the fuel failure 

rate associated with CILC. 

NEP has various LUA programs to continue collecting corrosion data for 

high burnup fuel. In general, visual inspection of the LUAs revealed 

excellent corrosion performance along the full length of the fuel rods. 

NEP will continue to irradiate LUAs for the purpose of extended burnup 

regime to confirm acceptable corrosion performance. 

Based on the improvement of plant operation and cladding fabrication to 

eliminate the nodular corrosion, and based on NEP's commitment to 

collect corrosion data during high burnup, the team concluded that the 

GE12 fuel design has adequately analyzed the corrosion performance, and 

thus meets the acceptance criteria. 

(4) Fuel Rod Internal Hydrogen Content, 2.2(4) 

Fuel rod internal hydrogen content is controlled during the manufacture 

of the fuel rod. To limit the maximum amount of hydrogen, NEP has 

speci f i ed standards (C776-83 and C934-85) def i ned by the Ameri can 

Society for Testing and Materials (ASTM). 

the GE12 design includes steps to remove hydroaen and verify that these 

ASTM standards are met. 

prevent internal cladding hydriding has been met. 

The manufacturing process for 

The team concluded that the design criterion to 

- 37 - 

119 
(5) Fuel Rod/Channel Bow, 2.2(5) 

Fuel rod and channel bow are included in the T/H analyses to prevent 

boiling transition, as discussed in Section on 3.5.3 of this report. 

The decrease in rod diameter of the GE12 design compared to earlier 

designs has the potential to increase the effects of rod bow. However, 

to help offset the effect of a decrease in rod diameter, NEP increased 

the number of spacers in the GE12 assembly as compared to the GEll 

design, and decreased the spacer spacing where rod bow is calculated to 

occur. The team concluded, therefore, rod bow is not expected to differ 

significantly for the GE12 design. 

on MCPR will not be discussed in the thermal-mechanical section, but are 

discussed in Section 3.5.3 of this report. 

The effects of rod and channel bow 

(6) Cladding Pressure Loading, 2.2(6) 

NEP has performed rod pressure analysis for the GE12 design using 

conservative upper bound rod powers and analysis methods described in 

NEDE-22148-P-A and NEDE-24011-p-A-6, Amendment 10. The fuel swelling 

and cladding creep equations rmain the same as those evaluated for the 

GEll design NEDE-31917. These equations are used to determine the 

maximum critical pressure at which the cladding creep out rate will not 

exceed the fuel swelling rate (NEP criterion for rod pressure). The 

team examined the GE12 rod pressure analyses, and verified that the fuel 

and Gd,O, rod pressures remain below the critical pressure threshold. 

The team also confirmed that the rod pressure criterion has been met for 

the GE12 design. 

(7) Control Rod Insertion, 2.2(7) 

Appendix A, "Evaluation of Fuel Assembly Structural Response to 

Externally Applied Forces," to SRP Section 4.2 described the 

requirements of control rod insertion during the combined seismic and 

LOCA loadings. For BWRs, the fuel assembly (including fuel rods and 

grid spacers) must maintain its structural integrity and resist fuel 

liftoff from the core plate during the combined seismic and LOCA 

loading. The liftoff of fuel assemblies would interfere with control 

blade insertion because of the fuel assembly lateral movement. To 

ensure GE12 structural integrity, NEP performed a fuel liftoff 

calculation based on the approved methodology described in NEDE-21175-3- 

P-A. 

previous staff evaluation of the GEll fuel design, which recommended 

that NEP select a referenced plant with significant liftoff. The result 

showed that the GE12 fuel liftaff it; the referenced plant is 

substantially below the fuel liftoff limit. Therefore, GE12 fuel 

assemblies remain seated in the core plate during the combined seismic 

and LOCA loading for the referenced plant. 

A referenced plant was chosen for this calculation based on the 

Based on the approved methodology and the xceptable referenced plant, 

the team concluded that NEP has demonstrated that the GE12 fuel conforms 

to the acceptance criteria of no fuel liftoff during the combined 

seismic and LOCA loading. 

- 38 - 

120 
(8) Cladding Collapse, 2.2(8) 

The team questioned NEP with regard to whether their analytical models 

or methods have changed. 

same as that defined in 1985 for GESTR-MECHANICAL, and the methods are 

the same as those approved in NEDE-20606-?-A. 

NEP indicated that the creep model remains the 

The NEP cladding coll apse analysis demonstrated that cl adding coll apse 

will not occur in the GE12 design at the maximum in-reactor times for a 

rod average burnup of 60 GWdIMTU, as required by the GENE design 

criterion. Therefore, the team concluded tnat the GE12 design is 

acceptable with regard to cladding collapse. 

(9) Fuel kiting, 2.2(9) 

The NEP criterion is to prevent fuel melting for normal operation and 

AOOs. 

NRC-approved analysis methods and models that demonstrate that fuel 

center melting will not occur as a result of normal operation or AOOs. 

The team concluded that the GEii fuel design is acceptable with regard 

to fuel melting. 

NEP has performed fuel melting analyses for the GE12 design using 

3.5.3 Safety Limit FICPR, 2.6 

The SLMCPR is influenced by the critical power correlation and by bundle 

design parameters which affect the bundle R-factor distribution and the core 

radial power distribution. These parameters include the spacer design, 

assembly dimensional geometry, enrichment level and distribution, and fuel 

discharge exposure. 

uncertainties in the core monitoring system and the GEXL critical power 

correlation. The determination of the SLMCPR for the GE12 and GE13 fuel 

designs employed the standard 1 imiting statepoint assumptions and analysis 

methods. However, because of the design improvements made in the GE12 and 

GE13 bundle designs, a series of ATLAS tests was performed to determine 

design-specific GEXL correlations for both the GE12 and GE13 fuel design. 

The SLMCPR provides the margin required to account for 

The critical power analyses for the GEll, GE12, and GE13 fuel designs were 

performed with the GEXL-07, GEXL-10 and GEXL-09 CPR correlations, 

respectively . The GEXL-07 and GEXL-09 correl at i ons were defined over a range 

of inlet subcoolings from zero to 70 British thermal units per pound (Btu/lb). 

For the GEXL-10 correlation, the upper limit of the subcooling range has been 

extended from 70 Btu/lb to 100 Btu/lb. 

database does not include data above 60 Btu/lb subcooling. This prevents the 

determination of the correlation statistics (i .e., the mean and standard 

deviation) in the extended range for use in the SLMCPR calculation. The team 

identified this concern to NEP and NEP responded that additional data (which 

is presently archived) can be used to resolve this concern. The team noted 

that this issue should be resolved during the NRC's GE12 and GE13 fuel design 

reviews. 

iesign reviews, the team requested that NEP provide the data used to extend 

the upper limit subcooling range from 70 Btu/lb to 100 Btu/lb, and considered 

this matter an open item. 

However, the GEXL-10 critical power 

Therefore, to resolve this concern relative to the GE12 and GE13 

- 39 - 

121 
NEP responded to the team's concerns in its letter JFK-95-073, "GEXL10 

Subcooling Data Extrapolation," dated October 17, 1995. 

GENE in response to the team's concern is currently under evaluation by SRXB 

and will be addressed separately by that branch. 

The data submitted by 

In order to properly account for the axial dependence of the fuel rod power in 

the case of PLRs, NEP changed the definition of the R-factor. The R-factor 

change applies to the GEll, GE12, and GE13 PLRs fuel designs, as well as the 

corresponding GEXL07, GEXL10, and GEXLO9 correlations. For previous (fulllength rod) fuel designs the R-faccor is defined using an axial integral of 

the "rod power." For the GEXL07, GEXL09, and GEXLlO correlations, the 

R-factor is defined using an axial integral of the "local R-factor." Although 

Amendment 22 to GESTAR I1 requires that changes in R-factor definition be 

reported to the NRC, this change was not reported. The team concluded that 

this indicated a weakness in NEP's adherence to the Amendment 22 to GESTAR-I1 

reportitlg requirements. 

NEP responded to the team's concerns in its letters JFK-95-092, "GEXL09 

Auditing Information," dated October 20, 1995; JFK-95-093, "R-Factor 

Calculation Method," dated October 20, 1995; and JFK-95-113, "GE13 MCPR Safety 

Limit," dated November 30, 1995. 

the team's concern is currently under evaluatior! by SRXB and will be addressed 

separately by that branch. 

The data submitted by GENE in response to 

3.5.4 Stability Licensing Acceptance Criteria, 2.9 

Comparison With Previously Approved Designs, 2.9( 1) 

During its audit of the GEll fuel design12, the team noted that the 

stability analyses procedures used by GE had been demonstrated to be 

non-conservative, and should be replaced by the BWROG stability analysis 

procedures that were then under review. NRC appryjal of the BWROG 

procedures was subsequently documented in the SER dated July 12, 

1993. The team was concerned with the acceptability of the GE12 

stability design because the smaller diameter fuel rods tend to reduce 

the margin to instability, and require compensating design measures to 

maintain acceptable stability decay ratio. 

In its stabil i ty compl iance evaluation, NEP compared the GE12 decay 

ratio to values for the previously approved R8x8R design. 

described to the team were performed for an appropriate range of 

operating conditions, and included both the regional and core-wide modes 

of instability. The results indicated that the GE12 fuel provides 

additional margin to the onset of regional oscillations. 

GE12 fuel shows a slightly greater propensity for core-wide instability 

than cores operating with the previously approved P8x8R fuel design. 

The analyses 

However, the 

'*documented in SER dated March 25, 1992 

13NEDO-31960 and Supplement 1, dated July 12, 1993 

- 40 - 

122 
The team concluded, therefore, that the GE12 fuel design does not comply 

with the first stability design acceptance criterion (i.e., 2.9(1), 

"Comparison With Previously Approved Designs") of Amendment 22 of 

GESTAR 11. 

(2) Exclusion Zone Evaluation, 2.9(2) 

The second stability acceptance criterion of Amendment 22 of GESTAR I1 

permits a new fuel design to be more limiting in core and channel decay 

ratios if it can be demonstrated that there is no change to the 

calculated exclusion zone to ensure stable operating conditions. GENE 

performed sensitivity studies to compare the calculated boundaries for 

stable behavior with the comparative fuel designs. The comparison 

indicated mixed results, with the GE12 stabil i ty boundary producing a 

slightly more restrictive operating condition for some regions of the 

power/flow map and a less restrictive condition for others. 

concluded that the differences between the two designs were very small 

and produce negligible differences (1% - 2% power) in the exclusion 

zone. The team also found the differences too small to be significant, 

and that the GE12 stability design 2ppears to be acceptable with regard 

to this criterion. 

GENE 

The team determined, however, that the design does not strictly satisfy 

the condition of "no change" to the stability boundary expressed by this 

Criterion. Therefore, in order to avoid a non-conservative creep in the 

exclusion boundary for successive fuel designs in accordance with 

Amendment 22 to GESTAR 11, the team concluded that the GE12 fuel design 

not be uszd as a reference design for future comparisons with Amendment 

22 of GESTAR 11. 

3.5.5 Refueling Accident, 2.13 

The NEP criterion is that the radiological consequences of this accident must 

either bound previous analyses or perform a new analysis that demonstrate that 

the consequences are within the countFy-specific limits on radiological dose. 

In a simple analysis in NEDE-32417P, NEP demonstrated that the radiological 

consequences of a GE12 bundle refueling accident are less than those for the 

previous 7x7 and 8x8 array fuel designs for plants equipped with a standard 

triangle refueling mast. 

However, for the application of the GE12 design to plants with the heavier 

NF500 cy1 indrical mast, the NEP analyses in NEDE-32417-P predict higher 

radiological consequences than the previous result for the 8x8 array fuel 

design. This is because there is an increase in the number of failed rods 

GE12 fuel, as compared to the 8x8 fuel design. Therefore, plants with the 

NF500 mast and an FSAR refueling accident analysis based on the 8x8 fuel 

design will require a re-analysis of this accident for GE12 fuel application. 

On the other hand, the application of GE12 fuel design to plants with the 

NF500 mast predicts less radiological consequences than the previous result 

for the 7x7 fuel design. 

analysis based on the 7x7 fuel design do not need a re-analysis of this 

accident for GE12 fuel application. 

Thus, plants with the NF500 mast and an FSAR 

- 41 - 

123 
The team concluded that the radiological consequences of the refueling 

accident of the GE12 application to plants with a triangular refueling mast 

are acceptable; however, a re-analysis will be required for plants with both 

the NF500 mast and an FSAR result based on the 8x8 fuel design for GE12 fuel 

appl i cati on. 

3.6 Nucl ear Fuel Fabr i cat i on 

To evaluate NEP's nuclear fuel and core component fabrication activities, the 

team evaluated NEP's activities in the fuel manufacturing operations (FMO) 

building, the fuel component operations (FCO) building, and the service 

components operations (SCO) building. This part of the team's inspection of 

NEP emphasized the manufacturing processes that relate to fuel rod failure 

mechanisms (e.g. ? hydriding, fretting , pel 1 et/cl addi ng mechanical i nteract ion 

(PCMI) , overheating, cladding collapse, bursting, and mechanical fracturing). 

The following paragraphs summarize the team's findings. 

3.6.1 Procurement 

The team found that NEP distinguishea the procurement of materials and 

products delivered to their customers from materials and products used in 

their own manufacturing operations. 

eventually shipped to their customers was referred to as "direct material 

sourcing," while procurement of materials 2nd products used in the 

manufacturing operations was referred to as "indirect material sourcing." 

Procurement of materials and products 

Vendor approval act i vi t i es were governed by "Vendor Approval and Survey , 'I PAP 

60-11, Revision 14, dated April 1, 1994, that described the methods for 

conducting vendor surveys and approving vendors for various types of 

procurement (e.9. , the American Society of Mechanical Engineers (ASME) Code 

materials, safety-related materials and services, and radioactive material 

shipping containers). All production materials and products were procured 

from vendors that were on the approved vendors 1 ist (AVL) . 

Production procurement activity was initjated by generating a purchase 

requisition. Before placing of the purchase order (PO), the requisition had 

to be reviewed and approved by the purchased material quality control (PMQC) and purchasing organizations. The team reviewed at least one PO for each of 

the following product forms: 

barrier tube-reduced extrusion (TREX) 

nonbarrier TREX 

endplug bar 

channel strip 

upper and lower tie plates 

The team's review of POs for the fuel channel strip procurement will be 

discussed here. 

(Zr) alloy (zircaloy) channel strip to NEP in accordance with GENE 

specification 23A7239, "Thermal Size Anneal Channel ,I' Revision 1, and NEP's QC 

plans for zirconium and zircaloy. 

channel strip was produced from double-width material that was slit into two 

The team noted that three suppliers had provided zirconium 

In accordance with the specification, 

- 42 - 

124 
pieces and shipped to NEP as a matched pair. 

with matched pairs of strip and identified with a heat, anneal, and strip 

number when received from the supplier. 

Each channel was manufactured I 

A fuel channel is fit over the fuel bundle to direct in-reactor coolant flow. 

Made of zircaloy, each channel is approxim,alely 14 feet in length. Newer fuel 

designs use a fuel channel that is interactive with the bundle itself, and is 

intended to enhance critical power performance. 

channel and channel shell as a safety-related basic component (as described in 

Section 3.6.2 of this report). 

NEP's requirements and the POs reviewed during this review required the 

channel strip suppliers to maintain a quality control system and provide or 

have available lot samples in a manner similar, although not in accordance 

with, the quality requirements of Appendix B to 10 CFR Part 50. In addition, 

NEP also imposed its own standards and requirements. One of these suppliers, 

Western Zirconium, supplied fuel channel strips to NEP specifications in 

accordance with QC plans A-208, "Zirconium and Zircaloy," as well as A-232, 

"General Documentation Requirements for Material, Services or Products." From 

its review of NEP's POs to Western Zircoci?rm, the team determined that NEP did 

not procure the channel strip as a basic componellt (as defined in 10 CFR Part 

21). The team concluded that NEP's procurement of the channel strip did not 

meet the definition in 10 CFR Part 21.3 for a commercial grade item because, 

in part, it was subject to NEP design and specification requirements that are 

unique to NRC licensees. 

NEP classified the fuel 

With the exception of the fuel channel fastener (discussed in Section 3.6.2 of 

this report) 

during this inspection did not impose the reporting requirements of 10 CFR 

Part 21 on its suppliers of materials, items, or services that NEP supplied to 

the nuclear industry as safety-related basic components. During the team's 

discussions with NEP regarding its procurement practices, NEP stated that it 

did not dedicate (as defined in 10 CFR Part 21) the materials, items or 

services. 

for basic components may not comply with the requirements of 10 CFR Part 21. 

NEP's procurement practices are further discussed in Section 3.6.2. 

the team determined that NEP's procurement activities reviewed 

Consequently, the team determined that NEP's procurement practices 

The team advised NEP that the requirements of 10 CFR Part 21 were being 

revised (affective October 19, 1995) and that it was expected that NEP will 

review the information for applicability to their procurement practices 

(specifically, the expanded definition for commercial grade items) and 

consider corrective actions, as appropriate. 

On the basis of the issues identified above, the team raised concerns 

regarding NEP's safety classification of the fuel assembly and its component 

parts, as discussed below. 

- 43 - 

125 
3.6.2 Safety Classification 

The team requested GENE to explain its safety classification of the fuel 

assembly and its component parts. In response to the team's concerns, NEP 

presented GENE'S Position Statement entitled, "Safety Classification of GE 

Fuel Assembly and Related Components," dated November 17, 1994. 

statement noted that, for the purpose of this position statement, the term 

"basic component" as defined in 10 CFR Part 21.3(a)(I) is equivalent to the 

term "safety-related" used by NEP for the ourpose of Functional 

classification. 

structures, systems, components or parts thereof (as used in 10 CFR Part 21) 

was given in EOP 65-2.10, "Safety-Related Classification," Revision 8, dated 

March 27, 1995. 

The position 

The GENE procedure for the functional classification of 

Paragraph 2.6, "Classification Determination," of EOP 65-2.10 stated, in pF[t, 

that the safety-related classification starts with the design-basis events (DBEs) and the functions required to prevent or mitigate these events, and 

extends to the system, component, and eventually to the part level. The 

paragraph continued by stating that the determination of a specific 

classification to be assigned to a system, component, or part was divided into 

two categories, functional and procurement. Additionally, the paragraph 

specified that the procurement category was then established either as safetyrelated or commercial grade, and concluded by adding that a commercial grade 

item was subject to dedication before it was used as a safety-related item (in 

accordance with EOP 65-2.20, "Dedication of Commercial-Grade Items," 

Revision 5, dated July 27, 1994). 

On the basis of its review of paragraph 2.6 and the POs reviewed and described 

above, the team concluded that NEP's procurement and safety classification 

practices for certain "direct material sourcing'( did not comply with the 

requirements of paragraph 2.6 of EOP 65-2.10 because NEP did not procure these 

materi a1 s as either safety-re1 ated basic components or commerci a1 grade i tems. 

For the same reasons stated in Section 3.6.1 of this report, the team 

identified this weakness in NEP's procedural compliance with paragraph 2.6 of 

EOP 65-2.10 as a non-cited nonconformance. 

According to GENE'S Position Statement dated November 17, 1994, a fuel 

assembly consists of only three safety-related components and gives the basis 

for that classification, as follows: 

Design basis events are defined in EOP 65-2.10 as conditions of normal 

operation, including Anticipated Operational Occurrences (AOOs) , design basis 

accidents, external events, and natural phenomena for which the plant must be 

designed to ensure three functions described in 10 CFR Part 21.3(a)(I), i.e;, 

(1) the integrity of the reactor coolant pressure boundary, (2) the capability 

to shut down the reactor and maintain it in a safo shutdown condition, or (3) 

the capability to prevent or mitigate the consequences of accidents which 

could result in potential off-site exposures comparable to those referred to 

in 10 CFR Part 100.11. 

14 

- 44 - 

126 
(1) Fuel Channel - The fuel channel and channel shell provide a guiding 

surface for control rod insertion and direct coolant flow within the fuel 

bundle during DBEs. GENE concluded that the fuel channel and channel shell 

help to ensure functions 2 and 3 described in footnote 14 and are therefore 

both safety-related. 

(2) Fuel Channel Fastener - The fuel channel fastener, through the 

channel fastener bolt, connects the fuel channel to the fuel bundle, thereby 

ensuring sufficient space for proper insertion of the control rods and proper 

flow paths for coolant flow to the fuel bundle during and following DBEs. 

Therefore, GENE concluded that the fuel channel fastener and channel fastener 

bolt support functions 2 and 3 described in footnote 14 and are therefore both 

safety-re1 ated. 

Note that the channel fastener bolt screws into a post that is one of four 

integral-cast posts of the upper tie plate which is not considered by NEP to 

be safety-related. 

(3) Fuel Bundle - The fuel bundle configuration provides reactivity 

characteristics (i.e., Gd O3 neutron :Ssorber and enriched UO, distribution) 

which ensure the capability to shutdown the reactor and maintain it in a safe 

shutdown condition during and following DBEs. 

retain it general mechanical and geometric integrities to maintain a coolable 

geometry during and following DBEs. 

therefore, helps to ensure functions 2 and 3 described in footnote 14 and is 

therefore safety-related. 

The fuel bundle must also 

GENE concluded that the fuel bundle, 

However, the Position Statement adds that GENE does not classify as safetyrelated fuel clad tubing, end plugs, upper tie plates, lower tie plates, 

spacers, or mi scel 1 aneous hardware typi f i ed by finger springs , coi 1 spri ngs , screws, nuts, lock tab washers, etc., used in fuel bundles (i.e., none of the 

fuel bundle's constituent parts are safety-related). 

The team therefore determined that GENE claimed that no bundle part 

contributed to the fuel bundle's ability to perform its safety function as 

described in (3) above. 

supports a functional evaluation of each part as follows: 

However, paragraph 2.6.l(e) of EOP 65-2.10 further 

"It will often be the case that a safety-related component contains 

items which are both safety-related and nonsafety-re7ated. This is 

strictly due to the function of the items within the component. 

item provides support for or is integral to the performance of the 

safety-related function of the consponent, or its fai7ure can prevent the 

component's satisfactory performance of the safety-related function, the 

item must be considered safety-related. 

re7ated. 'I 

If the 

If not, the item is nonsafety- 

- 45 - 

127 
However, contrary to the functional evaluation of each part prescribed in 

paragraph 2.6.1 (e) of EOP 65-2.10, GENE classified all of the fuel bundle’s 

constituent parts as nonsafety-re1 ated on the basis that these component 

parts, in and of themselves, could not contain defects which would credibly 

create a substantial safety hazard as set forth in 10 CFR 21.3. 

determined that this position was not based on a functional evaluation of the 

parts . 

The team held several discussions with NEP staff about the issues relating to 

its nonsafety-related classification of the constituent parts of the fuel 

bundle. The team then concluded that NEP had not established, during the 

course of this inspection, an acceptable basis for its position that none of 

the fuel bundle’s parts support or are integral to the performance of the 

safety-related function of the fuel bundle. NEP asserted that, on the basis 

of the safety-related function of the channeled fuel bundle in the unique BWR 

environment, an acceptable basis for its safety classification position could 

be developed. Therefore, the team agreed to provide GENE the opportunity to 

Submit such an evaluation for the NRC staff’s review. The team therefore 

considered this matter an open item and requested that NEP develop a 

functional evaluation for the fuel bundle parts, as prescribed in paragraph 

2.6.l(e) of EOP 65-2.10. 

3.6.3 Chemical and Ceramic Operations 

The team 

(Open Item 95-01-05) 

The team reviewed the chemical and ceramic operations from the receipt of 

uranium hexafluoride (UF,) through the production of sintered and ground 

pellets. All 

processes are housed in the FMO building. 

as they were performed. Items checked included the presence of written 

procedures at work stations, the calibration of instruments and gages, 

sampling points, analytical equipment, and analytical procedures. Interfaces 

with QC and QA oversight were reviewed. 

the team’s findings. 

The recycle of scrap material was included in this review. 

The team inspected all operations 

The following paragraphs summarize 

3.6.3.1 Chemical Conversion 

The UF, is converted to ammonium diuranate (ADU) by a conventional wet 

process. 

with water forming a uranyl fluoride solution. This solution, together with 

ammonium hydroxide solution, form a precipitate of ADU. The ADU in paste form 

is heated to where it decomposes thermally releasing NH, and then is reduced 

to UO,. The UO, is loaded into cans which have a bar-coded traveler card. 

The cans weight is automatically entered into the Material Inventory Computer 

System (MICS). 

card and also into the In-process Material Release System (IMRS) computer. 

The powder is transported to the hammermill in the mill-slug-granulate (MSG) 

area. 

The UF, cylinders are heated to vaporize the UF,, where it reacts 

The operator also enters the weight of the can on the traveler 

There are four lines for the MSG production of U!2z powfer. When the cans of 

UO, powder enter the MSG area, the transaction is entered into the MICS. The 

cans of UO powder are dumped into a milling machine and the resulting milled 

powder is aischarged to a slugging machine. The slugging machine produces a 

- 46 - 

128 
pressed compact which is pushed into a granulating machine. The granulated 

product is collected in cans which are weighed and receive new bar-coded 

traveler cards. Data on weights and composition are entered into the IMRS. 

The identity and weight of the cans are entered into the MICS system and the 

cans are placed in storage for later blending operations. 

Samples of the granulated powder are sent to the Chemet Lab for analyses 

(i .e., isotopics, metal impurities, carbon, chloride, fluoride, nitrogen, 

moisture, and O/U ratio). 

cards are attached to the sample containers. The results of analysis are 

entered into the Laboratory Material Contrcl System (LMCS) . 

Bar-coded sample labels received with the traveler 

The team concluded that the process controls and sampl ing process were 

defined and control 1 ed and ensured expected-qual i ty UF, powder. 

3.6.3.2 Uranium Recycle 

we1 1 

A fifth line of powder production originates in the uranium recycle un 

scrap material was divided' into two categories, dirty and clean scrap. 

scrap contained gadolinia (Gd,O,) and/or sLrap which did not meet the 

specifications for other impurities. 

per million (ppm) of Gd 0, and meets specification for other impurities. 

resulting powders are MSG processed, sampled and weighed, and then are 

available for the blending operation. 

Clean scrap contained less than 1 part 

The 

-8 . - * . . 1.1- .I ..-. I. .. . .. .. ine team touna tne ~3ti powaer was iaentitiaDie tnrougn tne entries in tne 

three computer systems (MICS, IMRS, and LMCS). Written procedures were found 

at workstations. Calibration records were checked for several of the 

instruments and were performed on schedule. 

analytical equipment and analytical procedures for the MSG powders were 

satisfactory. 

3.6.3.3 Powder Blending 

The team found that powder blending foilowing, the MSG operation, was 

performed in both the UO, shop and the Gd shop to produce press feed powder 

having uniform physical and chemical properties. 

The team concluded that the 

__- -- 

t where 

Dirty 

UO, lending 

The team's review determined that the blend may be all of a single 

enrichment that was blended solely to provide uniform pressabil ity 

characteristics and chemical properties; or a blend may include two or 

more materials of different enrichments. However, the maximum 

difference between enrichments can be no greater than 1.4% U,,,. 

The team observed the production control operator check the IMRS for the 

results of the MSG powder analysis. If IMRS releases the powder, it can 

be used for blending. The operator specified the identity of cans to be 

used for the blend. Blended powder was placed in cans that were weighed 

- 47 - 

129 
and given new card travelers. 

the enrichment were entered into the MICS system and then the cans were 

placed in storage awaiting pellet pressing operations. 

The identity and weight of the cans and 

The team found that blend samples were pressed in a sample press. 

green density was measured and the result entered in the IMRS. The 

samples were sent to the production area where they were sintered and 

ground in a production grinder. The sintered and ground pellets were 

examined for defects and sent to the laboratory for analysis. 

results of analysis were entered into the Chemet Lab LMCS. 

results were satisfactory the powder was released by the IMRS to pellet 

press. 

The team observed that if any sample analysis results were out of 

specification, the IMRS system printout instructed the shift production 

advisor regarding resampling, as described in the quality control 

inspection instructions (QCII). The IMRS system automatically released 

the blend if resampling results met specifications. The team concluded 

that the blending operations observed were satifactory. 

The 

The 

If the 

(2) 6d,0, Blending 

The team noted that there was a separate line for fuel rods which 

contain Gd 0 . UO enterei the Gd shop in containers specified for either low enriched 

(1$.013% UZ5) or high enriched (>4.013% U235) powder. 

parameter sieet identified for the operator the batches of blended UO, 

powder to be used, their weights, the w/o Gd 0, desired in the completed 

blend, the pellet press pressure, the green $ensity, and the sintering 

temperature for the batch. 

Pre-blended U02 powder was blended with Gd,O powder. 

A process 

The team observed one sample taken at random from one can of blended 

powder for fluoride analysis, and one from each of five randomly 

selected cans for Gd determination. 

Gd analysis on the samples. 

Final assessment of Gd content was performed on finished Gd-containing 

fuel rods using the magnetic and passive scanners (MAPS) system. 

Completed blends were placed in controlled storage until released for 

pressing by IMRS. 

observed were sat i factory . 

The vibromill operator performed a 

This is a process control measurement. 

The team concluded that the blending operations 

3.6.3.4 Pelleting 

The team observed that the IMRS system released the material for pellet 

pressing. 

system and dumped the powder into the hopper for the pellet press. 

A sinter test was performed on a sample taken from every 20th can of calciner 

powder, and every 20th can of MSG powder. A sinter tzst was also made on a 

sample from each powder blend. Sinter test pellets were evaluated based on 

density, cracks, surface pits, etc. Results were factored into the selection 

of specific material used in each powder blend. 

The blend operator recorded the transfer of product on the MICS 

- 48 - 

130 
(1) UO, Pelleting 

The team observed the UO, .pelleting on three rotary presses. The press 

operator monitored the weight and green density and made adjustments as 

necessary to maintain the green density within specifications. The 

pellet punch had a three digit code on the dish-forming surface which 

formed identifying marks on each pellet for enrichment and project 

i dent i ty . 

After a pellet was ejected from the die, it was guided into a molybdenum 

furnace boat and then transferred to the sintering furnace area. 

traveler card was clenerated for each boat at the Dressins station that 

A 

followed it through sintering, storage, and release for grinding by the 

IMRS. 

The team noted that the GENE fuel designs required several different 

enrichments. At the time of this inspection, 13 different enrichments 

were being processed at the same time. As there were only three rotzry 

presses, to change enrichment:, 't is necessary to frequently clean out 

the pressing system. The team determined that the frequent enrichment 

changes in the pressing operations introduced the possibility of crosscontamination of enrichments due to hold-up and release of pockets of 

powder in the equipment. This condition can not easily be detected 

during processing, but it did show up during active gamma scanning in 

the "Fat Albert" scanner. The phenomenon was referred to as "fuel 

spiking" (as discussed in Section 3.6.6.2 of this report). 

(2) 6d20,-UO, Pel ieting 

The team noted that the press feed for Gd,O,-UO, pellet pressing has 

poor-flowabil ity unsuitable for rotary press feed, therefore, special 

presses were used in the Gd shop for pelleting. The press tonnage used 

for pressing was taken from the process parameter sheet which was based 

on sinter test results for the U02 blend used in Gd 0,-UO blending. 

Pellets were marked with an identifying 3-digit codze as iescribed for 

UO, pellets. The team concluded that powder was supplied to the process 

as described previously for UO, pellet pressing. 

3.6.3.5 Sintering 

The sintering furnace contained six independently control led heating zones. 

The team determined that a uniform temperature exists over almost all of the 

high heat zones. The operator sampled five pellets from each boat for density 

determination, using gamma source densitometer and associated 1 aser 

micrometer. Verification standards were run following calibration, every four 

hours during densitometer operation, and immediately before and after any 

sintering test runs. 

high density, the low density, and the deviation. 

into the MICS system. The team observed a measurement of 10 samples; all 

samples were within specification. 

The results give the average density of 5 pellets, the 

The results were entered 

49 - 
Boats of pellets meeting density specifications were sampled for enrichment, 

hydrogen, densification, metal impurities, various chemical elements, total 

rare earths, total impurities, boron equivalent, and one archive sample is 

taken per every 15 blends. Pellets meeting specifications for the foregoing 

analyses were released by IMRS for grinding. 

From its review, the team determined that Gd,O,-UO pellet sintering was 

similar to UO sintering, with few exceptions. Pe\lets were approximately the 

same as for Ub pellets with some minor exceptions. As with UO, pellets, 

Gd2y-U0, pellets were released for grinding by the IMRS based on the analysis 

resu ts. 

this inspection were satisfactory. 

The team concluded that the sintering operations observed during 

3.6.3.6 Grinding 

Pellets were dry-ground using a centerless grinder. 

at each grinding station. 

while the other operator transferred the ground pellets to trays. 

There were two operators 

One operator monitored operation of the grinder 

The team noted that the pellet grinding operation was a "quzlity at the 

source" (QATS) process. 

Grinder operators received formal and on-the-job training. The training 

included operation of the machine, use of micrometers and other measuring 

devices and visual inspection of pellets. 

The operators also performed the first QC inspection. 

The team observed that as each pellet emerges from the grinder, a laser 

micrometer measured the diameter of each pellet several times. The inspecting 

operator visually evaluated the pellets and rejected pellets with excessive 

pits, chips, or cracks into a scrap container. 

located at the inspection station for this purpose. 

A set of visual standards was 

The team determined that, although every pellet was measured by the laser 

mike, there was no provision for automatically scrapping pellets that were out 

of diameter specifications. The QATS operator checked the diameter of one 

pellet from every other tray row using a hand held micrometer. 

observed that while checking diameter, the operator also checked for out-ofroundness. If a reject occurred, several adjacent pellets were checked to 

determine if there was a trend. 

The team 

The results of operator inspections were entered into the IMRS. The team 

found that the inspector put a hold on a tray (Blue Tag), this action was 

entered via a MICS transaction. 

grinder and observed that the tray was satisfactory for operator inspection 

and over-inspection. 

The team reviewed the records for a day shift 

The team determined that only pellet inspection results based on ground 

pellets used by the IMRS for releasing pellets for rod loading were surface 

finish and open porosity. 

Open porosity is determined on one pellet per five Pg2 powder blends and one 

pellet per each Gd,O,-UO, blend. 

on analyses of unground pel 1 ets. 

Surface finish was measured on one pellet per tray. 

All other pellet release criteria were based 

- 50 - 

132 
Because pellets are ground dry, they have visually observable grinding dust on 

their surfaces, and being ground, they have significantly greater surface 

area. The team determined that both factors may contribute to a greater 

amount of absorbed moisture than was present on unground pellets. The team 

concluded that NEP's practice of using the hydrogen analysis from unground 

pellets as a basis for estimating the hydtcyen content of the fuel was not as 

conservative a basis for hydrogen content as using the hydrogen analysis from 

pel?sts after grinding. 

3.6.4 Fuel Clad Tubing 

Fuel clad tubing was manufactured by NEP either with and without a pure Zr 

barrier clad on the inside surface of the tubing, 

either case, the GENE designs incorporated a heat treated outer surface for 

corrosion resistance. The zircaloy tubing was initially received in the form 

of tube-reduced extrusions (TREX). 

prepared for reduction (reduction of both diameter and thickness dimensions). 

The tubes were then saw cut to length, cleaned and annealed. After annealing, 

a sample was sent to the chemet lab for evaluation of the heat treatment. If 

acceptable the tubes were straightened, cleaned again and subjected to an inprocess heat treatment (IPHT) for increased corrosion resistance. 

The team determined that in 

After shop release, the TREXs were 

3.6.4.1 Alpha-Beta Quench 

From its review, the team determined that the alpha-beta quench heat treatment 

process for zircaloy tubing dated back to the early 1970s. 

heat treatment produced a metallurgical structure of intermetal 1 ic 

precipitates which provided good corrosion resistance with no change in 

mechanical properties and characteristics (grain growth). In the 1980s (after 

an extensive corrosion test program by GENE), a solution quench heat treatment 

was developed to assure uniform resistance to nodular corrosion of the tubing. 

This process was accomplished by induction heat treating the tubing while 

simultaneously running cold water through the tube inside surface. 

method of heat treating the tubing resulted in a "skin effect" protection of 

the outside surface to corrosion processes. 

developed for fuel tubing. 

treatment on the tubes but at a later step in the process. 

effect of providing enhanced corrosion resistance with a faster quenching time 

and no change in the final tubes' mechanical properties or texture. 

This particular 

This 

In the late 1980's, the IPHT was 

Thiz process used the solution quench heat 

This had the 

The team observed the IPHT also referred to as the alpha-beta quench. Quality 

Control Inspector Instructions (QCII) No. 15.2.1, "In-Process Heat Treatment," 

Revision 16, June 9, 1995, was used to assure that proper heat treatment 

results had been attained. A sample metallurgical cross section was prepared 

to observe the depth of penetration of the heat affected zone. The thickness 

of the heat affected zone was measured and compared to the specified limits. 

Product heat treatment was accepted when the heat affected zone was within the 

specified limits. Other visual criteria were checked and the product accepted 

for continued processing when the tube shell had pass2d the c r i t e r i a o f 

QCII 15.2.l. 

- 51 - 

133 
I The team determined that manufacturing and qual i ty control operations were 

performed according to the required operation parameter sheets (OPS) and QCIIs 

in the order specified by the shop traveler. 

3.6.5 Fuel Bund1 e Components 

The team determined that, at the time of this inspection, NEP produced fuel 

bundles for six fuel designs, GE8 through GE13. 

for GE8, GE9 and GElO, 9x9 fqr GEll and 6i13, and 10x10 for GE12. All fuel 

rods offered the GENE patented barrier cladding option and all fuel designs 

offered axial gadolinia zones. 

enrichment zones, the large central water rod and ferrule spacers. Designs 

GEll, GE12 and GE13 offered an anti-debris lower tie plate and part-length 

rods. 

reviewed by the team. 

The rod matrix design was 8x8 

All fuel designs except GE8 offered axial 

NEP's manufacturing of the following fuel bundle components was 

3.6.5.1 Upper and Lower Tie Plates 

The team reviewed manufacturing and inspection operations for upper and lower 

tie plates, including debris-filter lower tie plates. 

lower tie plates were machined from a single stainless steel investment 

casting. The debris-filter lower tie plates were machined from two separate 

investment castings welded together. Manufacturing and inspection operations 

were specified on shop travelers for all tie plates. The team noted that two 

QCII's reviewed, QCII 501, "Upper Tie Plate (GE9)," Revision 5, February 10, 

1994, and QCII 501, "Lower Tie Plate (GE9)," Revision 4, August 28, 1992, had 

the same QCII number which had the potential for creating some confusion. 

Both the upper and 

The team observed that activities affecting quality were being performed in 

accordance with written instructions and suitable tests and verifications had 

been satisfactorily completed for activities affecting quality. 

(1) Lower Tie Plate Debris 

The team evaluated NEP's response to a recent issue regarding 

manufacturing debris in lower tie plates. 

sumarize the team's findings. 

The following paragraphs 

During an audit of NEP by Carolina Power and Light Company (CP&L) in 

January 1995, its auditor discovered a piece of metallic debris lodged 

in a GElO lower tie plate between a tie rod and an adjacent peripheral 

rod boss. The debris was = l/2-inch long and 1/8-inch wide. The team 

noted that if such debris was allowed to remain in the lower tie plate, 

it can potentially lead to a failure of fuel rods in reactor operations 

due to a phenomenon known as debris fretting. 

After an additional piece (= 1/8-inch long and 1/8-inch wide) of debris 

was identified that had not been detected by NEP quality inspectors, NEP 

instituted an augmented inspection program it-, 00th SCO, where the tie 

plates were manufactured, and in FMO, where the tie plates were 

assembled into the completed fuel bundle. The augmented inspection 

consisted of an improved visual inspection and a 100% mechanical sweep 

- 52 - 

134 
of the peripheral locations in which the debris had been discovered. 

NEP also formed an investigation team to determine the source of the 

entrapped debris and to assure that the necessary steps were taken to 

prevent recurrence. 

In addition, NEP decided to reinspetl ,he GElO lower tie plates for 

CP&L's Brunswick Steam Electric Plant Unit 1 reload which had already 

been assembled and which were awaiting shipment in shipping containers 

at NEP (Quality Notice F-P-1299, "Reinspection of Bundles for Lower Tie 

Plate Debris, Revision 0, dated January 18, 1995). 

the assembled lower tie plates resulted in the detection of one 

additional piece of debris (= IllO-inch long and 1/5-inch wide). 

This reinspection of 

The team reviewed NEP's response to these findings and found that, 

according to NEP, the difficulty with which the debris was observed 

indicated that inspection personnel needed a greater awareness of the 

condition to observe it. As a result of its investigation, NEP 

performed debris awareness training of inspection personnel, improved 

lighting to aid in debris observation, and revised procedures and 

training documents to reflect these changes. 

manufacturing activities to prevent debris from being caught in the tie 

plate. 

the number of burrs generated. 

established. Hollow cutting tools providing coolant flushing action 

were incorporated in addition the standard external coolant flush as 

well as high pressure washing of the tie plates. 

The team observed that foreign material exclusion zones were established 

and general improvements in housekeeping requirements were instituted in 

the foreign material exclusion zones. Debris generating activities were 

not permitted in the foreign material exclusion zones. Part storage 

areas were removed from proximity to the manufacturing areas and placed 

in the foreign material exclusion zones. 

changed to improve debris shedding abilities and various other 

improvements in material hand1 ing have been incorporated to minimize the 

possibility of debris contamination. 

NEP also changed its 

Cutting tools were retained in a sharper condition to minimize 

Improvements in deburring were 

Wrapping for storage was 

From this review, the team determined that NEP also planned the 

f ol 1 owi ng future improvements : 

The fuel bundle assembly area will become a foreign material 

exclusion zone with 1 imited access by personnel. 

NEP will establish cleaning lines for the exclusive use of tie 

pl ate cleaning operations. 

Air-tight containers will be used for tie plate storage. 

NEP also developed two analyses of the debris problem. 

type problem analysis and decision analysis WE~C performed in February 

1995, and a Tap Root type analysis was performed and documented in "GENE 

Energy Production Root Cause Analysis Report RCA-SCOQA9501, " dated March 

A Kepner Tregoe 

135 

- 53 - 
17, 1995. The team evaluated both reports and found them both to be an 

exhaustive analysis of the debris problem. 

thoroughness in its analysis of the debris problem, the team noted that 

NEP determined that an outside auditor had first identified the lower 

tie plate debris in March 1994. 

time did not react vigorously to the earliest signal of debris in its 

lower tie plates. 

Additionally, on January 20, 1995, iiEP requested a potential safety 

concern (PSC) review in accordance with its 10 CFR Part 21 procedures: 

EOP 65-4.00, "Potentially Reportable Conditions," Revision 2, dated 

October 23, 1992; NEP P&P 30-07, "Reporting Defects and Noncompliance, I' 

Revision 10, dated February 15, 1995; and NEDE-31746, GENE P&P 70-42, 

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

August 1994. On February 6, 1995, GENE concluded that PSC 95-01, 

"Metallic Debris in Fuel Bundle Lower Tie ?late," was not a potentially 

reportable condition (PRC) nor a safety concern for the following 

reasons : 

As evidence of NEP's 

NEP noted that the NEP system at that 

There was no safety limit impact. There was no reason to 

believe the debris would reduce the margin to the MCPR safety limit, nor 

that it would affect the reactor water level or the coolant system 

pressure. 

As the debris was inside the channel, there was no concern with 

interference with control rod motion. Any debris small enough to find 

its way into the bypass region would be too small to pose a problem. 

resulting from debris fretting. 

through the offgas system and poses an operational concern, not a safety 

concern. 

There was a concern for potential localized fuel rod failure 

This would be readily detectable 

Additionally, GENE first called all of its customers to inform them of 

the lower tie plate metallic debris and followed up with at least three 

written communications to its customers in February, March, and April , 1995. The NRC has tracked the lower tie plate debris as a 10 CFR Part 

21 issue. On the basis of the its review of GENE'S corrective actions 

taken, problem analyses, and PSC evaluation, the team concluded that the 

10 CFR Part 21 issue can be closed on the basis that the GENE'S actions 

were adequate. 

Additionally, GENE provided assistance to those 1 icensees that decided 

to inspect the peripheral flow region in the lower tie plates in 

recently delivered fuel that was fabricated prior to implementation of 

the augmented inspection procedures described above. For the inspection 

and removal of lower tie plate debris from unirradiated fuel bundles at 

1 icensee' s sites , GENE i ssued Fuel Examination Techno1 ogy procedure 246- 

GP-18, Revision 0, dated February 1995, 

- 54 - 

I 136 
[deleted pursuant to 10 CFR 2.790 - document described specific values] 

The team’s investigation concluded that NEP management, in conjunction 

with and in full agreement with the licensee, did make a conscious 

decision to ship certain fuel bundles to the licensee without 

reinspecting the bundles at NEP’s facility. However, in these instances 

NEP implemented its augmented inspection of the fuel bundles at the 

licensee’s site, as describe above. NEP’s rational for inspecting the 

fbel at the licensee’s site as opposed to inspecting in the NEP facility 

after the bundles were packaged in shipping containers awaiting shipment 

was to avoid possible damage in NEP’s facility by utilizing the 

1 icensee’s new fuel inspection equipment for up-ending and inspecting fuel bundles. 

The team concluded that with the exception of not reacting vigorously to 

the first indication of a debris problem in lower tie plates that 

occurred in March 1994, the actions taken by GENE to the January 1995 

Brunswick event were very proactive and beneficial to other licensees 

concerning the problem. 

3.6.5.2 Spacers 

The team observed the fabrication of GE9 spacers. 

performed under the control of an computer record-keeping system referred to 

as the Traceway System. Documentation regarding all the components was 

entered into the system to provide traceability from the spacer assembly to 

the raw material. 

accommodated welding and was loaded into a computer controlled 5-axis 

positioning system and laser welder. 

identification was performed in the same setup. 

Weld penetration samples were taken as required and submitted to the Chemet 

Laboratory to be evaluated for weld penetration of one joint of each of the 

following types: 

to band. Evaluation of the weld penetration information was completed and 

determined to be satisfactory before the iot represented by the sample was 

released for assembly into fuel bundles. 

assemblies were constructed and inspected in accordance with approved 

drawings, procedures and specifications using materials that met design 

requ i rement s . 

Spacer fabrication was 

All the components were assembled into a fixture that 

Laser marking for assembly 

cell to cell, cell to band, cell to support plate and band 

The team confirmed that spacer 

3.6.5.3 Fuel Channel s 

Fuel channel strips were procured from NEP’s suppliers in matched pairs with 

significant refinement and machining prior to the strip being welded into a 

fuel channel box. NEP assumed credit for supplier provided machined channel 

- 55 - 

137 
strip devices in the safety analyses. 

procurement and safety classification described in Section 3.6.2 of this 

report, the team did not identify any issues of concern during the fabrication 

or inspection processes for fuel channels. 

Other than the issue of fuel channel 

3.6.5.4 Control Rods 

GENE control rods are of two designs, DuraLifeN and MarathonN blades. Both 

designs were manufactured from high purity stainless steels and the designs 

had been approved by the NRC. The DuraLife" blades utilize stainless steel 

tubes encased in a stainless steel wrapper which protects the individual tubes 

and provides stiffness to the design. The Marathon" design reduces both the 

volume and surface area of stainless steel in the irradiation zone. The 

design utilizes square tubing which is laser welded. 

current (ET) inspected before welding. These tubes are not wrapped in a 

stainless shell, but are open to the reactor environment. 

The tubes are 100% eddy 

While observing the start of mechanical measurement of a thrust limiter, the 

team inquired into the availability of the dye penetrant procedure which had 

been performed on the part. The tedrlc nlas informed that the procedure could be 

at another location, and was not available at the work station. Further 

discussion with the employee disclosed that the dye penetrant examination had 

been performed from memory. The team concluded that this lack of procedures 

in the work area was contrary to Section 5, "Instructions, Procedures, and 

Drawings," of the QA topical report. 

From its review, the team made a second observation from this incident that 

concerned a shop traveler which had two different inspection steps in 

sequence. Since the traveler was unsigned and unstamped for both of the 

inspections (the dye penetrant examination had been previously performed) it 

was not possible for the team to determine the status of the inspections 

performed. 

work station. The team concluded that this practice was also contrary to 

Section 5 of the QA topical. 

evidence of completion of the operaticin and certification of the acceptance 

status of the item(s) inspected. 

The QCII for the first operation was also not available at the 

The team found that this traveler also served as 

In response to the team's observations, NEP management instructed floor 

personnel (the same afternoon as the observation) on the importance of having 

procedures available for use at the work station. These instructions were 

documented (by signature attendance) and were shown to the team. The team 

noted that the attention to corrective action coupled with no additional areas 

noted without procedures (during the inspection) demonstrated prompt and 

adequate short term response by NEP management. 

addressing the culture of having procedures available for use at the work 

station-will be addressed in the training plan improvements in Section 3.8.2 

of this report. 

The larger issue of 

- 56 - 
3.5.5.5 Finger Springs 

During the inspection, NEP identified a problem in a shipment of finger 

springs. Of the 503 finger springs received, 9 - 11 of the finger springs may 

have oversized thickness. NEP stated that the spring’s function is to prevent 

bypass flow; the loss of a spring leaf will allow some bypass flow. 

determined that a broken leaf cannot travel to an area where it could cause 

problems as either a loose part or a contributor to unanticipated flow 

restriction. A1 though the discussions with NEP disclosed that the oversize 

finger springs did not present a safety issue but could cause a degradation in 

fuel bundle performance, the team monitored NEP’s response to this event. 

NEP 

NEP initiated corrective actions to perform 100% inspection of the received 

503 finger springs, including those installed in completed fuel bundles. In 

addition, NEP initiated Supplier Corrective Action Request W3102, dated August 

15, 1995. NEP’s supplier initiated corrective actions to review all previous 

coils of material supplied by their sub-contractory to review the continuous 

thickness chart of the master coil in addition to the material supplier charts. 

The team concluded that the actions taken by NEP were adequate and prudent for 

the complexity and severity of the oversize finger spring material. 

3.6.6 Fuel Rod Fabrication 

From its review, the team determined that the fuel rod fabrication process 

starts with the welding of the lower end plug. 

welded into the fuel rods using a butt weld between the end plug and tubing. 

The butt weld presented a smooth surface for the examination of the weld using 

an ultrasonic (UT) examination. 

read and related to the number stamped on the lower end plug. The results of 

the UT inspection and the identity of the fuel tube were entered into the fuel 

business system computer. 

The lower end plugs were 

During the UT inspection, the bar code was 

(1) UO, Fuel Rods 

The team observed the operator of the rod loading machine draw from 

storage a specified batch of empty fuel rods with lower end cap welded. 

The operator also drew trays of released pellets from storage to meet 

the loading pattern for the tubes. 

linked to the MICS as well as reads the bar code and provided approval 

to proceed. 

The loading machine was computer 

The team also observed the operator load pellets from the tray up to the 

push rod. The push rod checked the stack length; the stack was weighed 

and the weight and length were entered into the computer. When all 

zones were loaded and weighed, the entire stack of pellets was inserted 

into the tube. As the fuel was pushed into the tube, a laser micrometer 

measured the diameter on each pellet, The weight measurements for each 

enrichment were reported as the net weight in fuel certification 

reports. 

- 57 - 

139 
The team determined that the upper end plug weld was made with bead 

welders or flush butt welders. As each tube enters the welding chamber, 

the bar code was read by the machine and the rod type must compare with 

the specification for rods being welded, otherwise the welding operation 

would not be allowed to proceed. 

(2) 6d20,-U0, Fuel Rods 

The team found that the operation of the gadolinia line was similar to 

the UO, line. 

3.6.6.1 Rod Inspection 

The team determined that in the case of bead welded end plugs, a sample was 

taken for X-ray inspection. 

each weld was subjected to a UT examination. 

each completed rod while rotating. vi sua1 standards. 

In the case of butt (flush) welded end plugs, There was a visual inspection of 

Each weld must be acceptable based on 

A sample of the rods, including the I irst rod welded, was checked for end plug 

parallelism. 

including the first acceptable rod, and all reworked rods were segregated and 

tagged for X-ray analysis. 

analysis of rod atmosphere. 

corrosion test and for a daily visual examination and weld parameter printout. Individual rods were not subjected to a helium test. 

performed on the completed fuel bundles. 

In the case of bead welded end plugs, a sample of the rods, 

A sample of production rods was taken weekly for 

Sample welds were made for a weekly 72 hour 

A helium test was 

3.6.6.2 Rod Scanner 

Each UO, rod was sent through an active scanner used to detect enrichment 

blending. 

CFR 2.790 - document described a specific value] percent probability of 

detecting a single pellet that was greater than [deleted pursuant to 10 CFR 

2.790 - document described a specific value]% difference in relative 

enrichment than the adjacent pellets. 

map which was compared to the specification trace. 

within limits, the rod was accepted and the computer sent its acceptance to 

the MICS computer system. 

The scanner, nicknamed “Fat Albert,” had a [deleted pursuant to 10 

The detector for enrichment provided a 

If the comparison was 

The team reviewed NEP’s report of a 4.2% enriched, 0.376-inch diameter pellet 

found in a zone of 3.95% enriched 0.411-inch diameter pellets in a fuel rod 

that was down-loaded for other reasons. According to NEP, the active scanner 

failed to detect this error because the difference in diameters compensated 

for the difference in enrichments. 

inspection (described in Section 3.6.6.1 of this report) was added to prevent 

a recurrence. NEP concluded that there was no definitive explanation for the 

presence of the rogue pellet; although it must have occurred somewhere between 

pellet grinding and fuel loading. 

Since that event, the laser micrometer 

- 58 - 

140 
The team noted that if the laser micrometer inspection had been a part of the 

rod loading process when the defective rod was loaded, the loading machine 

would have stopped pushing pellets into the rod when the misplaced pellet was 

encountered. 

occurring again was small, as long as the laser micrometer inspection was 

continued during rod loading. 

The team concluded that the probability of this type of defect 

From its review, the team determined that the Gd,O, bearing rods cannot use a 

neutron source for scanning, therefore, a magnetic and passive scanner (MAPS) 

was used. The results obtained from Fat Albert and MAPS systems was used as 

certification for the content of U235 in each rod. 

3.6.6.3 Storage and Re1 ease 

Fuel rods that have successfully passed the Fat Albert and MAPS scanners were 

placed in storage. 

for bundle assembly. 

established to control the release of rods to bundle assembly. 

These rods were then available for accumulation of rods 

The team verified that adequate measures were 

3.6.7 Fuel Bundle Assembly 

From its review, the team determined that production travelers for fuel 

assembly operations were avail able on computer terminals at the workstations 

and acknowledgment of the completion of each production and inspection 

operation was also made at the computer terminals. The team observed the 

assembly of a GE9 fuel bundle. Skeleton assembly began with the assembly 

personnel drawing from storage the appropriate number of spacers and a center 

water rod, all of which had been released for assembly. Identifying numbers 

on these components were entered into the computer system which confirmed 

their released status. When all spacers were properly positioned, the 

skeleton was clamped into the automated bundle assembly machine (ABAM). The 

lower tie plate was also properly oriented and clamped into position in the 

ABAM. This completed the skeleton assembly process. 

ABAM examined the preestablished bundle matrix pattern and loaded the rods 

into the bundle. 

the preestablished bundle matrix pattern caused ABAM to stop until the 

discrepancy was resolved. 

plate was placed in position and lock tab washers were installed. 

Any discrepancy between rod identification information and 

After all the rods were installed, the upper tie 

3.6.8 Fuel Certification 

The team reviewed NEP’s basis for reportipg the enrichment of fuel delivered 

to licensees. NEP sends three reports to licensees, as follows: 

Final Certification Bundle Verification for Bundle - provided a 

detailed report for each fuel bundle. 

the bundle and the drawing numbers for all components as well as a fuel bundle 

matrix which identified the rod type and number Cor each location in the 

mati-ix. This report also listed the composition of each fuel rod in the 

It referenced the drawing number for 

- 59 - 

141 
bundle by zone; include the zone number, the design enrichment of the zone, 

the load date, the actual enrichment of the zone, the net weight of the zone, 

the uranium weight in the zone, and the UZs5 weight in the zone. The uranium 

weight and U,, weight are determined from measurements by the active or 

passive scanner devices (Fat Albert or MAPS). 

gives the net weight, uranium weight, and U,, weight for each enrichment in 

the bundle. 

a1 so reported. 

Product Q u a l i t y C e r t i f i c a t i o n Document - provided an itemized list of 

fuel bundles in the shipment, the product configuration for each bundle model 

number, and the number of bundles of this configuration in the shipment. 

Bundle Shipping Document - listed each bundle in the shipment and 

The totals for each bundle and the total for the shipment are 

3.6.9 laboratories 

The team evaluated NEP's Chemet Laboratory and the Gage Laboratory. 

following paragraphs summarize the team's findings. 

The 

3.6.9.1 Chemet Laboratory 

The team found the Chemet Laboratory an integral part of the in-process 

release of raw materials, process control, and final release of finished 

products. 

laboratory, and spectrographic/uranium recycle laboratory. 

The Chemet Laboratory consisted of a metallurgical 7 aboratory, wet 

Hetall urgical Laboratory 

During the team's review of lab examination techniques, using some weld 

samples chosen at random from a Marathon" control rod final package for 

Philadelphia Electric Company, Peach Bottom Atomic Station Unit 3 (Peach 

Bottom 3) (WA HE3958058), the team determined that daily weld samples 

(DWSs) weld penetration examinations were performed at the lab without 

procedures. While verifying the metallurgical examination of weld 

samples from Drawing 107E61226002, Lot 76MRO for welds on the Marathon" 

control rod, the team determined that the weld penetration examination 

was performed in accordance with a memorandum and not an approved 

procedure. (The memorandum, dated April 7, 1992, was a request for the 

lab to issue these instructions as a procedure.) The applicability of 

the memorandum was directed to the following Marathon" DWSs: 

weld-9 annulus seal (Sl,S2) weld-12 square tube to end plug (Al,Bl,AZ,B2) 

weld-14 square tube to end plug (D,E,F) 

As a result of its investigation into the DWS evaluations in response to 

the team's concerns, NEP determined that other Marathon" design 

weldments were also evaluated for weld pene?; ation without benefit of 

procedures (e.g., weld-5 tube to tie rod, and welds-6 and -7 tube to 

tube). 

- 60 - 

142 
Therefore, the team concluded that since there was no other procedure 

for the determination of weld penetration on the Marathon" 

configuration, either the lab personnel have been using this nonapproved procedure or they have been evaluating this manufacturing 

parameter with no procedure. for the Marathon"' design this issue became 

a concern for the team because there was no NDE performed on these welds 

and the destructive evaluations of the DWSs was the only criteria with 

which to judge the quality of the laser welder. 

The team raised the additional issues of how and who evaluates 

metallurgical tests since many requests come to the lab with specific 

instructions while others simply have a miscel Janeous weld designation 

assigned to them. The team concluded that the only reason the system 

currently works was that the Jab had a staff comprised of personnel with 

many years cif metal 1 urgical 1 aboratory experience and could have 

resulted in a worse condition if there was a large turnover of 

personnel. 

nonconformance during this part of the inspection. 

As a result, the team identified this issue as a potential 

In response to the team's determinatinn that DWSs weld penetration 

examinations were performed by the lab without procedures as required by 

P&P 30-10, "QA Program - GENE," the following corrective actions were 

either planed or taken by NEP during the inspection. 

Request (CAR) 95-12, dated August 28, 1995, included the following 

recommended actions to correct the procedural oversight: 

Corrective Action 

e Issue Chemical, Metallurgical and Spectral (CM&S) procedure for 

the evaluation of the DWSs for the welds identified above. Provide 

training in usage thereof to lab analysts. Conduct training of lab 

personnel to identify to management any requested analysis for which 

written instruction was not provided. 

The team reviewed CM&S Analytical Method 2.2.1.0, Revision 0, 

"Metallographic Evaluation of Laser Welds," and concluded that it 

adequately covered the appropriate welds to be evaluated and the 

necessary procedural information to inspect for weld penetration as well 

as other weld attributes on the Marathon" blades. 

0 

characterize and document the margin of acceptance criteria. 

margin of acceptance criteria to satisfy Marathon'" production to date. 

Review reported weld penetration results for Marathon" DWSs, 

Document 

The team reviewed Quality Notice E-G-064, Revision 0, dated 

August 31, 1995, "Review of Postulated Vari abi 1 i ty in Marathon"' Laser 

Weld Penetration." NEP reviewed a random sample of 603 weld penetration 

evaluations from the Marathon" production period of 1989 to the present. 

This review indicated that all of the welds could be accepted after 

considering such factors as visual inspection or bonus tolerance. 

According to NEP, the end result of this exercise indicated that the 

- 61 - 

143 
design requirements provided adequate assurance that any postulated 

variability in the measurement technique for the weld penetration 

results would not affect product safety or perfwmance. The team 

concluded that from this review that NEP’s review provided the necessary 

actions. 

Review lab weld analyses and notify affected parties if there are 

any evaluations performed by the lab analysts without instructions. 

The team reviewed Chemet Lab memorandum CL-95-056, “Marathon” Weld 

Evaluations,” dated August 30, 1995. The memorandum documented that 

training had been provided to the analysts and that three additional 

evaluations were being performed without benefit of procedures. 

included the bar code depth heat-affected zone, water rod weld, and 

channel clip adapter weld. This element of NEP’s corrective actions 

remains open, pending evaluation by NEP to determine whether any safety 

concerns were raised for products which had been evaluated for these 

attributes. 

evaluation was complete. 

These 

The team requested that NEP notify the NRC when the 

(See Open Item 95-01-06 below.) 

0 Review lab activities and cuntrols. Issue instructions to ensure 

that approved written procedures were used to perform any lab analysis. 

Conduct training of personnel on control of written instructions for lab 

analysis. 

This action was to be completed by NEP by October 31, 1995. This 

element of NEP’s corrective actions remains open. 

that NEP notify the NRC when the review was complete. 

(Open Item 95-01-06) 

The team requested 

As a result of its review of CAR 95-12 and the corrective actions taken 

by NEP, the team determined that its concern regarding DWSs weld 

penetration examinations that were performed by the lab without 

procedures had been satisfied and that the potenti a1 nonconformance was 

closed. Additionally, in its letter dated September 11, 1995, NEP 

submitted its documentation for CAR 95-12. 

I (2) Wet Laboratory 

The wet chemical laboratory performs a variety of chemical and physical 

analyses on UO, powder and pellets. 

components are also analyzed in the wet lab. 

samples of all of these materials are catalogued and placed in storage 

through the wet lab. 

The team determined that analyses performed on UO, powder samples 

include isotopic, carbon (C), Chloride (Cl), fluoride (F), nitrogen 

(N,), water (H O), and Oxygen/Uranium ratio (O/U) . Analyses performed 

on pellets incjude O/M (Gd pellets), isotoc’c, C, C1, F, N, hydrogen 

(H,), total outgas, densification, O/U, and porosity. 

observations were made. 

Several fuel rod and assembly 

In addition, archive 

The following 

- 62 - 

144 
A LECO hydrogen analyzer was used for the analysis to measure H, release 

from UO, pellets and zirconium and stainless steel .QN F-Q-2006, 

"Qualification of LECO RH404 Hydrogen Analyzer," Revision 1, dated 

July 30, 1993, documents the basis for using this equipment for hydrogen 

analysis. 

The team determined that there was no direct measurement of the actual 

temperature reached by the sample during measurement. 

temperature was > 1800°C. A power setting of 2600 watts was found to he 

sufficient to melt chromium metal using the heating schedule used for 

UO, pel 1 et analysi s. 

check once each month. Pure chromium melts at 1857°C. 

The specified 

The procedure specified a chrome me1 t temperature 

Upon further investigation, the team determined that a Fisher Chemical 

Company j a r o f chromium granules (stock number C-317) was used for this 

test. There were no certificates or purchasing papers for the chromium 

metal and the same jar of material had been used for several years. 

fisher Chemical Company was contacted by NEP for information regarding 

the purity of the chromium. NEP found that item C-317 was a 

discontinued item that had not been offered for several years and that 

the purity claimed by Fisher Chemical Company for this material was 

typically 99% pure. 

NEP stated that given the 99% purity of the chrome metal, the chrome 

me1 t temperature check, together with the unchanging induction heating 

operating parameters through the years, had given adequate confidence 

that the samples were indeed Deing heated to > 1800°C. 

From the "Handbook of Binary Phasz Diagrams," Update Number 46, December 

1994, by William C. Moffat, Genium Publishing Co., Schenectady, NY., the 

team examined binary phase diagrams for chromium with 38 different 

metallic elements. From this review, the team determined that the 

greatest lowering of the melting point by any of these at low 

concentrations were by platinum, which lowered the melting point to 

= 1700°C at about 3 atom per cent, and lead, which required about 4 atom 

per cent for the same effect. Even though NEP had not verified the 

alleged purity of chromium samples, the team concluded that the chrome 

melt test, along with the record of unchanging power parameters, were 

serving the intended purposes of verifying temperatures reached in the 

H, analysis. 

The team observed the chemical analyst perform a verification of the 

system using a H standard. 

12.93, 4.74 and 1.67 ppm H,. The operating procedure gave the analyst 

the option of rejecting a result for sufficient cause; the 12.93 ppm was 

rejected by the analyst based on "common sense and logic." The team 

requested the analyst to explain how far a value had to be from the 

standard's stated value to be discarded. The analyst stated that any 

number that differed from the accepted value by more than three times 

the f value for the standard would be rejected (as provided for in 

procedure COI 003, "Chemet Laboratory Rerun Out1 ier Criteria, 'I 

Revision 8, dated May 13, 1994). 

The results for four tests were 5.89, 

- 63 - 

145 
The team observed the analyst analyze three of the 1.4 ppm H, standards 

which obtained an average result of 2.42 ppm. This value represented a 

recovery of 172.86% of the stated value. Whenever the instrument was 

calibrated, the results were automatically entered into the LMCS. When 

results fall between the lower and upper zlarm limits, the system is 

considered operating satisfactorily. The team observed that NEP did not 

determine whether these values were statistically significant in 

determining the control of the calibration process for the analyzer. 

The team determined that in this case the value of 172.86% is above the 

alarm limit but within the control limits. The analyst was required to 

make an investigation but it was not clear to the team how extensive an 

investigation was required. 

received a standards results report covering the last 25 calibrations, 

which covered a period of 14 days. 

between the upper alarm limit and the upper control limit, one occurring 

on each of four consecutive days. All of the results for the preceding 

10 days were inside the alarm limits. The team concluded that, based on 

NEP's failure to correct this situation after four days, the 

investigative procedure invoked ;.hen the alarm limit was exceeded was 

not effective. 

During the team's review, the analyst 

Four of the last six results were 

The team determined that the specification maximum limit for H in UO, 

fuel pellets is 1 ppm. Typical results for production materia? is about 

0.1 ppm. In the qualification document (F-Q-2006) the average found for 

15 pellets from the same tray was 0.09 ppm (+ 0.025 ppm). 

weights are six to eight grains. Thus the amount of H, measured is about 

0.7 pg. The "blank" value for empty crucibles is about 0.5 pg, a value 

almost equal to the sample value. The team concluded, therefore, that 

the hydrogen content of UO, fuel pellets was so low that the analytical 

technique was pushed to its lower limit for detecting hydrogen. 

Pellet 

The team concluded that this system was highly automated in handling 

data when results are all nominal, but it left considerable leeway to 

the operator for discarding poor results. 

in some instances there seemed to be a lack of in-depth understanding of 

the equipment operators regarding the equipment they were running and 

the reason they were performing certain steps in the operation. 

The team also observed that 

(3) Spectrographi c/Urani um Recycle Laboratory 

The team observed that the spectrographic lab equipment in operation 

included a grating spectrometer, two ion coupled plasma mass 

spectrometers, and a conventional iilass spectrometer. The grating 

spectrometer was used to determine the concentration of 26 metal1 ic 

impurities in UO powders and pellets, and also analyzed other materials 

used in the plant. 

- 64 - 

146 
With the exceptions noted above, the team concluded that the overall 

procedures and practices in the Chemet Laboratory were well organized and well 

run. Operating personnel were generally well qualified for their positions, 

or were monitored by experienced personnel. Cali brations of instruments used 

for operations and analysis were carried out in a timely fashion. 

3.6.9.2 6age Laboratory 

The ttrdm reviewed the calibration, tracking, and control of measuring and test 

equipment (MTE), such as micrometers and calipers used by QC inspectors and 

workers throughout NEP. The inspectors noted that the NEP gage lab was 

responsible for all of the calibrations and repair of M&TE. 

was staffed by 5 individuals who maintained and updated the QC Gage Control 

System, a computerized central mainframe database that was used to track and 

control M&TE. The database contained an active equipment list of 

approximately 3200 gages and other instruments with an archive file of over 

30,000 M&TE records that have accumulated since the late 1970s. 

system, created over 20 years ago, had been updated several times, although 

the team determined that the system still had only limited sort and no trend 

capability. 

impact on the use of the QC Gage Control System. 

The team determined that there did not exist a user's manual for the 

database, either at the terminal workstations, or at the computer mainframe 

programmer interface, even though this system had been designated by NEP as a 

"QA-controlled software record file." This was a concern to the team because 

NEP procedure 120-13, "Development , Control and Maintenance of NF&CM Computer 

Software," required that user manuals be avail able for QA-control led record 

files unless a valid exemption could be established. The team noted that 

although the lab personnel were intimately familiar with the gage control 

system, only their many years of experience in use of this database has 

enabled them to avoid potential gage control problems and prevent situations 

which could allow out-of-cat ibration equipment to go unnoticed. In response 

to the team's observation, NEP stated that a user's manual was in development 

and would be completed in the near future (as discussed in Section 3.7 of this 

report). 

The NEP gage lab 

This computer 

The team identified the following observations that could have an 

The team observed the lack of procedural control concerning new gages 

that had been calibrated but were found to be out-of-tolerance or in a 

nonconforming condition. A1 though these gages had been tagged as 

nonconforming, they were stored in the same location with new, calibrated 

gages and NEP's existing procedures had not required these gages to be tracked 

or accounted for in a log. 

NEP revised procedure A13, "Entering Gages into the Gage Control System," to 

include additional requirements on disposition of new gages that were found to 

be nonconforming. 

After discussions of this observation with NEP, 

The team found that each of NEP's supervisors (who were also the 

designated "owners" of gages), receicled a weekly status report of gages that 

were past due, delinquent, or lost. 

was based on a 1% delinquent or overdue threshold which was maintained and 

tracked by the Gage Laboratory. M&TE owners who had 1% or more of assigned 

NEP's system for controlling these gages 

- 65 - 

147 
gages overdue or delinquent could not receive additional gages. 

found that during the week of August 21, 1995, 1.7% of 3233 gages were overdue 

or delinquent (approximately 56 gages). 

functional departments, 14 of these supervisors were at or above the 1% 

threshold. In three of the 22 department areas, between 9% - 33% of the gages 

assigned to these supervisors were overdue or del inquent. The team determined 

that, although these numbers change on a weekly basis, various individuals who 

needed additional gages were routinely denied gages at the beginning of each 

work kreek. 

The team 

Of the 46 NEP supervisors within 22 

0' The team observed that, in the past, the many years of experience and 

However, the team concluded that the 

low turnover rate of NEP's Gage Laboratory personnel had minimized problems 

associated with the control of M&TE. 

lack or unavailability of gage trending data could allow an increase in gage 

control problems to occur. 

3.7 Corrective Actions 

The team evaluated NEP's correct i vn action program. 

summarize the team's findings. 

The fol 1 owing paragraphs 

A corrective action request (CAR) form was implemented and tracked to correct 

a problem in the design/analysis process that did not fit in the design change 

requests or in the engineering computer program (ECP) problem report. 

can be initiated from within NEP or from a customer. Once the CAR is 

initiated there was a 30 day time period for a corrective action plan to be 

identified although there was no time limit to implement the corrective 

action, or to close the CAR. 

A CAR 

A ECP Problem Report was generated for a software problem that was categorized 

as either a Type A or Type B problem. 

calculated design value and a Type B as having no impact on calculated design 

values.) If a Type A problem was identified it was up to the ECP responsible 

engineer to report the problem to the ECP users within 30 days of the problem 

by memo (per EOP 40-3.00, "Engineering Computer Programs," issued March 27, 

1995). The memo requests each ECP user to determine the impact of the problem 

on analyses, identification of the specific analyses, and to recommend a 

corrective action. 

to implement the corrective action, notify users that corrective action has 

been implemented and the corrective action taken (software changes implemented). The team reviewed several Type A software problem report to 

determine the types of problems, assessed the adequacy of the corrective 

actions and evaluate the time taken to close out the CARS. 

(Type A was defined as changing the 

It was the responsibility of the ECP responsible engineer 

Two ECP problem reports (94018 and 95004) were greater than 90 days old and 

not closed out. The team observed that neither of these ECP problems were of 

great significance in terms of calculational errors; but concluded that a 

maximum time limit to close previously identified problems would eliminate 

Tvoe A problems from being left open for a year or :anger, which could have a 

significant impact. 

- 66 - 

148 
Type A problem report 95049, issued on July 7, 1995, identifie" in the 

RELSBOEV program was also reviewed by the team and found to be significant 

because it would always list the rotated bundle as having the most limiting 

MCPR when both the mis-located and rotated bufidle dnatyses were performed. 

The team concluded that this problem was significant because the mis-located 

bundle analysis could be more limiting. A? ',he time of the team's exit, NE? 

was continuing its evaluation of this problem report to address the team's 

concerns. 

The Nuclear Fuel Users Forum (NFUF) performed an audit at the NEP facility 

during the week of May 1, 1995. The NEP response to the NFUF findings was 

included in the team's review, as part of its evaluation of NEP's corrective 

action program. 

For CAR 95-003-083-2, omission of a user's manual for the rod load 

pellet diameter laser micrometer computer software, NEP stated that the reason 

for omission was "operator action was not required with the software" and has 

since determined that this statement was incorrect. The corrective action 

taken by NEP included changing the reason why the manual was omitted. 

However, the team observed that NEP did not consider the broader corrective 

action to re-review the "grandfathering and basel !ne" efforts that were 

undertaken in 1992 to bring up-to-date approximately 175 pre-existing 

computer-controlled software and databases, which was the basis for the 

original incorrect statement. 

During the team's review of this CAR, the team identified another 

database where the QC Gage Control Database System user's manual did not 

exist. 

review during the 1992 baseline effort. Although this basel ine review 

identified that no user manual was required, the team questioned this 

rationale because this system was a QA-controlled database listed in the 

software records file; users frequently need to manipulate and change data 

within the database and perform other manipulations in order to obtain 

database printouts. 

NEP's justification for not requiring a user's manual was again its 

As a result of interviews with several of the users and the designated 

QC Gage system engineer, the team concluded that without a users manual 

present, competent operation and use of the database system was not feasible 

without significant on-the-job training by other competent users. NEP agreed 

with the team's conclusion and stated that the QC Gage Control Database manual 

would be completed in September 1995. 

The team noted that the adequacy of the 1992 baseline effort was also 

questioned by a GENE NQA internal audit of the information management systems 

being conducted concurrently during this inspection. 

from this audit, as presented to the team, appear to address the same 

observations of the team. 

Preliminary findings 

For CARS 95-003-OBS-1, and -3, the team concluded that NEP provided 

an acceptable response. However, in CAR 95-003-062-9, NFUF noted that the 

basis for allowing the rod to exceed the Rod Internal Pressure (RIP) guideline 

was not provided in the DRF (as required by EOP 42-10.00). The fuel rod 

- 67 - 

149 
thermal-mechanical evaluation of the GEll BWR2 high Gd fuel rod was given in 

DRF Jll-01663. 

concluded that the response also did not provide an adequate basis for 

accepting the RIP guideline violation. When this concern was brought to the 

attention of NEP, the necessary DRF updates were made immediately and the CAR 

report was reissued. 

concern was acceptabl e. 

After reviewing the NEP response CAR 95-003-OBS-9, the team 

NEP's corrective actions taken in response to the team's 

3.8 Training 

The team evaluated NEP's personnel training program for those activities that 

affect quality in reload core engineering and fuel production. The following 

paragraphs summarize the team's findings. 

3.8.1 Reload Core Engineering 

The team noted that the methodology for core design and licensing at NEP had 

clearly evolved over a number of years, as demonstrated by 77 technical design 

procedures (TDPs) and a siini 1 ar number of engineering operating procedures 

(EOPs). There was a set of 7 separate TDPs that relate to critical eigenvalue 

and reactivity calculations alone indicating the complexity of the core design 

and licensing process. Given this complex design pr-ccess, the team determined 

that training for the design engineers was needed to acquire and maintain 

sui tab1 e proficiency. 

The team concluded that the recent relocation of the reload core engineering 

group from San Jose, California, to Wilmington, North Carolina, resulted in a 

need to hire new staff members to make up for lost personnel as well as 

emphasize the critical role played by training in core design and licensing 

activities. NEP appears to have come to this conclusion as well, and has 

recently issued a TDP on training (TDP-0077, "Training," Revision 0, issued 

August 1995). 

Training at NEP is provided to the responsible engineer within a loosely 

hierarchical matrix that includes the responsible manager, the process 

development leaders (who are assigned to five technology areas, and are 

responsible for continual improvements of the work flow process within these 

areas), mentors (who have demonstrated in-depth know1 edge of a specific 

technology area, and are utilized as training and consultation resources for 

responsible engineers) and verifiers (who have demonstrated sufficient 

proficiency in a specific technology area to allow detailed verification and 

review of the work performed by the responsible engineer in that area). The 

responsible engineer works closely with an assigned mentor till the 

responsible engineer demonstrates proficiency in independent analyses and 

verification. Each stage of proficiency was evaluated and documented by the 

responsible manager in concurrence with a process leader and a mentor, as 

appropriate. In addition to the proficiency demonstration by the responsible 

engineer discussed above, TDP-0077 required development of a process-based 

training plan to be maintained by each process l?ider in his area. This 

- 68 - 

150 
training was required for all responsible engineers on a continuing basis, and 

was to be conducted either in a formal class environment, or as one-on-one 

training. 

technologies and procedures, a review of document or procedural changes, and 

discussion of CAR findings and other reported errors and preventive measures. 

The elements of process training were to include a review of new 

Recently, the reload process has been subdivided into 22 subprocesses, each 

one of which is scheduled to be covered approximately once a year in a 

training course. The team noted that cold shutdown rrtdrgin calculation, an 

area that had recently exDerienced several instances of calculational errors 

and procedural shortcomings, had been covered in a training session in June 

1995. 

In order to evaluate whether the training programs discussed above are 

responsive to the specific needs of the design and licensing efforts at NEP, 

the root cause evaluation of the errors nade in Duane Arnold 1 Cycle 14 

eigenvalue calculations was reviewed by the team as a representative example. 

The root cause evaluation identified two causal factors, and ten associated 

root causes that led to the errors in the calculation. All ten root causes 

identified were human performance difficulty on the part of either the 

technical program engineer (TPE) or Management. 

on the part of the TPE were related to missing knowledge or skill in 

procedures used, lack of training, incorrect use of procedures, incorrect or 

inadequate procedures, and inadequate training or controls provided to 

interpret 1 imitations of analyses used. 

(related to TPE performance) included training in relevant areas and 

modification of procedures. Most importantly, the recommended preventive 

actions also included a management plan to obtain consistency between 

procedural detail and training, an adequacy review of the procedures to ensure 

that they are structured to facilitate understanding and use by the newer 

engineers, and an increased emphasi s on automation options . The recommended 

preventive actions re1 ated to management performance included a focused Fuel 

Engineering training on the verification process and the responsibilities for 

this process. 

The performance difficulties 

Preventive actions recommended 

The level and extent of training available until recently at NEP, viewed in 

the context of the extensive and complex set of procedures used and the recent 

hiring of engineers after the move to Wilmington, North Carolina, was 

considered a weakness by the team. The team noted, however, that the training 

process had recently been formal ized in TDP-0077, and a program of training in 

the reload subprocesses had been initiated. The team also noted that the root 

cause evaluation of the errors in Duane Arnold 1 Cycle 14 eigenvalue 

calculations had identified specific areas for focused training that should 

benefit the newer engineers and enhance the verification process. 

potential benefits from the new training program need to be monitored to 

determine whether the weakness had been removed. 

The 

- 69 - 

151 

I 

I 
3.8.2 Fuel Product i on 

From its review the team found that NEP had 

training program to address common technical 

site-wide training, track and manage records 1 essons 1 earned. 

At the time of this inspection, training was 

for each of the three production areas (FMO, 

acked a formal, integrated 

training needs, plan and schedule 

and feedback audit findings and 

conducted and handled separately 

FCO, and SCO). Managers in these 

areas were administerirg and managing their own training needs without benefit 

Of common training needs or resources to provide the same level and quality of 

training to all NEP activities affecting quality. 

isolationism and resistance to change still existed at some levels within the 

NEP organization and that this had hampered past efforts to focus attention on 

site-wide training issues. 

acknowledge training as a general issue, the team observed that weaknesses in 

administrating and organizing technical training across the site may not have 

been fully recognized by management. 

The team found that 

A1 though NEP management appeared to recently 

For example, the team’s interviews with several managers indicated that the 

highly specific training requirements, such as on-the-job training for many Of 

the fuel manufacturing activities would not have benefited from an integrated 

training program. The team found that many of the highly-experienced workers 

were now area coordinators or supervisors, and were also relied upon to 

conduct and administer training in their areas on an as-needed basis. This 

team effort had assisted management in maintaining the high skill level and 

experience necessary to maintain product qual i ty. The team a1 so concluded 

that for the majority of workers within NEP, a historical low turnover rate 

coupled with an average experience level of 10 to 15 years per worker were 

principle reasons why NEP had been able to minimize any detrimental effects 

from manpower shortages and the lack of an integrated training program. 

However, the team determined that an aggressive training program would 

facilitate NEP’s future ability to maintain product quality. 

During the inspection, the team found several examples in NEP’s QA audit 

findings and other documents where a lack of training was identified as root 

causes and could have potentially impacted product quality if allowed to go 

uncorrected. 

identified several areas where over 40 CARS were written that had root causes 

traceable to inadequate training. 

observed were personnel not following instructions and documentation errors in 

the SCO processes. This audit finding and the subsequent management briefing 

appeared to be the first indication to the team that at some level within NEP, 

training was fully recognized as a potential site-wide issue. 

For example, a 1994 internal QA audit in the SCO building 

Many of the identified deficiencies 

In another example, the team found that, in 1994, two SCO managers identified 

that their staff lacked adequate micrometer training. 

subsequently organized formal training on micrometers and administered this 

training for all individuals who use micrometers in the SCO area. Although 

micrometer training was also a critical issue in the FMO building and FMO 

SCO management 

- 70 - 

152 
managers had experience problems in the past with individuals receiving 

inadequate micrometer training, the team found that FMO managers had nct been 

aware that special micrometer training was planned or administered for workers 

in the SCO building. This example was typical of how training had been 

conducted in the past. 

The team found that training conducted thrc*:;h out the NEP facilities was a 

complex and evolving area that did not have a central focub. The team noted 

that in 1994, several NEP QA auditors, training personnel, and others, 

presented to NEP management the need for improving the training programs 

currently being conducted at NEP. A training program manager was created in 

order to centrally focus and track training for all of NEP. Although these 

actions were only recently initiated, the team concluded that NEP management 

was taking appropriate steps in order to improve the weaknesses identified by 

the team.and the NEP staff. 

3.8.2.1 Quality At The Source 

In 1992, NEP initiated a program whereby operator/worker job tasks and inprocess inspection activities were combined into a single function performed 

by one individual. 

Source," or QATS. The program was in varying stages of implementation across 

the site. A1 though combined production/inspection tasks were not novel to 

fuel fabrication vendors, the team decided to perform an in-depth review of 

the QATS program in order to assess how the program was functioning, since it 

had not been used by NEP before. Since the team was told that QATS was only 

fully implemented in the FMO production facility, the team limited its indepth review to the FMO production facility, although general aspects of the 

NEP QATS program were included in the assessment. 

The team found that overall, the QATS program appeared to be a sound and 

practical approach for NEP to control product quality. However, in several 

instances, weaknesses in QATS training has not enabled NEP to achieve the 

level of effectiveness that was expected from such a quality program. 

Specifically, a lack of training for certain FMO QATS-qualified 

operators/inspectors on how to use inspection gages necessary for individuals 

in inspector-qual ified positions. 

NEP called this program "Implementing Quality At The 

For example, the team observed pellet grinding and inspection. 

observation, a QA inspector identified 5 trays of fuel pellets which contained 

unacceptable amounts of pellet defects and were rejected. 

told the team that it appeared the QATS-operator/inspector apparently failed 

to follow established procedures for the proper use of micrometers. 

found the that these individuals had completed QATS training and had been 

certified to the qualifications stated in the American National Standards 

Institute (ANSI) standard N45.2.6, "Qualifications for Inspection, 

Examination, and Testing Personnel .'I The team determined that this level of 

certification should have provided an acceptable level of performance to 

erlsure product quality was maintained. 

During the 

The QA inspector 

The team 

- 71 - 

153 
The team identified other examples which support these concerns, including for 

example, certain CARs issued by QA that contained the following findings: 

QATS-qual ified operatorslinspectors have left uncertified trainees 

Although final product quality was unaffected, poor product 

unattended at the workstations who were performing pellet grinding/inspection 

activities. 

quality was passed through the first inspection station. 

Unqualified trainee operators have used the QATS-qual ified operator 

identification number in order to sign-on or login-in in order to operate 

computer-control led equipment and some QATS-qual i fied operators condoned this 

practices and had repetitively allowed poor product quality to proceed unidentified, after performing the first quality inspection. 

Root causes identified in some of these CARs indicated a lack of appropriate 

training for new employees. 

causes as failure to follow established procedures, failure to follow 

instructions and properly document work processes. 

This root cause may include such identified 

3.9 10 CFR Part 21 

During this inspection, the team evaluated the GENE procedure and the NEP 

procedures that address the requirements of 10 CFR Part 21. 

evaluation determined that the GENE and NEP procedures met the requirements of 

10 CFR Part 21, the evaluation determined that revisions to the procedures and 

NEP’s procurement practices (as described in Sections 3.6.1 and 3.6.2 of this 

report) will be necessary when the revised 10 CFR Part 21 is effective 

(October 19, 1995). 

While the 

- 72 - 

154 
APPENDIX A 

WEAKNESSES, STRENGTHS, AND OBSERVATIONS 

Neither the weaknesses nor the observations described in the inspection report 

require any specific action by or written response from NEP. The more 

significant weaknesses, strengths, and observations identified by the team are 

given below. 

Sections 3.4.2.1, "Eigenvalue Selection and Uncertainties," and 

3.4.2.2 , "Col d Shutdown Margin , 'I descri be a recogni zed weakness in core 

react i vi ty ca 1 cui at i ons . 

Section 3.4.2.3, "Steady-State Nuclear Methods Improvements," 

describes new design methods as a potential strength. 

Section 3.4.3.1, "Cycle-Dependant Safety limit Minimum Critical Power 

Ratio," describes observations regarding a cycle-dependant SLMCPR methodology. 

Section 3.4.4.2(1), "La Salle 1 Cycle 7," describes the development 

of a comprehensive and workable design interfacing procedure (DIP) as a 

strength. 

Section 3.4.4.2(2), "Hatch 1 Cycle 16," describes the lack of a 

comp'lete verification of the FRED data as a weakness. 

Section 3.4.4.2(2), "Hatch 1 Cycle 16," describes the implementation 

of the Transiefit Selection Review Committee as a strength and noted, however, 

that it was a weakness to not require documenting the committee's selection 

basis. 

Section 3.4.4.2(3), "Limerick 1 Cycle 16," describes the high level 

of automation in the design and analysis process as contributing to the 

overall quality of the analysis process. 

Section 3.5.5, "2.6 Safety Limit MCPR," describes a weakness in 

adherence to the reporting requirements in Amendment 22 of GESTAR 11. 

Section 3.6.3.6, "Grinding," describes an observation regarding the 

hydrogen analysis of unground pel 1 ets. 

Section 3.8.1, "Training," describes a weakness in previous level and 

extent of training and noted that the new training program will need to be 

monitored to determine whether the weaknsss has been remcved. 

- A-1 - 

155 
APPENDIX B 

OPEN ITEMS 

The report includes several open items that could not be resolved during the 

inspection, or for which the team needs additional information to reach its 

conclusions. NEP is requested to submit a written response, as described in 

the enclosed report, for the open items given below. 

Open Item 95-01-01, Section 3.4.2.2(2), "Pilgrim Cycle 11," requested 

the current shutdown margin value at the most reactive time in Cycle 11. 

Open Item 95-01-02, Section 3.4.3.4, "Quadrant-Symmetric Fuel 

Failures," requested that the NRC be notified when the failure mechanism was 

ident i f i ed . 

Open Item 95-01-03, Section 3.4.5.1, '#Rod Bowing," requested the 

correlation of the measurements of rod bow. 

Open Item 95-01-04, Section 3.4.6.2(1), IIGESTRM," requested a 

response to the burnup extension issue. 

Open Item 95-01-05, Section 3.6.2, "Safety Classification," requested 

a functional evaluation for fuel bundle parts. 

Open Item 95-01-06, Section 3.6.9.1(1), "Metallurgical Laboratory," 

requested that the NRC be notified when all corrective actions for CAR 95-12 

are complete. 

I - B-1 - 
APPENOIX C 

PERSONS CONTACTED 

The U.S. Nuclear Regulatory Commission staff participating in the inspection 

of General Electric Nuclear Energy activities conducted at the Nuclear Energy 

Production facilities in Wilmington, North Carolina and the Nuclear Energy 

Production personnel contacted during this inspection are 1 isted below and 

designated as follows: 

entrance meetings; a dagger (t) indicates that person attended the interim 

exit meetings. 

a bullet (0) indicates that person attended the 

Reload Core Desian. Safety Analysis. and Licensincr Processes: 

August 14, 1995, Entrance Meeting 

August 24, 1995, Interim Exit Meeting 

Nuclear Energy Production: 

Akerlund, S.O. 

t Alzaben, A.F. 

Armijo, J.S. 

t Babb, S.J. 

Baka, G.M. 

t Batchlor, C.D. 

t Baumgartner, J.A. 

Bolger, F.T. 

t Bowman, J.S. 

Brayman, K.W. 

t Brechtlein, T.D. 

t Brohaugh, T.R. 

t Butrovich, R.M. 

t Congdon, S.P. 

t Currier, J.W. 

t Dix, G.E. 

t Elkins, R.B. 

Embley, J.L. 

t fawks, R.A. 

t Galloway, G.D. 

Gardner, K.E. 

t Gibbs, E.W. 

Halls, J.L. 

t Harmon, J.L. 

Hoffman, P.K. 

Hauser, T.M. 

t Hull, G.R. 

Jackson, R.O. 

t Kingston, R.E. 

t Klapproth, J.F. 

Marlowe, M.O. 

Principal Enqiqeer, Operating Fuel Performance/Support 

Technical Program Manager, Nuc! ear Fuel Americas 

General Manager, Nucl ear Fuel 

Manager, Components Quality Assurance, 

Technical Program Manager, Nuclear Fuel Americas 

Manager, Channe?s Product Line 

Fuel Project Manager, Nuclear Fuel Americas 

Transient Analysis Process Leader, 

Design Process Improvement (DPI) 

Team Member, D31 

Manager, GENE Systems, NQA 

Acting Manager, Nuc’lear Fuel Asia 

fuel Project Manager, Nuclear fuel Americas 

Fuel Project Manager, Nuclear Fuel Americas 

Manager, DPI 

Manager, Customer Service 

Manager, Development Program, Nuclear Fuel 

Manager, Mechanical Design 

Licensing Program Manager, Fuels & Facility Licensing 

Sr. Engineer, OPI 

Process Team Leader, DPI 

Process Team Leader, DPI 

Technical Program Manager, Nuclear Fuel Americas 

Program Manager, Technology Licensing 

Manager, New Product Introduction & Value Engineering 

Manager, GENE Audits 

Manager, Environmental Health & Safety 

and Nuclear Quality Assurance (EH&S/NQA) 

Fuel Project Manager, Nuclear Fuel Americas 

Process Team Leader, DPI 

Fuel Project Manager, Nuclear Fuel Americas 

Manager, Fuel s and Faci 1 i ty Licensing , EH&S/NQA 

Manager, Fuel Materi a1 s Programs, Nuclear Fuel 

Nuclear Quality Assurance (NQA) 

- c-1 - 

157 
APPENDIX C Continued 

McCaughey, D.A. 

t Mills, V.W. 

Milmoe, C.J. 

Nichols, K.W. 

Pogosian, A. 

t Potts, G.A. 

Rand, R.A. 

t Rash, J.L. 

Reda, C.A. 

t Russell, W.E. 

Sependa, W.J. 

t Serell, D.C. 

Sick, P.W. 

t Smith, C.W. 

Soulis, R.E. 

t Stepp, M.R. 

t Stier, D.P. 

t Stirn, R.C. 

t Tuttle, J.L. 

Wei, P. 

Wileman, J.T. 

t Williams, R.D. 

t Stachowski, R.E. 

Manager, Fuel Qual i ty 

Design Automation Project Manager, DPI 

Legal Counsel 

Acting Manager, Manufacturing Facilities 

Production Manager, Control Rods 

Manager, Operating Fuel Performance/Support 

Principal Engineer, 3perat i ng Fuel Performance/Support 

Licensing Program Manager, Fuels & Facility Licensing 

Manager, Information Management Systems 

Technical Program Manager , Nuclear Fuel Americas 

Manager, Chemical Product Line 

Technical Program Manager, Nuclear Fuel Americas 

Manager, NQA 

Sr. Licensing Engineer, Fuels & Facility Licensing 

Acting Manager, Nuclear Fuel Asia 

Nuclear Methods Project Manager, DPI 

Manager, Nuclear Fuel Europe 

Technical Program Manager, Nuclear Fuel Americas 

Manager, New Prcdwt Introduction 

Project Manager, Fuel Engineering QA 

Principal Engineer, DPI 

Technical Program Manager, Nucl ear Fuel Americas 

Fuel Project Manager, Nuclear Fuel Americas 

U.S. Nuclear Regulatory Comnission: 

t Beyer, C.E. 

t Brewer, D.H. 

Carew, D+. J.F. 

Grow, R.L. 

t Kendrick, E.D. 

Lacy, P.S. 

t Matthews, S.M. 

Neogy, Dr. P. 

t Phillips, L.E. 

t Wu, Dr. S.L. 

Neutronics Specialist, Battelle-Pacific Northwest 

Labor at ory 

Metallurgical Engineer, Division of Inspection 

and Support Programs (DISP), Special Inspection 

Branch (PSIB) , Vendor Inspection Section (VIS) 

Physicist, Brookhaven National Laboratory 

Neutronics Specialist, Parameter, Inc. 

Reactor Engineer, Division of Systems Safety 

Neutronics Specialist, Parameter, Inc. 

Qual i ty Assurance Speci a1 i st, DISP/PSIB/VIS 

Physicist, Brookhaven National Laboratory 

Reactor Engineer, DSSA/SRXB 

Reactor Engineer, DSSA/SRXB 

and Analysis (DSSA), Reactor Systems Branch (SRXB) 

- c-2 - 

158 
- c-3 - 

159 

APPENDIX C Continued 

Fuel Desians and Fuel Production: 

August 21, 1995, Entrance Meeting 

August 25, 1995, Interim Exit Meeting 

Nuclear Energy Production: 

a.rmstrong, R.P. 

Babb, S.J. 

Baldwin, J.W. 

t Baumgartner, J.A. 

Bianchi, R.R. 

Brayman, K.W. 

Brechtlein, T.D. 

Bradberry, J.H. 

Brown, D.W. 

t Butrovich, R.M. 

Calcaterra, R.F. 

t Congdon, S. P. 

Currier, J.W. 

Davis, T.C. 

Downs, J.M. 

Dowker, D.K. 

Elkins, R.B. 

t Embley, J.L. 

English, D.G. 

Farella, M. 

Fawks, R.A. 

Gallerani, 3.5. 

t Frederickson, C.D. 

Haughton, R.A. 

t Hauser, T.M. 

Henderson, M.L. 

t Hull, G.R. 

t Kaiser, B.J. 

Keenan, R.J. 

Kipp, C.P. 

t Klapproth, J.F. 

Lamb, M.C. 

Laufer, S.C. 

Landry, J.D. 

Lion, E.A. 

McCaughey, D.A. 

McGowan, R.R. 

Milmoe, C.J. 

Mobley, E.R. 

Nunalee, J.J. 

Ogden, W. 

Palmer, E.E. 

t Pensinger, D.L. 

QA Engineer, Fuel Quality, NQA 

Manager, Components Qual i ty Assurance, NQA 

Qual i ty Auditor, Qual i ty Audits & Programs, NQA 

Fuel Project Manager, Nuclear Fuel Americas 

Qual i ty Engineer , Fuel Qual i ty , NQA 

Manager, GENE Systems, NQA 

Acting Manager, Nuclear Fuel Asia 

Leader, FMO Regulatory Team, Chemical Product Line 

Program Manager, Environmental Programs Team 

Fuel Project Manager, Nuclear Fuel Americas 

Sr. Engineer, Purchased Material QC, NQA 

Manager, DPI 

Manager, Customer Service 

Special i st, Organization Effectiveness 

Quality Auditor, Quality Audits & Programs, NQA 

Fuel Support Team Leader 

Manager, Mechanical Design 

Licensing Program Manager, Fuels & Facility Licensing 

Quality Auditor, Quality Audits & Programs, NQA 

Sr. Manufacturing Engineer, Channels Product Line 

Sr. Engineer, DPI 

Project Manager-GEI2, New Product Introduction 

Manager, SourcingjPurchased Materi a1 QC & 

Team Leader , Rod Fabrication 

Manager , EH&S/NQA 

Sr. Specialist, Fabrication Product Line 

Fuel Project Manager, Nuclear Fuel Americas 

Manager, Fabrication Product Line 

Program Manager, Compliance Auditing 

General Manager, Nucl ear Energy Production 

Manager, Fuels and Facility Licensing, EH&S/NQA 

Team Leader, Powder Prep and Pack 

Manager, Software QA & New Product Introduction 

Leader, Techno1 ogy Team, New Product Introduction 

Qual i ty Engineer, Fuel Qual i ty , NQA 

Manager, Fuel Qual i ty 

Leader, URLS/Waste Treatment Team 

Legal Counsel 

Quaiity Control Specialist, Fuel Quality, NQA 

Qual i ty Auditor, Qual i ty Audits & Programs, NQA 

Manager, Manufacturing Facilities 

Project Engineer, Channels Product Line 

Manager, QA Programs, Europe, NQA 

Shipping/Traf f ic 

and Value Engirwering 
APPENDIX C Continued 

Plotycia, G.D. 

Pope, D.D. 

t Potts, G.A. 

Quintana, L.M. 

t Rash, J.L. 

Reda, C.A. 

Reda, R.J. 

Rochelle, D.L. 

Rochelle, S.J. 

t Schardt, J.F. 

Scott, A.W. 

Selby, S.E. 

Sependa, W.J. 

t Serell, D.C. 

Sick, P.W. 

t Smith, C.W. 

Smith, G.H. 

Smith, M.W. 

Snell, D.H. 

t Stepp, M.R. 

t Stirn, R.C. 

Strickler, H.R. 

Sullivan, A.E. 

Sweet, F.W. 

Tarrer, C.C. 

t Tuttle, J.L. 

Theriault, K.M. 

Williams, W.K. 

Manager, Materials Services Product Line 

Qual i ty Auditor, Qual i ty Audits and Programs, NQA 

Manager, Operating Fuel Performance/Support 

Manager, Services Components QA, NQA 

Licensing Program Manager, Fuels 81 Facility Licensing 

Manager, Information Management Systems 

Manufacturi ng Techno1 ogi st, Fabri cat i on Product L i ne 

Quality Auditor, Ouality Audits & Programs, NQA 

Specialist, Fuel Release, Mechanical Design 

Manager, Tubing Product Line 

Quality Auditor, Quality Audits & Programs, NQA 

Leader, UO Production Line 

Manager, Ckemical Product Line 

Technical Project Manager, Nuclear Fuel Americas 

Manager, NQA 

Sr. Licensing Engineer, Fuels & Facility Licensing 

Leader, FMO Maintenance Support Team 

Nuclear Engineer, DPI 

Manager, Control Rods Product Line 

Manager, Nuclear Fuel Europe 

Manager, New Product Introduction 

Manager, Environmental Protection & Industrial Safety 

Sr. Program Manager, Order To Remittance 

Quality Auditor, Quality Audits & Programs, NQA 

Leader, Configuration Management 

Project Manager, Fuel Engineering QA 

Leader, Uranium Recovery/Process Team, 

Qual i ty Control Speci a1 i st, Fuel Qual i ty, NQA 

Chemical Product Line 

U. S. Nucl ear Regulatory Cod ss i on : 

Brewer, D.H. 

t Beyer, C.E. 

Brite, D.W. 

Czajkowski, C.J. 

t Kendrick, E.D. 

t Matthews, S.M. 

t Phillips, L.E. 

Schafer, A.C. 

Thompson, J.W. 

t Wu, Dr. S.4. 

Met a1 1 urg i cal Eng i neer , D I S P/ PS I B/V I S 

Neutronics Specialist, Battelle-Pacific Northwest 

Labor ir tory 

Sr. Research Engineer, Battelle-Pacific Northwest 

Laboratory 

Sr. Research Engineer, Brookhaven National Laboratory 

Reactor Engineer, DSSA/SRXB 

Qual i ty Assurance Special ist, DISP/PSIB/VIS 

Chief, DSSA/SRXB 

Chemical Engineer, Parameter, Inc. 

Reactor Engineer, Office for Analysis and €valuation 

of Operational Data (AEOD) , Di agnostic Eva1 uat i on 

and Incident Investigation Branch (DEIIB) 

Reactor Engineer, DSSA/SRXB 

c-4 - 

160 
APPENDIX C Continued 

Exit Meetina. Nuclear Enerw Production - September 1, 1995 

Nuclear Energy Production: 

Armi joy J.S. 

Babb, S.J. Manager , Fuel Components QA 

Baker, W. B. 

Brechtlein. T.D. Acting Manager, Nuclear Fuel Asia 

General Manager, Nuclear Fuel 

Manager, Qual ity Audits & Programs 

Congdon , S P. 

Creech, M.G. 

Currier, J.W. 

Dowker, D.K. 

Davis, T.C. 

Diersing, M. L. 

Downs, J.M. 

Elkins, R.B. 

English, D.G. 

Fawks, R.A. 

Guilliams, B.S. 

Hanvelt, R.A. 

Hauser, T.M. 

Haughton, R.A. 

Hoffman, P.K. 

Howard, D.C. 

Hull, G.R. 

Kaiser, B.J. 

Kingston, R.Z. 

Kipp, C.P. 

K1 approth, J. F. 

Laing, C.F. 

Lamb, M.C. 

McCaughey , D. A. 

McGowan , R R . Mobley, E.R. 

Nixon, C.L. 

Nunal ee, J T. 

Ogden , W. 

Pace, R.C. 

Parnell, R. I. 

Pinckney, C.C. 

Plotycia, G.D. 

Pope, D.D. 

Quintana, L.M. 

Reda, C.A. 

Reda, R.J. 

Scott, A.W. 

Schardt, J.F. 

Selby, S.E. 

Sependa, W.J. 

Manager, DPI 

Specialist, Shipping & Traffic 

Manager, Customer Service 

Team Leader, Environmental Process Team 

Specialist, Organization Effectiveness 

Team Leader, Ceramic Production, 

Fabrication Product Line 

Quality Auditor, Quality Audits & Programs, NQA 

Mariager, Mechanical Design 

Qual i ty Auditor, Qual ity Audits & Programs, NQA 

Sr. Engineer, DPI 

Sr. Program Manager, Information Management Systems 

Manager , Nucl ear Fuel Americas 

Manager , EH&S/NQA 

Team Leader, Rod Fabrication 

Manager , GENE Audi t s 

Acting Manager, Channels Product Line 

Fuel Project Manager, Nuclear Fuel Americas 

Manager, Fabrication Product Line 

Fuel Project kianager, Nuclear Fuel Americas 

General Manager , Nucl ear Energy Product ion 

Manager, Fuel an3 Facility Licensing, EH&S/NQA 

Manager, Purchased Materi a1 QC 

Team Leader, Powder Prep and Process 

Manager, Fuel Qual i ty 

Leader, URLS/Waste Treatment Team 

Qual i ty Control Speci a1 i st , Fuel Qual i ty, NQA 

Material Control and Accountabil ity (MC&A) Operator, 

Emergency Preparedness and MC&A, Fuel and Faci 1 i ty 

Licensing 

Quality Auditor, Quality Audits & Programs, NQA 

Manager, Manufacturing Facilities 

Manager, Chemet Lab 

Principal Engineer, Metallurgy, Chemet Lab 

Quality Support Coordinator 

Manager, Materials Services Product Line 

Qual i ty Auditor, Qual i ty Audits & Programs, NQA 

Manager, Services Components QA, NQA 

Manager, Information Management Systems 

Manufacturing Technologist, Fabrication Product Line 

Quality Auditor, Quality Audits & Programs, NQA 

Manager, Tubing Product Line 

Leader, UO Production Line 

Manager, Ckemi cal Product Line 

- c-5 - 

161 
APPENDIX C Continued 

Sick, P.W. 

Smith, C.W. 

Snell, D.H. 

Stepp, M.R. 

Stirn, R.C. 

Sweet, F.W. 

Theriaul t, K.M. 

Torres, R.L. 

Tuttl e, J. L. 

Walker, H.F. 

Williams, W.K. 

Manager, NQA 

Sr. Licensing Engineer, Fuels and Facility Licensing 

Manager , Control Rods Product Line 

Manager, Nucl ear Fuel Europe 

Manager, New Product Introduction 

Quality Auditor, Quality Audits & Programs, NQA 

Leader, Uranium Recovery/Process Team, 

Auditor, Compliance Auditinq 

Project Manager, f-uel Engineering QA 

Leader, Shipping & Traffic Team 

Qual i ty Control Speci a1 ist , Fuel Qual i ty, NQA 

Chemical Product Line 

U.S. Nuclear Regulatory Conmission: 

Cwalina, G.C. 

Brewer, D.H. 

Matthews, S.M. 

Schafer, A.C. 

Thompson, 3. W. 

Chief, DISP/PSIB/VIS 

Metal 1 urgi cal Engineer, DISP/PSIB/VIS 

Qual i ty Aszlircnce Speci a1 i st, DISP/PSIB/VIS 

Chemical Engineer, Parameter, Inc. 

Reactor Engineer, AEOD/DEI IB 

- C-6 - 

162 
r 

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 205554001 

Apri 1 8, '1996 

Mr. L. Charles Spriggs 

Vice President, Product Assurance 

PROMATEC, Inc. 

P.O. Box 309 

Cypress, TX 77429 

SUBJECT: NRC INSPECTION NO. 99901292/95-01 

Dear Mr. Spriggs: 

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

(NRC) inspection of Promatec Incorporated facilities in Cypress, Texas by this 

office on December 6-8, 1995. 

The inspection was conducted to evaluate Promatec's qual ity assurance program 

as implemented in the process for providing fire barriers penetration seals to 

the nuclear industry. 

design change controls, product testing and installation, procurement of 

materials, and document controls. In addition, the inspectors reviewed 

Promatec's program for implementing Part 22, "Reporting of Defects and 

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

Part 21). 

Specific areas of review included, product design and 

During the inspection the inspectors examined procedures and representative 

records, for activities such as receipt inspections and product dedication, 

and conducted discussions and interviews with personnel. 

observations were discussed with the staff at the conclusion of the 

inspection, on December 8, 1995. 

The inspectors' 

Based upon its observations and evaluation, the team concluded that the 

silicone-based nuclear power plant fire barrier applications designed, tested, 

and installed by Promatec are commercial grade items, as defined in 

10 CFR 21.3. Although the silicone products supplied for nuclear plant fire 

barrier applications are not subject to the NRC's Part 21 regulation, it 

appears that Promatec established and implemented a system which meets the 

intent of certain 10 CFR Part 21 requirements. 

inspection and our conclusions are discussed in the enclosed inspection 

report. 

Areas examined during the 

163 
L. Charles Spriggs -2- 

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

this letter and the enclosed inspection report will be placed in the NRC 

Public Document Room. If there are any questions concerning this inspection 

we will be pleased to discuss them with you. 

requ i red. 

No response to this letter is 

Si ncerel y , 

arb\\ -' 

onald P. Norkin, Acting Chief 

Speci a1 Inspection Branch 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

Docket No. 99901292 

Enclosure: Inspection Report No. 99901292/95-01 

164 
I 

U.S. NUCLEAR REGULATORY COMMISSION 

OFFICE OF NUCLEAR REACTOR REGULATION 

DIVISION OF INSPECTION AND SUPPORT PROGRAMS 

ORGANIZATION: 

REPORT NO: 

ORGANIZATIONAL 

CONTACT : 

NUCLEAR INDUSTRY 

ACTIVITY: 

INSPECTION CONDUCTED: 

LEAD INSPECTOR: 

OTHER INSPECTORS: 

REV1 EWED : 

Promatec, Inc. 

P.O. Box 309 

Cypress, TX 77429 

99901 292/95-0 1 

Mr. L. Charles Spriggs, Vice President 

Designed, tested and installed rated fire barrier 

penetration seal assemblies in nuclear power 

electrical generating stations. The assemblies 

combined commercial grade si1 icone products and 

various provr4etary additives developed to meet 

specific fire barriers penetration seal fire test 

criteria. 

December 6-8, 1995 

.. . .- '7 

jc. I e.:&- k.- "L.\ 

c, . YC,L-. 

Peter-5. Kol tay, Team Leader 

Special Inspection Section 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nucl ear Reactor Regul at i on 

Armajit Singh, NRR 

Christofer Bajwa, NRR 

T.L. Tinkel , Brookhaven National Laboratory 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nucl ear Reactor Regul at i on 

APPROVED : '1 

Donald 

,&\~\,L 

P. Norkln!v Actinq Chief 

a Date 

Special Inspection Branch 

Division of Inspection and Support Programs 

Office of Nucl ear Reactor Regul ati on 

Enclosure 

165 
1 .o INSPECT ION SUMMARY 

Based on the scope of this inspection, the team determined that Promatec, Inc. 

does not manufacture and supply any "basic components" as defined in Part 21, 

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

Requlations (10 CFR Part 21). While some customers may contractually impose 

the requirements of 10 CFR Part 21 on Promatec, Inc. for their fire barrier 

penetration seal assemblies supplied for use in nuclear power plants, the 

inspectors verified that the products did not meet the definition of basic 

components as defined in 10 CFR 21.3. However, the inspectors found that 

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

certain sections of 10 CFR Part 21, such as 10 CFR 21.21 (b) . Promatec, Inc. 

personnel stated to the team that they will continue to implement the existing 

program for meeting the intent of 10 CFR Part 21. 

In response to NRC concerns regarding the fire endurance qualification testing 

of cable slot type fire barrier penetration seals at the Watts Bar Nuclear 

Plant, additional testing on site specific penetration configurations were 

conducted. The test assemblies were constructed by the penetration 

contractor, Peak Seals (Promatec). The tests were performed on October 22, 

1995, at Omega Point Laboratories, Elmenford, Texas. All test assemblies met 

the acceptance criteria (section 3.5). 

Appendix A to this report. 

Details of the test are described in 

2.0 STATUS OF PREVIOUS INSPECTION FINDINGS 

This was the first NRC staff inspection of this vendor; thus, there were no 

previ ous f indi ngs . 

3.0 

3.1 

INSPECTION FINDINGS AND OBSERVATIONS 

Entrance Meeting 

The entrance meeting was held on Wednesday, December 6, 1995. 

meeting, Mr. Koltay, the NRC inspection team leader, discussed the scope of 

the inspection with the Promatec staff. 

At this 

3.2 Background and Description of Facilities 

Following the Browns Ferry nuclear power plant fire, increased emphasis was 

placed on fire protection at commercial nuclear power plants. 

ftre barrier penetration seals, using silicone based foam and elastomer 

materials, were developed to seal openings created in safety related 

structures designed to accommodate piping, cables, and ventilation duct work. 

A new line of 

Promatec was formed in September 1983 from the fire safe division of B&B 

Insulation company. 

results of fire tests of fire barrier penetration seals of several companies 

including TechSil, ICM and Peak Seals. 

Subsequently, Promatec also obtained the rights to the 

Promatec provides a full line of fire barrier penetration seal services 

including engineering, supply of materials, training, and inspections. The 

2 

166 
company’s qual i ty assurance program sat i sf ies the requirements of 10 CFR Part 

50, Appendix B and ANSI N45.2. Promatec has performed fire barrier 

penetration seal work at over 40 domestic nuclear utilities. 

3.3 Penetration Seal Configurations and Sealing Materials 

The Promatec corporate test procedure index, dated July 10, 1995, indicates 

that Promatec has tested and qualified dozens of fire barrier penetration seal 

design configurations between 1976 and 1994. 

The penetration seal design configurations vary depending on many factors 

including the type of fire barrier construction, the type of penetration, and 

the size of the opening being sealed. Examples of the various kinds of 

penetration seal configurations include: 

0 Penetrations for electrical cables 

Penetrations for electrical cable pipe conduits 0 Penetrations for electrical cable trays 

Penetrations for piping and ducts 

0 

0 

Raw materials, silicone foam, silicone elastomers, and silicone gel, are 

purchased from the Dow Corning and the General Electric corporations. 

Generally, Promatec reformulates the raw materials using proprietary 

ingredients and sells them under various names such as PROMAFLEX (FIREFLEX), 

LDSE, HDSE, RADFLEX, and others. 

3.3.1 Material Receipt Inspection 

The inspectors reviewed the receipt inspection and Quality Control (QC) 

procedures. 

procedures and are clearly marked with QC tags which define the status of the 

materials. 

comnercial nuclear facilities are tested by Promatec or by an independent 

laboratory to verify manufacturing specifications, prior to use. For example, 

elastomer and silicone seal materials are tested by Promatec, while filler 

materials are tested by independent laboratories to assure that the materials 

are acceptable according to Promatec’s specifications. These independent 

verifications ensure that the fillers are of the correct purity and are free 

of contaminants. QC involvement is required in all repacking operations and 

batching processes so that inadvertent use of untested material will be 

prevented. 

Raw materials are receipt inspected in accordance with detailed 

Materials that are used in products supplied to domestic 

Overall the inspectors found that Promatec has a well maintained material 

control program. All stored material is clearly marked with its inspection 

status. Items are identified with a sticker which contains the receiving 

information for the material. Generally, all material is identified with the 

receiving report number, date of receipt, purchase order (PO) number, lot 

number when applicable, and shelf1 ife or expiration date. 

3 

167 
Proportioning records and procedures for the mi xi ng and formul at i on of 

Promatec products, were reviewed by the inspectors. The inspectors noted that 

complete and detailed batching records are established and the identification 

of materials is maintained throughout the operation. Additionally, the 

procedures are established and very well implemented for maintaining shelf 

life of stored materials and assignment of shelf life to repackaged and 

blended products. 

The inspectors also reviewed the dedication activities which are defined in 

procedure QAP-0001, "Procurement Control," Issue J, dated March 28, 1994. 

inspection and test requirements are prescribed for material specifications. 

A review of the material specifications and inspection instruction against 

test results received for a sample of items found that appropriate test 

results were provided. 

established an adequate test and inspection program. 

identify any deficiencies in this area of the inspection. 

The 

The inspectors concluded that the vendor has 

The inspectors did not 

3.3.2 Procurement Document Control 

The inspectors reviewed a sample of Promatec customer procurement documents 

including some recent customer Purchase Orders (POs). 

customers impose the provisions of 10 CFR Part 21 on individual POs, while 

other customers do not invoke 10 CFR Part 21 at all. 

reviewed by QA prior to issuance. A Customer Traveler form is prepared which 

is used to record all vital information, from the receipt inspection number, 

to the classification of the order as safety or non-safety related, to the 

final shipping information. This sheet is attached to the PO. Also attached 

to the POs is a purchase specification, which provides information required to 

assure the quality of the materials to be provided. 

inspectors were in compliance with the established procedures. 

required, POs are placed with suppliers listed on the approved vendor list 

(AVL). 

suppliers. 

will be used and whether there is additional testing required to complete 

dedication, special instructions, etc. The inspectors concluded that the 

vendor has implemented a good program in the area of procurement. 

3.3.3 Fabrication/Assembly and Special Processes 

Some of Promatec's 

Requisitions and POs are 

All POs reviewed by the 

Where 

Promatec's AVL contains both Appendix B and commercial grade 

The list clearly distinguishes how the supplier was qualified and 

The procedures for producing Promatec products address the batching of foam 

and elastomeric materials. 

special processes performed by Promatec on the materials they supply. 

batching of formulations is conducted by specified individuals and is overseen 

by a QC inspector. The batching is performed in accordance with the 

instructions provided in proportioning procedures and is recorded on the 

batching record. Scales used to weigh the materials are confirmed to be 

accurate by calibration by an outside supplier. 

identify any discrepancies in this area of inspection. 

The inspectors also noted that there are no 

The 

The inspectors did not 

4 

168 
3.3.4 Cal i brat ion 

The inspectors reviewed the calibration program at Promatec. 

cal i brat i ons are subcontracted to a survey qual i f i ed suppl i er . The vendor 

provided adequate evidence of traceability, accuracy and certification of its 

supplier. 

calibration sticker indicating the calibration date, due date for next 

calibration and the identification of the person who did the calibration. The 

calibration records provided the status, location, calibration frequency, and 

as-found and as-left data for each scale. 

vendor has an adequate cal i bration program. 

3.3.5 Document Control 

The weight scale 

The weight scales were found to be uniquely identified and had a 

The inspectors concluded that the 

The vendor has implemented a documentation control program which meets its 

procedure requirements and QA program commitments. The procedures were found 

to provide sufficient detail and direction. Personnel are assigned controlled 

copies of appl icable procedures; the control led procedures issued were the 

correct revision and stamped with a control number. 

maintained and drawing files conta;nc.;1 previous revisions to drawings, where 

applicable. 

the inspection. 

Drawings are well 

The inspectors did not identify any deficiencies in this area of 

3.4 Quality Assurance and Design Control 

The inspectors reviewed the vendor’s Quality Assurance Program (QAP) document 

QAM 20188, Issue D, dated June 28, 1995, which establishes those planned and 

systematic actions necessary to assure that products provided by the company 

will perform satisfactorily in accordance with their intended purpose. In 

addition, that such products, including services associated therewith, comply 

with customer specifications and all corresponding regulatory requirements. 

The Qual ity Assurance Manual incorporates the appl icable requirements of 

ANSI/ASME NQA-1-1989 EDITION (ANSI N45.2). 

to the following; (1) penetration seals and protective coatings for fire, 

pressure, smoke and hot gas, hydrostatic, and/or thermal protection; (2) fire 

protection systems for separation of electrical trains, boundary separations 

including flexible blankets and rigid barriers; (3) thermal insulation 

specified for use with safety-related equipment; and (4) activities and 

services affecting quality, including design, testing, materials, 

manufacturing , instal 1 ati on, inspection, and training . The inspector 

concluded that the vendor’s QA program and manual establishes the 

organizational structure, the responsibilities of individuals, and departments 

performing activities governed by this program. 

identify any discrepancies in this area of inspection. 

3.4.1 

Promatec qualifies the design of a penetration seal configuration using one of 

two methods ., 

The vendor’s QA program is applied 

The inspectors did not 

Methodology for Qualifying Penetration Seal Configurations 

a. The first method consists of subjecting the penetration seal design 

configuration to a fire endurance test. The results of successful 

5 

169 
testing are used to qualify the particular parameters of the design. 

This method was used extensively during the development of the fire 

barriers penetration seals. 

penetration test results on file. 

Promatec maintains a library of fire 

b. The second method consists of performing an engineering analysis for 

the penetration seal design configuration requiring qualification in 

the field and establishing a correlation between the field 

penetration and tested penetration configurations on file. 

The engineering analysis method for qualifying penetration seal configurations 

permits Promatec to use a combination of engineering judgment and results from 

multiple prior tests to qualify new or modified design configurations without 

perf ormi ng addi ti onal testing . 

Engineering analysis for qualification consists of analyzing reports of 

pertinent past tests performed by or available to Promatec, and identifying 

qualitative and quantitative information for selected design parameters that 

represent the candidate design configuration and are covered by prior testing. 

The inspectors noted that the Promatec methodology for qualifying 

configurations using engineering analysis is not documented except for the 

notes on drawings. Based on a review of drawings and discussions with the 

Operation’s Manager and the Product Assurance Manager, Promatec’ s past practice and methodology were reviewed. 

valid range of the following design parameters that can be used by Promatec 

for qualifying candidate penetration seal configurations: 

The team identified and verified the 

a. 

b. 

C. 

d. 

e. 

f. 

Primary seal material twe. The candidate seal material must be the 

same as the tested seal material. In the event the particular seal 

material is available from more than one vendor, a comparative test 

of the two seal materials must be performed. 

Primary seal material deDth. Tfie candidate seal material depth must 

be equal to or greater than the tested seal material depth. 

Fire barrier orientation. Tested floor orientations can be used to 

qualify a candidate floor or wall orientation. Tested wall 

orientations can only be used to qualify candidate wall orientations. 

Service temDerature. The candidate design service temperatures must 

be equal to or less than the service temperature of the seal 

material. 

Radiation resistance. 

candidate configuration must be equal to or less than the radiation 

exposure of the seal material. 

The required radiation exposure of the 

Maximum Denetrant movement (1 ateral and/or axial 2. The allowable 

motion of the candidate configuration i,bUSt be equal to or less than 

the allowable motion of the seal material. 

6 

170 
9. 

h. 

i. 

j. 

k. 

1. 

m. 

n. 

0. 

P* 

Maximum openina size. 

Configuration must be equal to or less than the maximum opening size 

of the tested configuration. 

The maximum opening size of the candidate 

Maximum free area fa1 so referred to as maximum UnsuDDorted area). 

The maximum free area of the candidate configuration must be equal to 

or less than the maximum free area of the tested configuration. 

Minimum clearance of Denetrants. The minimum clearance between 

penetrants or between penetrants and the substrate (fire barrier 

boundary) in the candidate configuration must be equal to or greater 

than the minimum clearances in the tested configuration. 

Maximum annular area (for DiDe/sleeve desisnsl. The maximum annular 

area of the candidate configuration must be equal to or less than the 

maximum annular area of the tested configuration. 

Penetratinq ob.iect material tvoe. The penetrating object material 

type of the candidate configuration must be the same as the tested 

configuration. 

Fire barrier ratinq. 

penetration configuration shall be equal to or less than the fire 

barrier rating of the tested configuration. 

The fire barrier rating of the candidate 

Cable wav fill. The percent fill of the candidate penetration 

configuration must be less than or equal to the percent fill of the 

tested configuration. The ignitability of the insulation for the 

candidate configuration must be equal to or less than that in the 

tested configuration. 

Damminq material. The same type of damming material and thickness 

used on the tested configuration must be used on the candidate 

penetration if the tested configuration did not use a dam.) 

Penetrant item wall thickness. The wall thickness of the candidate 

configuration must be equal to or less than the wall thickness of the 

tested configuration. 

Penetrant item size. 

equal to or less than the size of the tested configuration. 

The size of the candidate configuration must be 

Since engineering analysis is an important method for qualifying Promatec 

penetration seal configurations, the inspection team recommended that 

consideration should be given to formally documenting and controlling the 

engineering analysis method in a Promatec engineering standard or procedure. 

The acceptance criteria and limitations for each of the selected design 

parameters along with the technical basis or technical justification for the 

analysis method should be included. 

7 

171 
The inspectors verified that parameters are 1 imited by tested configurations; 

however, there is no limitation on the length and width for the parameter that 

establishes the acceptable range in the maximum free area parameter 

assessment. Promatec management agreed to review this weakness and establish 

appropriate 1 imitations as needed. 

The inspector assessed the quality of completed engineering analyses by 

reviewing the following resultant Fire Barrier Seal Design Test Report: 

Test report CTP 1063 dated 21 January 1985 indicated that the unexposed 

side temperature limits defined by ASTM E-814 of 325 OF plus initial 

temperature (325 OF+ 84 OF = 409 "F) were exceeded during the 3 hour test 

for a number of penetrations included in the test. The high temperatures 

occurred on thermocouples located at the unexposed side at the interface 

of the sealant material and the penetrating items. 

The inspectors noted that portions of this test were used as the 

qualification basis for certain site specific penetration configurations 

at a power plant. Additionally, the power plant penetration seal 

specification (Rev. 1 dated 3 April 1992) invokes a maximum temperature 

of 325 "F on the unexposed side surface and penetrating items. All the 

above noted penetrations, as well as a number of others in this test, 

failed to meet this requirement. This matter was discussed with the 

Promatec Product Assurance Manager. The inspectors verified that the 

Promatec f i re qual i f i cat i on test were conducted i n accordance wi th 

ANI/AMERP for which the limiting end point temperatures on the unexposed 

side of the penetrations are less restrictive than those of ASTM E-184 in 

that the ANI test does not specify the temperature of the penetrating 

item on the unexposed side. 

regulatory expectations, were satisfied by the test. 

for the power plant installation were appropriately adjusted. 

The ANI acceptance criteria, that meets 

The requirements 

In conclusion, the review of actual design configurations did not identify any 

design issues. However, the implementation of the engineering analyses to 

qualify tested fire barrier penetration seals on file, for installation into 

fire barrier penetrations of various configuration, partially re1 ies on 

informal engineering judgment. 

and controlling such aspects of the engineering analysis methods in a 

engineering standard or procedure. 

This could be improved by formally documenting 

3.5 Fire Endurance Test of Cable Slot Penetration Seals 

On October 22, 1995, the concrete test slab, containing the 14 cable slot 

penetration seal test assemblies was subjected to a 3 hour fire endurance test 

which followed the ASTM E-119 standard time-temperature curve and a fog nozzle 

hose stream test. The acceptance criteria of IEEE 634-1978, "Standard Cable 

Penetration Fire Stop Qualification Test," was used to evaluate the 

thermal/fire resistive performance of the test assemblies. 

requires the test assemblies to: 1) withstand the fire endurance test without 

the passage of flame or hot gases hot enough to 'gnite cables on its unexposed 

This criteria 

8 

172 
side; 2) heat transmission through the penetration seal shall not raise the 

temperature on its unexposed surface above 700 OF; and 3) not allow water to 

be projected through the penetration seal during the hose stream test. 

All test assemblies, met the IEEE 634 acceptance criteria. 

unexposed cable/seal interface temperatures ranged from 323 "F for instrument 

and control cables and 601 OF for power cables. 

The maximum 

The staff, based on the results of the applicant's supplemental cable slot 

fire endurance tests, finds those "as-built" penetrations which have been 

installed in accordance with WBN cable slot penetration seal design details 

A4, H1, L1, and M4 and that are bounded by the tested cable fill (thermal mass 

of copper conductors) conditions will provide an equivalent level of fire 

safety to those which were tested and, therefore, they are acceptable. 

3.6 Exit Meeting 

The exit meeting was conducted on December 8, 1995, by the NRC team leader 

prior to the team's departure from the Promatec facility. 

9 

173 
APPENDIX A 

TRIP REPORT 

TRIP DATE: October 15-22, 1995 

REVIEWER: Patrick Madden, Senior Fire Protection Engineer 

Fire Protection Section 

P1 ant Systems Branch 

Division of Systems Safety and Analysis 

LOCATION: Omega Point Laboratory 

San Antonio, Texas 

APPLICANT: Tennessee Valley Authority (TVA) 

SUBJECT : CABLE SLOT 3-HOUR PENETRATION SEAL TEST SPECIMEN CONSTRUCTION AND 

FIRE ENDURANCE TEST 

1.0 APPLICANTKONTRACTOR PERSONNEL CONTACTED: 

Joseph A. Lisa, TVA (QA) 

3. J. Pierce, TVA (ENGINEERING) 

Charles Spriggs, PROMATEC/Peak Seals 

Michael Jordan, PROMATEC/Peak Seal s 

DeDe Smi thwick, PROMATEC/Peak Seals 

James Granc i 0, PROMATEC/Peak Seal s 

Mike Murphy, PROMATEC/Peak Seal s 

2.0 DESCRIPTION OF TEST ASSEMBLY 

The test assembly consists of a 8' x 13' by 12-inch thick concrete test 

slab with 14 5" x 20" cable slot blockouts. On one half of the slab, 

8 cable tray slots (A1 through A8) are arranged in two parallel columns 

with 4 cable slots in each column. The slots in each column are 

separated by a 7" wide concrete mullion and a 6" concrete mullion exists 

between the cable slot ends between the columns. 

columns were constructed so that they were maintained at least 24" away 

from the edge of the test slab. 

through B6) are located on the second half of the slab and are arranged 

in two parallel columns with 3 slots in each column. 

separated by a 6" wide concrete mullion and each cable slot within each 

column is separated by a 7" wide concrete mullion. 

columns were constructed so that the edges of cable slots columns were 

maintained at least 24" away from the edge of the test slab. 

following summarizes the cable fill of each cable slot test specimen: 

The two cable tray slot 

The remaining 6 cable tray slots (B1 

The columns are 

These two cable slot 

The 

- Penetration Seal (PS) Test Specimen A1 - single layer of 4/c-#16 

(43 cables) 

A- 1 

174 
- PS Test Specimen A2 - 100% visual fill of 4/c-#16 (230 cables) 

- 

- 

- 

- 

- 

- 

PS Test Specimen A3 - 100% visual fill of 4/c-#16 (230 cables) 

PS Test Specimen A4 - 50% visual fill of 4/c-#16 (150 cables) 

PS Test Specimen A5 - 50% visual fill of 4/c-#16 (150 cables) 

PS Test Specimen A6 - single layer of 4/c-#16 (43 cables) 

PS Test Specimen A7 - blank spare (no cable fill) 

PS Test Specimen A8 - blank spare (no cable fill) 

(3 cables) , 2/0-8Kv (9 cables) , 4/0 (2 cables) , #2 (2 cabl es) , #6 (4 cables), 3/c-#10 PXMJ (8 cables) , and 3/c-#10 CPJJ (3-cables) 

I - PS Test Specimen 81 - contains 300MCM (14 cables), 2/0-600v 

I 

I - PS Test Specimen B2 - ccnf?i(*s 300MCM (9 cables) 

- PS Test Specimen B3 - contains 300MCM (7 cables), 2/0-8Kv (3 cables), 4/0 (4 cables), #2 (1 cables), #6 (2 cables), 3/c-#10 PXMJ (10 

cabl es) , and 3/c-#10 CPJJ (6-cab1 es) 

- PS Test Specimen 84 - contains 300MCM (20 cables), 2/0 (12 cables), 

4/0 (4 cables), #2 (2 cables), #6 (4 cables), 3/c-#10 PXMJ 

(14 cables), and 3/c-#10 CPJJ (6-cables) 

PS Test Specimen B5 - contains 300MCM (9 cables) - 

- PS Test Specimen B6 - contains 300MCM (7 cables), 2/0-8Kv (3 cables), 

4/0 (4 cables), #2 (1 cables), #6 (2 cables), 3/c-#10 PXMJ (10 

cabl es) , and 3/c-#10 CPJJ (6-cab1 es) 

3.0 PENETRATION SEAL 

General DescriPtion of Foam Installation 

For PS test specimens A1 through A6, each seal was constructed by 

installing a damming board (Carborundum Fiberfax 1-in thick low density 

board) on the exposed (fire) surface of the specimen and filling the 

blockout void with 12-inches of Dow Corning RTV 3-6548 silicone foam and 

flush with the surface of the concrete. Once the foam has been injected 

into the blockout void a damming board is installed on the unexposed 

surface of the slab. 

PS test specimens A7 and A8 are spare penetrations with a sleeve that 

extends 4-inches out away from the wall on each side. On the exposed 

(fire) side of PS test specimen A7 the damming board is attached to the 

end of the sleeve and 11-inch foam fill is injected into the cable slot 

blockout penetrations thus creating a 9" air gap between the damming 

175 
board and the foam on the unexposed side of the seal. PS test specimen 

A8 was constructed in the same manner except that the 9" air gap was on 

the exposed (fire) side of the seal. 

On the exposed (fire) side of PS test specimens B1 through 86 a damming 

board is installed and ll-inch thickness of foam was injected into the 

blockout to fill the void. On the unexposed side l-inch thickness of 

ceramic fiber (Carborundum Durablanket) is installed between the foam 

seal and the outer damming board on the unexposed surface. 

Procedure Review 

The following procedures related to the installation of the silicone foam 

seal test specimens were reviewed: 

1. Promatec Procedure OCP-0067, 7/1/91, "Density rerification" 

(procedure sets a minimum density of Iqlbs/ft as compared to TVA 

specified density range of 15-30 lbs/ft ) 

2. PCI/ICMS - Tennessee Valley Authority Procedure QC-101, Revision 9 

(7/15/95) "QC Inspection Silicone Foam" 

(procedure establishes the inspection criteria for the inspection of 

foam samples and installed seals) 

3. PCI/ICMS - Tennessee Valley Authority Procedure QC-103, Revision 4 

(7/17/95) , "Density Measurement Mu1 ti-Component Si1 icon; Material" 

(procedure sets a minimum density range of 15-30 lbs/ft . This is 

consistent with TVA's penetration seal engineering report) 

4. PCI/ICMS - Tennessee Valley Authority Production Work Instruction 

PWI-051, Revision 9 (9/22/94) , "Installation of Damming Material" 

(procedure provides guidance to the craft regarding the installation 

of damming materials prior to pouring silicone materials) 

5. PCI/ICMS - Tennessee Valley Authority Production Work Instruction 

PWI-052, Revision 7 (4/14/94) "Installation of Si1 icone Foam" 

(procedure sets the criteria for inspecting blockouts and damming 

prior to silicone installation) 

6. TVA Maintenance/Addition Instruction MAI-3.6, "Cable Tray and Sleeve 

Seals," Revision 7 (9/19/95), Unit 0, Appendix E, Sealing Cables in 

Vert i cal $1 eeves . 

Penetration Seal Materi a1 s 

Damming Materials Used: 

Carborundym Fiberfax 1-in thick low density board (density - 15 to 

18 lbs/ft ), Lot Nos. 3053 and 4311, date of manufacturer 3/30/95 

Carborundum Durablanket, 1-in thick, density - 6 lbs/ft3, Lot No. 5117, 

date of manufacturer 5/19/95 

A-3 

176 
Silicone Foam Material: Dow Corning 3-6548 Silicone Foam 

Part A - Lot No. ET55392A (shelf life 10/31/96) 

Part B - Lot No. ET065158B (shelf life 10/31/96) 

General Methods of Installation 

The applicant used MAI-3.6, "Cable Tray and Sleeve Seals," Revision 7 

(9/19/95), to install the penetration seal test specimens B-2, B-4, and 

B-6. The following summarizes the general installation methods used by 

the craft to install the foam in these penetration seals: 

1. Distribute the cables evenly, lift cables and firmly pack a two-inch 

depth of ceramic fiber (Kaowool or equivalent) around and between 

each cable at both ends of the sleeve to provide approximately 1/4-inch separation between cables and metal surfaces. (Note: the 

2. 

3. 

4. 

5. 

6. 

7. 

a. 

ceramic fiber extends at least one-inch on each side of'the minimum 

depth of seal material .) 

Cut a temporary form and install temporary form over bottom end of 

sleeve . 

Apply silicone foam around cables and between the cables from the top 

side of the sleeve. (Note: An initial sealant application is made 

and allowed to cure. This acts as a base which eliminates excessive 

leakage of uncured material through the bottom side of the sleeve.) 

Cart a temporary form and cut three one-inch holes in the form for 

foam applications. Install the temporary form over the top side of 

the cable sleeve. 

Apply silicone foam inside the sleeve through the application holes. 

Remove the temporary forms and any excess sealing material. 

(Note: Random surface irregularities or voids less than 1/2-inch in 

their major dimension are acceptable.) 

Apply ceramic fiber board cement to the surface of foam and edges of 

the sleeve and allow to set 3-5 minutes. 

Install pieces of one-inch ceramic fiber board over each cable sleeve 

face with a 1/2 inch overlap of the sleeve edges. 

fiber board firmly onto previously prepared surfaces. 

Press ceramic 

For the remaining penetration seal test specimens (6-2, 6-4, 6-6, and A-1 

through A-8) , the appl icant used QC-101 , QC-103 , PWI-051 , and PWI-052 to 

install these penetration seal test specimens. The following summarizes 

the major quality verification attributes and general installation 

methods used by the craft to install and QC personnel to inspect these 

foam penetration seal s: 

E\-4 

177 
QC-101 - Quality Control Inspection Silicone Foam 

1. Permanent forms (dams) shall fit snug of shall be mechanically 

fastened to the face of the barrier on both openings. 

2. Electrical cable trays in floor/wall configurations are to be formed 

to allow a minimum depth of 12 inches of foam. 

3. Electrical cables are spread where possible to allow foam to flow 

freely around them. 

4. Silicone foam samples are prepared in accordance with PWI-052. One 

of the two samples shall be cut open and examined for the 

following: (a) foam color shall be dark grey to black; (b) cell 

structure shall be uniform; (c) foam texture shall be set and firm, 

and (d) material tear shall be crisp and firm. The second sample 

shall be trimmed flush with the top of the cup and weighed to the 

nearest tenth of a gram. (Note: Acceptable density range for 

silicone foam is 15 to 30 pounds per cubic foot. 

5. Key inspection criteria are: (a) seals are neatly trimmed where 

needed; (b) foam color is dark grey to black; (c) foam cell structure 

uniform; (d) foam texture is set and firm; (e) penetration fill 

sufficient; (f) seal conforms with detail shown on customer supplied 

field data sheet, and (9) permanent forms are properly fitted; and 

materi a1 is acceptable. 

QC-103 - Density Measurement Multicomponent Si1 icone Material 

1. Determine the weight and volume of sample cup. 

2. Determine weight of sample cup filled with silicone material and 

cal cul ate density . 

PWI-051 - Installation of Damming Material 

1. Forming materials be cut and fit snug into or against the barrier to 

prevent 1 eakage. 

2. Where cables protrude through penetration, care should be taken to 

tightly pack damming materials to prevent excessive leakage. 

3. In electrical cable trays, cab?es will be spread where possible to 

allow material to flow around and between them. Blanket or bulk 

fiber material may be used for this purpose. 

PWI-052 - Installation of Silicone Foam 

1. Distribute a light layer of silicone foam on the form to seal off any 

potential leaks. 

A- 5 

178 
2. Electrical cable trays in floor/wall configurations are to be formed 

in accordance with a minimum twelve inches of silicone foam. 

4.0 CONSTRUCTION ACTIVITIES WITNESSED: 

OCTOBER 17. 1995 - OBSERVATIONS 

The following construction activities were witnessed: 

1. To facilitate the installation of damming materials the test slab 

(assembly) was configured in a vertical position. 

fitting of the damming boards for the exposed side of the individual 

penetration test specimens was started and welding of damming board 

studs to cable slot steel liners was completed on the unexposed side 

of the test assembly. 

Cutting and 

2. Filling void spaces between the cables, cables and the cable tray, 

and the cable tray and the sides of the cable slot blockout (test 

specimen B-2, 8-4, and 8-6) with Durablanket per TVA MAI-3.6 and 

installation of Carboruncknl 'iberfax damming board. 

OCTOBER 18, 1995 - OBSERVATIONS 

The following construction activities were witnessed: 

1. 

2. 

3. 

4. 

5. 

Cutting and fitting of the damming boards and welding of damming 

board studs to cable slot steel liners. 

Installation Carborundum Fiberfax damming board and filling void 

spaces between the cables in the damming board cable cutouts with 

Durablanket per PWI-051. 

Setup of foam gun and mixing air pump and the pre-mixing of silicone 

foam Part A and B prior to filling the Part A and B injection pump 

reservoirs. 

Calibration of the foam density sample cup (verified that the scale 

used to determined weights was calibrated - calibration date 10/2/95, 

re-cal i brati on due 3/31/96) 

Test assembly was placed in a horizontal configuration to facilitate 

filling the penetration seal test specimens with foam. 

OCTOBER 19. 1995 - OBSERVATIONS 

The following construction activities were witnessed: 

1. The cables in the individual penetration seal test specimens were 

2. 

spread in order to facilitate the flow of foam between the cables. 

Foam samples were taken to determine its density. Using the PCI/ICMS 

density method the initial density was 24.1 lbs/ft3 (average of three 

A-6 

179 
samplel). 

1 bs/ft was achieved. 

When the slab was placed in the horizontal condition the cables slid 

from their original position. The cables had to be re-pulled to meet 

the test plan specification. This resulted in some of the 

aurablanket damming material to be dislodged. 

the cables from sliding during seal installation, the cables were 

secured on the unexposed side of the test assembly to the cable trays 

This was done prior to re-working the 

dams and filling voids around the cables and the damming boards with 

Durabl anket ceramic fiber materi a1 . 

Using the PROMATEC density method a density of 19.64 

3. 

In order to prevent 

- with stainless steel tie wire. 

4. During the initial pouring/injection of the silicone foam into 

penetration seal test specimen A-4 (first seal to be filled), the 

silicone gun experienced a problem with the silicone foam Part B 

injection nozzle and shutoff valve. This resulted in improper mix 

of Part A and B in the mixing chamber and poor foam color qual i ty. 

Work was stopped to repair the gun. 

5. As result the foam gun problems, Penetration Seal A-4 was complete 

re-worked. The damming' and fo 

seal opening. The damming was 

re-sealed with si1 icone foam. 

6, As result of the foam gun prob 

was performed tgree additional 

was 24.5 1 bs/ft . 

Y 

m was removed from the penetration 

re-cut and re-installed and the seal 

ems a reveri f i cat i on of foam density 

times. The final average foam density 

OCTOBER 20. 1995 - OBSERVATIONS 

The following construction activities were witnessed: 

1. Final installation of foam completed, penetration seal test specimens 

were trimmed to meet specification requirements established by TVA's 

test plan. At the start of work a foam fensity measurement was made. 

The average foam density was 25.6 lbs/ft . 

Installation of engineering thermocouples on unexposed side si1 icone 

foam surfaces by 1 aboratory personnel . 

Installation of Carborundum Fiberfax 1-in thick low density damming 

board on unexposed side of each penetration seal test specimens. 

Installation of qualification thermocouples on the individual 

penetration seal test specimens by the Laboratory personnel. The 

test assembly was instrumented with a total of 286 thermocouples. 

2. 

3. 

4. 

5.0 FIRE ENDURANCE AND HOSE STREAM TEST 

DATE OF TEST: October 22, 1995 

A- 7 

180 
TEST DURATION: 3-hour (ASTM El 19 - standard time-temperature curve) 

TEST ACCEPTANCE CRITERIA: IEEE 634-1978, "Standard Cable Penetration Fire 

Stop Qualification Test" 

This criteria requires that: (1) the test specimen withstand the fire 

endurance test without the passage of flame or gases hot enough to ignite 

cables on its unexposed side; (2) heat transmission through the penetration 

seal shall not raise the temperature on its unexposed surface above 700 OF, 

and (3) not allow water to be projected through the penetration seal during 

the hose stream test. 

FURNACE PRESSURE: 0.01 inches of water 

TEST ASSEMBLY ORIENTATION: 

TYPE OF HOSE.STREAM TEST: 

Horizontal 

Fog (30 "F pattern, 75 gpm @ 75 psi); Duration 

of hose stream application is 2-1/2 minutes and 

was applied within 10 minutes after the fire 

endur:.r.c s? test. 

INITIAL TEMPERATURE CONDITIONS: 69 OF 

TEST OBSERVATIONS: 

0:02 Cables on the exposed side fully ignited 

1:lO Penetration seal test specimens 63, 65, and 66 emitting smoke 

through the unexposed side of the seal. 

discoloration is occurring on the ceramic fiber damming board on 

the unexposed side of these penetration seals. 

Slight smoke streaming 

1:40 

2: 17 

Penetration seal test specimen 64 emitting smoke fairly steady. 

Smoke streaming discoloration is occurring on certain areas of the 

unexposed side ceramic fiber damming board. 

All "6" penetration seal test specimens emitting steady smoke 

through the penetration seal s and cab1 e bund1 es. 

RESULTS: (maximum temperatures noted during the test) 

a. Test Specimen: A1 

Cable Fill: single layer of 4/c-#16 (43 cables) 

Unexposed side temperatures: 

- On top of the ceramic fiber damming board: 112 OF 

- Interface between the seal and cable: 205 OF 

- Interface between the seal and tray side rail; 160 "F 

- On the steel liner of the cable slot: 144 OF 

- Under damming board on top of silicone foam: 150 "F 

A-8 

181 
I 

b. Test Specimen: A2 

Cable Fill: 100% visual fill of 4/c-#16 (230 cables) 

Unexposed side temperatures: 

- On top of the ceramic fiber damming board: 150 OF 

- Interface between the seal and cable: 280 OF 

- Interface between the seal and tray side rail: 196 OF 

- On the steel liner of the cable slot: 179 OF 

- Under damming board on top of silicone foam: 229 OF 

c. Test Specimen: A3 

Cable Fill : 100% visual fill of 4/c-#16 (230 cables) 

Unexposed side temperatures: 

- On top of the ceramic fiber damming board: 156 OF 

- Interface between the seal and cable: 323 OF 

- Interface between the seal and tray side rail: 160 OF 

- On the steel liner of the cable slot: 151 OF 

- Under damming board on top of si 1 icone foam: 263 OF 

Cable Fill: 50% visual fill of 4/c-#16 (150 cables) 

Unexposed side temperatures: 

- On top of the ceramic fiber damming board: 150 "F - Interface between the seal and cable: 264 OF 

- Interface between the seal and tray side rail: 137 OF 

- On the steel liner of the cable slot: 159 OF 

- Under damming board on top of silicone foam: 

Cable Fill: 50% visual fill of 4/c-#16 (150 cables) 

Unexposed side temperatures: 

- On top of the ceramic fiber damming board: - Interface between the seal and cable: 318 OF 

- Interface between the seal and tray side rail: 151 OF 

- On the steel liner of the cable slot: 180 OF 

- 

d. Test Specimen: A4 

210 OF 

e. Test Specimen: A5 

156 OF 

Under damming board on top of silicone foam: 

Cable Fill : single layer of 4/c-#16 (43 cables) 

Unexposed side temperatures: 

- On top of the ceramic fiber damming board: 123 OF 

- Interface between the seal and cable: 243 OF 

- Interface between the seal and tray side rail: 135 OF 

- On the steel liner of the cable slot: 187 OF 

- Under damming board on top of silicone foam: 161 OF 

227 OF 

f. Test Specimen: A6 

A-9 

182 

i 
h. 

i. 

j. 

k. 

Test Specimen: A7 

Cable Fill : SPARE 

Unexposed side temperatures: 

- On top of the ceramic fiber damming board: 155 OF 

- On the steel liner of the cable slot: 295 OF 

- Under damning board on top of silicone foam material: 311 OF 

Test Specimen: A8 

Cable Fill : SPARE 

Unexposed side temperatures: 

- On top of the ceramic fiber damming board: 110 OF - On the steel liner of the cable slot: 328 OF - Under damming board over the air space: 141 OF 

Test Specimen: B1 

Cable Fill : 300MCM (14 cables), 2/0-600v (3 cables), 2/0-8Kv (9 cables), 

Unexposed side temperatures: 

4/0 (2 cables), #2 (2 cables), #6 (4 cables), 3/c-#10 PXdJ 

(8 cables), and Sjc-%lO CPJJ (3-cables) 

- On top of the ceramic fiber damming board: 192 OF 

- Interface between the seal and cable: 574 OF 

- Interface between the seal and tray side rail: 225 OF - On the steel liner of the cable slot: 194 OF 

- Under damming board on top of ceramic fiber blanket: 385 OF 

Test Specimen: 82 

Cable Fill: 300MCM (9 cables) 

Unexposed side temperatures: 

- On top of the ceramic fiber damming board: 

- Interface between the seal and cable: 527 OF 

- Interface between the seal and tray side rail: 139 OF 

- On the steel liner of the cable slot: 162 OF 

- 

158 OF 

Under damming board on top of ceramic fiber blanket: 138 OF 

Test Specimen: 83 

Cable Fill : 300MCM (7 cables), 2/0-8Kv (3 cab1 es) , 4/0 (4 cables), #2 

(1 cables), #6 (2 cables), 3/c-#10 PXMJ (10 cables), and 

3/c-#10 CPJJ (6-cables) 

Unexposed side temperatures: 

- On top of the ceramic fiber damming board: 

- Interface between the seal and cable: 595 OF 

- Interface between the seal and tray side rail: 233 OF 

- On the steel liner of the cable slot: 168 OF 

- 

208 OF 

Under damming board on top of ceramic fiber blanket: 264 OF 

183 
1. Test Specimen: B4 

Cable Fill : 

Unexposed side temperatures: 

300MCM (20 cables) , 2/0 (12 cables) , 4/0 (4 cables) , #2 

(2 cables), #6 (4 cables), 3/c-#10 PXMJ (14 cables), and 

3/c-#10 CPJJ (6-cables) 

- On top of the ceramic fiber damming board: 

- Interface between the seal and cable: 562 "F 

- Interface between the seal and tray side rail: 298 "F 

- On the steel liner of the cable slot: 195 OF 

- 

263 "F 

Under damming board on top of ceramic fiber blanket: 434 OF 

m. T,est Specimen: B5 

Cable Fill : 

Unexposed side temperatures: 

300MCM (9 cables) 

- On top of the ceramic fiber damming board: 

- Interface between the seal and cable: 564 OF 

- Interface between the seal and tray side rail: 148 OF 

- On the steel liner of the cable slot: 165 "F 

- Under damming board on top of ceramic fiber blanket: 381 OF 

236 OF 

n. Test Specimen: 86 

Cable Fi 11 : 300MCM (7 cables) , 2/0-8Kv (3 cables) , 4/0 (4 cab1 es) , #2 

Unexposed side temperatures: 

(1 cables), #6 (2 cables), 3/c-#10 PXMJ (10 cables), and 

3/c-#10 CPJJ (6-cab1 es) 

- On top of the ceramic fiber damming board: 200 OF 

- Interface between the seal and cable: 601 "F 

- Interface between the seal and tray side rail: 180 OF 

- On the steel liner of the cable slot: 141 "F - Under damming board on top of ceramic fiber blanket: 287 OF 

All test specimens, met the IEEE 634 acceptance criteria. 

unexposed cable/seal interface temperatures ranged from 323 OF for instrument 

and control cables and 601 OF for power cables. 

The maximum 

6.0 POST-FIRE TEST SPECIMEN EXAMINATION 

Post-fire tear down and examination of the test specimens was performed on 

October 23, 1995. Due to the reviewer's travel schedule, only a portion of 

these examination was wi tnessed . The fol 1 owing summarizes the reviewer's 

observations made during the examination process: 

Specimen A1 - Bottom side of cable bundle closest to cable tray side rail 

had no foam remaining. 

sleeve and the top of the cable bundle had 6" of foam 

remaining. Cable jacket and insulation damage was present 

through the full thickness of the seal. 

The penetration space between the 

A-11 

184 
Specimen A3 - 

Specimen A7 - 

Specimen A8 - 

Specimen B1 - 

The ceramic fiber damming board on the unexposed side 

sustained smoke damage. Bottom side of cable bundle closest 

to cable tray side rail had no foam remaining. 

penetration space between the sleeve and the top of the 

cable bundle had 3" of foam remaining. Cable jacket and 

insulation damage was present through the full thickness of 

the seal. 

The 

3" of undamaged foam was present at the sleeve interface and 

approximately 1" at the center of the penetration seal. The 

lower ceramic damming board was heavily fire damaged but, 

remained intact and the upper damming board was undamaged. 

9" of undamaged foam was present at the sleeve interface and 

approximately 7" at the center of the penetration seal. The 

lower ceramic damming board was heavily fire damaged but, 

remained intact and the upper damming board was undamaged. 

Bare copper and sever jacket and insulation fire damage up 

to the unexposed ceramic fiber damming board. 
.... -_. 

.. ..... .- 

A 

. 

........ ... 

-* 

.. 

.. 

-- ........ 

.. .-.- .___ ..... 

-- ._. . ---. 

....... ..--.- 

.-. .-- ... -- . -- 

.- .. 

'r 

....... .... 

.............. __.. 

..- .-.- ........ 

... - .. . .I ... 

.-- -. ... ...-. - .__.. 

-.--._I-.-.--. , 

.. .-.-- ............ 

... 

... 

.......... .--- 

- 

.. 

..- 

I.. 

.. 

j- 

... ._- ..... 

........ .-- .... -..- .-. ... ......... 

... .._- .. -. .-__............. -.-..-.- 

\ 

. ._ 

I ... -.-. , 

I ......... .- . 

I 

I . ...----.. 

I ....--.- j .... 

... ..-...-. I 

1 ..... --. .. -. .. 

.... _-.- ...... _.-- I 

OCT 17 '95 18:50 

615 36s 1750 

. .-- .... 

PFICE. 42 
CC 

OCT 17 ’5 10:S0 

187 

615 365 1750 PRGE. 13 
~___ 

. TLA J89H (EN DES.%-78) TENNESSEE VALLEY AUTHORITY SnErt OF 

.. . . )t' 

I 

'h" '. 

. A- & 

18& 
I 'mail rl 

189 

615 365 1758 PAGE. 1s 
swctt OF 

- .. 

€9-€9 
OCT 17 ’95 10:s~ 

-01. 

a 

0 * 

0 0 

a 

e 

I 

TOT% P. 17 191 

PC?GF 17 
TENNESSEE VALLEY AUTHORITY S w EET OF 

DATI; 

w - 

--- 

34 

k I 

Y 4, \- 

I 

A/ -Al 

192 
131 .. .* .. .. 

ii II ii I/ 

iI I1 I1 Ii 

1 

-.-.. 

-.-.. 

24" 

7- 

-q 

'p I 

I 

I 

I 

I 

+"llF/ I 1 IL? Scale:l/2"=1' 

5" .----pf 

PLAN VIEW 

p" Po 0 

% *e 

0 ?L 

ORAIO 

r 

ATTACHMENT 3 193 
f 9" 1 10" f- 13" t 12" f-- 20" - 

PLAN VIEW 

3 

- 

11" 

9" - 

r 

-4 '*" 

OMEGA POINT LABORATORIES, INC. 

Project No. 1 121 0-99050 

TENNESSEE VALLEY AUTHORITY - WBN 

Fig. 2 Slab Rebars 

Drawn By: 

Approved By: 

Client Approval: 

194 
PLAN VIEW A 

1 -1 ............. 

............ 

............. 

............. 

.............. 

.............. 

............... ............. 

............. 

............ 

............ ........... 1 

.... .A ............. 

............. 

............. 

............. 

.............. 

............. ........... 

.......... ............ 

............ 

............ 

............ ......... 

SECTION B-B 

1-t 2" 

I/ /I 1 3" 

SECTION A-A 

*O" -7 ........ ........... 

........... 

........... 

............ 

............ 

............. 

............. 

............ 

........... 

........... 

........... 

............ 

............ 

............. 

............. 

............. 

........... 

........... 

........... 

............ 

........... 

............ 

.......... 

........... r ........... 

.......... 

T I 

12" 

NOTE: 

The upper and lower edges of the blockouts 

containing penetrations, included "picture 

frame" surrounds, fabricated of 3" x 2" x 3/8" 

steel (A36) angle, welded together at the 

ends and welded to 1/2"0 steel round stock, 

which was in turn welded to the concrete 

rebar as necessary to ensure structural 

stability. This blockout edging method was 

utilized for Penetrations A1 -A6 and B1 -B6. 

195 

Scale: 1-1/2"=1' 

I OMEGA POINT LABORATORIES, INC. 

Project No. 1 121 0-99050 

Tennessee Valley Authority - WBN I 

Fig. 3 Blockout Edge Details I 

Drawn By: g&&!$Tr d74- 

Approved By:P+ e/7/95- 

Client Approval: &flL @/, 2) 9 5 
i 20" -7 

5" rr 

PLAN VIEW 1/4" plate steel blockout liner 

(welded comers) 

............... .............. 

.............. 

............. 

.............. 

.............. 

.............. 

.............. 

.............. ............. 

............ 

............ 

........... 

........... ............ 

............. 

............ 

............. 

............. 

............. ............ 

........... ............. 

............. 

............ 

............. ........... 

.......... 

NOTE: 

The blockouts not containing penetrations 

included sheet steel sleeve surrounds, 

fabricated of 1/4" thick plate steel (A36), 

welded together at the comers and welded 

to 1/2"0 steel round stock, which was in tum 

welded to the concrete rebar as necessary 

to ensure structural stability. Additional 

studs were welded to the faces of the 

sleeve in contact with concrete to insure a 

solid ankhoring. This blockout edging 

mdthod was utilized for Penetrations A7 - --I AScale: 1-1/2"=1' 

I OMEGA POINT LABORATORIES, INC. 

Project No. 1 121 0-99050 

Tennessee Valley Authority - WBN I 

Fig. 4 Blockout Edge Details - A7, A8 I 

196 
14 steel angle perimeter frame 

318" 0 hole - 3/8% round 

Bolts - Hd inside tray (typ of 2) 

-qdHD Phillips HD ASTM A307 4 

................. .. ... A A A.A *'A 

PLAN VIEW 

112" 

.................. *-**A***# 

................. ******I* 

................. LAbAAAAA 

................ L+.**L**, 

NOTE: 

The above figure shows details of the 

cable tray mounting dips used at 

both side rails of each penetrating 

cable tray, on both the top and 

bottom metal blockout frames. 

112" .c------).: 

I 

4 4 CUTmAWAY VIEW 5" 

~~ 

1 OMEGA POINT LABORATORIES, INC. 4 

I Project No. 1 121 0-99050 

1 

! 

I TENNESSEE VALLEY AUTHORITY I 

I 

I Fig. 5 Cable Tray Mounting Clip Details 
36" 

! 

18" wide x 4 deep steel cable tray / 

)I' I 

***a 

, .. .. .. 

I 

i 

i 4s 

! 

12" , 

!! 

'I 

NOTE: 

The 18" wide x 4" deep steel cable trays extended 

36" above and 12" below the concrete slab. The 

trays did not pass through the penetration seal - each 

tray section extended 3' into the seal material. The 

trays were secured to me steel Mockout frames with 

the tray mounting clips detailed in Fig. 5. Due to the 

location of the bolts securing the tray sections to the 

tray mountiong clips, a minimum of 3' was atlowed 

between the end of the tray embedded in the seal 

and the first tray rung (5' for the lower tray sections). 

1 OMEGA POINT LABORATORIES, INC. 

1 Project No. 1 121 0-99050 I 

I 

TENNESSEE VALLEY AUTHORIIY 

I Fig. 6 Typical Cable Tray Elevation Details 

P/7/95- 

Client Approval: I$? 6?.? I o/;.-/ 95 

/0/'7/5 s1 Drawn by:%* I Approved by: C 

j 

! 

198 
18" wide x 4 deep steel cable tray 

I 

I 

19 

! 

:I ;/ I, 

3% i: I 

I i I! 

T ra 

*A*** 

......... 

........ 

1 , I I 

I 

! '< 

I I 

12" 

, 

A*. 

...... A*. 

h *A* .... 

h *A1 .. 4. 

...... 4.6, 

...... *A, 

a. *A. 4. 4. 

...... ** 

...... LA. 

...... 

= **a 

1. 

- ;I 

' 1 'I thick ceramic damming board, 

to outer edge of angle perimeter, 

I secured with wing nuts threaded I 

NOTE: 

The penetration seal consisted of : 1) a 1" thick layer 

of ceramic damming board, installed flush with the 

bottom surface of the slab, secured to 1/4" studs 

welded to the metal blockout frames with washers 

and wing nuts, 2) a 12' depth of silicone foam, 3) a 

1" thick layer of ceramic damming board, installed 

flush with the top surface of the slab, secured to 114" 

studs welded to the metal blockout frames with 

washers and wing nuts. The gaps between cables 

and damming board were filled with bulk ceramic 

fiber. 

i 

L-r onto studs welded to framework. (typ) 

1 OMEGA POINT LABORATORIES, INC. 

1 Project No. 1 121 0-99050 

TENNESSEE VALLEY AUTHORITY 

Fig. 7 Typical Penetration Seal Detail 

I I Penetration Set A (AI -A6) 

199 
'I 

w L a 

4' I 

1" 

.d 

A 

NOTE: 

The penetration seal consisted of : 1) a 1 " thick layer 

of ceramic damming board, installed flush with the 

bottom of the steel penetrating sleeve, secured to 

114' studs welded to clip angles welded to the outside 

of the steel sleeve with washers and wing nuts, 2) an 

11-112" depth of silicone foam, 3) an air gap of 8-1/2', 

and 4) a 1' thick layer of ceramic damming board, 

installed flush with the top of the steel penetrating 

sleeve, secured to 1/4' studs welded to clip angles 

welded to the outside of the steel sleeve with 

washers and wing nuts. 

1 1 1/2" 

! 

\ 1 I' thick ceramic damming board, 

secured with wing nuts threaded 

onto studs welded clip angle attached 

to steel sleeve. (typ) 

OMEGA POINT LABORATORIES, INC. 

Project No. 1 121 0-99050 

TENNESSEE VALLEY AUTHORITY 

Fig. 8 Penetration Seal Detail 

Penetration A7 

200 
A I 

1 ' thick ceramic damming board, 

secured with wing nuts threaded 

onto studs welded clip angle attached 

to steel sleeve. (typ) 

NOTE: 

The penetration seal consisted of : 1) a i' thick layer 

of ceramic damming board, installed flush with the 

bottom of the steel penetrating sleeve, secured to 

114" studs welded to clip angles welded to the 

outside of the steel sleeve with.washers and wing 

nuts, 2) an air gap of 8-1/2", 3) an 11-1/2" depth of 

silicone foam, and 4) a 1 thick layer of ceramic 

damming board, installed flush with the top of the 

steel penetrating sleeve, secured to 1/4" studs 

welded to clip angles welded to the outside of the 

steel sleeve with washers and wing nuts. 

1 OMEGA POINT LABORATORIES, INC. 

I Project No. 1 121 0-99050 f 

I 

TENNESSEE VALLEY AUTHORITY 

Fig. 9 Penetration Seal Detail 

Penetration A8 

Drawn by: /dW/ IJ' 

Approved by: W/b./kr 

Client Approval: 

201 
36" 

18" wide x 4" deep steel cable tray / $1 ,J 

!I 

I' 

i 

'I 

I 

19 

I 

Bl ! 

A I * !i 

!' I 

! 

I: 

I \ 

I 

I 

I 

12" I \ I 

i i! 8 ' 1 thick ceramic damming board, 

i' I to outer edge of angle perimeter, 

I: i, secured with wing nuts threaded 1 

I 

'I It 

I 

- onto studs welded to framework. (typ) 

NOTE: 

The penetration seal consisted of : 1 ) a 1 " thick layer 

of ceramic damming board, installed flush with the 

bottom surface of the slab, secured to 1/4" studs 

welded to the metal blockout frames with washers 

and wing nuts, 2) an 1 1 " depth of silicone foam, 3) a 

1' thick layer of ceramic fiber blanket material, and 

OMEGA POINT LABORATORIES, INC. 

Project No. 1 121 0-99050 I 

I 

4) a 1 " thick layer of ceramic damming board, 

installed flush with the top surface of the slab, 

TENNESSEE VALLEY AUTHORITY 

secured to 1/4" studs welded to the metal blockout 

frames with washers and wing nuts. The gaps 

between cables and damming board were filled with 

Fig. 10 Typicai Penetration Seal Detail 

Penetration Set B (81-66) bulk ceramic fiber. 

Drawn by: 

Client Approval: 

202 
1 

vertical member of tray support Wp.) 

-1/4-20 x 2" threaded steel studs, 

-metal blockout frame 

welded 

PLAN VIEW - TOP OF SLAB 

1/4-20 x 2" threaded steel studs. welded (typ.) 

PLAN VIEW - BOTTOM OF SLAB 

NOTE: OMEGA POINT LABORATORIES, INC. 

The 1/4"-20 x 2" Ion steel studs were welded to the 

blockout perimeters as shown above. The ceramic 

damming board was then installed over the studs, 

with the board covering to the outer edges of the 

blockout frames, and was secured with washers and 

wing nuts. 

I Project No. 1 121 0-99050 

I TENNESSEE VALLEY AUTHORITY 

; 

i 

I 

i Drawnby: 

I ; Apmved&%$%$ 

I Client Approval: ,./A$ 35 

Fig. 11 Stud Placements For Damming 

Board Installation (A1 -A6, B1 -B6) I /+d/W 

/p/ao/9J203 
1/4"-20 x 2" threaded steel studs, welded 

to 1-1/2" angle dips welded to end of sleeve (typ.) 

the studs, and was secured with washers and wing 

nuts. 

\metal blockout sleeve 

3 1/2" 

Fig. 12 Stud and Clip Angle Placements 

For Damming Board Installation (A7-A8) I 

PLAN VIEW - TOP OF SLAB 

204 
I 

I 

! 

/ 

12 13 29 64 \ 

1 30 

/ TC tree #105- 1 10 46 

80 81 97 every 2" of seal depth 

/ 

110 

/ 1 28 127 

OMEGA POINT LABORATORIES, INC. 

Project No. 1 121 0-99050 

TENNESSEE VALLEY AUTHORITY 

Thermocouple Locations - Penetration Set A 

Under Upper Damming Board Material 

NOTE: 

24 gauge, Type K, Chromel-Alurnel 

Error: fl.l "C, purchased with lot traceability and 

calibration certifications) were placed in the locations 

illustrated above. All TCs shown above were placed 

prior to installation of the top damming board material 

on the seals. TCs were placed on the top surface cf 

the silicone foam (TC Nos. 12, 13, 29, 30, 46, 47, 63, 

64,80,81, 97, 98, 118, 119, 127, 128), on the 

surface of one side of the blockout perimeter frames 

11 1-1 14) 1 I' from the 1/4"-20 x 2" damming board 

stud. Additional TC trees were installed on one 

4C/#16 cable and in the field of the seal in Pen. A6. 

electrically-welded thermocouples (Special Limits of 

(TC NOS. 14-16, 31-34, 48-51, 65-68, 82-85, 

*+=A 

0 *A 

205 
A? A8 

NOTE: 

24 gauge, Type K, Chromel-Alumel 

electrically-welded thermocouples (Special Limits of 

Error: k1.1 "C, purchased with lot traceability and 

calibration certifications) were placed in the locations 

illustrated above. TCs were placed on the surface of 

the ceramic fiber damming board on the unexposed 

face of the seals. The TCs wer covered with 2" x 2" 

x 0.4' felted fiber pads. 

1 OMEGA POINT LABORATORIES, INC. 

Project No. 11 21 0-99050 

I 

! TENNESSEE VALLEY AUTHORITY 

I 

! 

1 

Thermocouple Locations - Penetration Set A 

Surface of Unexposed Seal Face 

206 
23 24 25 

40 41 42 

A7 A8 

I OMEGA POINT LABORATORIES, INC. 

Project No. 11 21 0-99050 

NOTE: 

24 gauge, Type K, Chrornel-Alumel 

electrically-welded thermocouples (Special Limits of 

Error: k1.1 OC, purchased with lot traceability and 

calibration certifications) were placed in the 

locations illustrated above. TCs were placed on 

cables (TC Nos. 6-8, 23-25, 40-42,57-59, 74-76, 

91 -93) and on tray side rails (TC Nos. 4-5, 21 -22, 

38-39,55-56,72-73, 89-90), at the interface 

between the penetrants and the upper damming 

board. The TCs were covered with 3/4" x 3/4" x 

0.4" felted fiber,pads. 

TENNESSEE VALLEY AUTHORITY 

I Thermocouple Locations - Penetration Set A 

Interface Between Seal and Penetrants 

207 
26 27 

63 44 45 

A7 A8 

OMEGA POINT LABORATORIES, INC. 

Project No. 1 121 0-99050 

TENNESSEE VALLEY AUTHORITY 

Thermocouple Locations - Penetration Set A 

NOTE: 

24 gauge, Type K, Chromel-Alurnel 

Error: k1.1 "C, purchased with lot traceability and 

calibration certifications) were placed in the 

locations illustrated above. TCs were placed on 

cables, 1" to 3" above the interface between the 

Penetrants and the UPPr damming board. The TCs 

were covered with 3/4" x 3/4" x 0.4" felted fiber 

electrically-welded thermocouples (Special Limits of 1 

pads. I On Cables Above Unexposed Seal Face 

208 

I 
A7 - 120 

121 

22 

+ 

123 

NOTE: 

24 gauge, Type K, Chromel-Alumel 

electrically-welded thermocouples (Special Limits 

of Error: fl.l°C, purchased with lot traceability and 

calibration certifications) were placed in the 

locations illustrated above. TCs on A1 -A6 were 

placed on the unexposed face of the seal system, 

directly above the steel perimeter framework (on 

ceramic damming), at least 1" from mounting studs. 

TCs on A7 & A8 were placed on the outer surface 

of the steel penetrating sleeve, 1" above the slab 

surface. TCs were covered with 2" x 2" x 0.4" 

felted fiber pads. 

t, : 

L 

- 

112 - 

113 

A8 - 129 

131 

13 

- 

132 

OMEGA POINT LABORATORIES, INC. 

Project No. 1 121 0-99050 

I TENNESSEE VALLEY AUTHORITY 

Thermocouple Locations - Penetration Set A 

I On UnexDosed Seal Face Above Frames 

I Drawn by: 

' Approved 

' Client Approval: 

209 
NOTE: 

24 gauge, Type K, Chromel-Alumel 

electrically-welded thermocouples (Special Limits of 

Error: _tl .1 OC, purchased with lot traceability and 

calibration certifications) were placed in the locations 

illustrated above. TCs Here placed on the surface of 

the ceramic fiber damming board on the unexposed 

face of the seals. The TCs wer covered with 2" x 2 

x 0.4" felted fiber pads. 

l 

OMEGA POINT LABORATORIES, INC. 

Project No. 1 121 0-99050 

TENNESSEE VALLEV AUTHORlTY 

Thermocouple Locations - Penetration Set B 

Surface of Unexposed Seal Face 

21 0 
161 \ 

NOTE: 

24 gauge, Type K, Chrornel-Alumel 

electrically-welded thermocouples (Special Limits of 

Error: fl.l "C, purchased with lot traceability and 

calibration certifications) were placed in the 

locations illustrated above. TCs were placed on 

cables (TC Nos. 160-162, 177, 190-192, 207-209, 

224,237-239) and on tray side rails (TC Nos. 

235-236), at the interface between the penetrants 

and the upper damming board. The TCs were 

covered with 3/4" x 3/4" x 0.4" felted fiber pads. 

158-159, 175-176, 188-189,205-206,222-223, 

OMEGA POINT LABORATORIES, INC. 

Project No. 1 121 0-99050 

TENNESSEE VALLEY AUTHORITY 

~~ 

Thermocouple Locations - Penetration Set B 

Interface Between Seal and Penetrants 

I 

21 1 
'3 #2 @UO (8KV) 

Q 210 s3/C#l0 -#6 

NOTE: 

24 gauge, Type K, Chromel-Alumel 

electrically-welded thermocouples (Special Limits of 

Error: fl.1 "C, purchased with lot traceability and 

calibration certifications) were placed in the 

locations illustrated above. TCs were placed on 

cables, 1 " to 3" above the interface between the 

penetrants and the upper damming board. The TCs 

were covered with 3/4" x 314'' x 0.4" felted fiber 

pads. 

OMEGA POINT LABORATORIES, INC. I 

Project No. 1 121 0-99050. 

TENNESSEE VALW AUTHORITY 

Thermocouple Locations - Penetration Set 6 

On Cables Above Unexposed Seal Face 

21 2 
156" 

! 

-m 

A 

278 cno 

- B5 1. ' B6 .- 275 - 

- 

273 274 

' OMEGA POINT LABORATORIES, INC. i 

! 

Project No. 1 121 0-99050 

TENNESSEE VALLEY AUTHORITY 

I 

i 

I 

NOTE: 

24 gauge, Type K, Chromel-Alurnel 

electrically-welded thermocouples (Special 

Limits of Error: fl.1 OC, purchased with lot 

traceability and calibration certifications) were 

placed in the locations illustrated above. TCs 

were placed on the slab surface adjacent to the 

blockouts, and in several locations 12" from the 

penetration blockouts. All thermocouples were 

covered with 2" x 2" x 0.4" felted fiber pads in 

accordance with ASTM E814. 

I 

I 

I ' Thermocouple Locations - Slab Surface I 

21 3 
Selected Generic Correspondence on the Adequacy of 

Vendor Audits and the Quality of Vendor Products 

Ident i f i er 

Information Notice 96-24 

Information Notice 96-29 

Information Notice 96-30 

Title 

Preconditioning of Molded-Case Circuit Breakers 

Before Survei 11 ance Test i ng 

Requirements in 10 CFR Part 21 for Reporting and 

Eva1 uat i ng Software Errors 

Inaccuracy of Diagnostic Equipment for MotorOperated Butterfly Valves 

214 
SRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER 

2-89) 

IRCM 1102. 

kdgned by NRC. AW Vol., Supp., Rev., and Addendum Numbers, if any.) 

201,3202 BIBLIOGRAPHIC DATA SHEET 

(See instructions on the reverse) NU R EG-00 40 

Vol. 20, No. 2 . TITLE AND SUBTITLE 

Licensee Contractor and Vendor Inspection 

Status Report 3. DATE REPORT PUBLISHED 

Quarterly Report MONTH YEAR 

April - June 1996 August 1996 

4. FIN OR GRANT NUMBER 

. AUTHORIS) 6. TYPE OF REPORT 

Quarterly 

7. PERIOD COVERED f/ncluuveoates~ 

April - June 1996 . PERFORMING ORGANIZATION - NAME AND ADDRESS (IfNRC,pmvide Division, OfficeorRegion, U.S Nuclear Resulamw Commission, andmailingaddress, if contractor,provide 

name and mailing addmd 

Division of Inspection and Support Programs 

Office of Nuclear Reactor Regulation 

U.S. Nuclear Regulatory Commission 

Washington, D. C. 20555-0001 . SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, type ‘Sam asabove’lifcontractor, provide NRCOivision, Officeor Region, U.S. Nuclear Regularow Commision, 

and mailing addred 

Same as above. 

0. SUPPLEMENTARY NOTES 

1. ABSTRACT 1200 words or less1 

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

Inspection Branch, Vendor Inspection Section, that have been distributed to the 

inspected organizations during the period from April through June 1996. 

L KEY 44ORDSIOESCR:PTORS 113. AVAILABILITY STATEMENT (Listwordsorphraoerthat willassistrssearchers in 1ocatinetherewrr.l 

1 Unlimited 14. SECURITY CLASS1 FlCATlON 

Vendor Inspection I (This fagel 

Uncl assi f i ed 

(This Report1 I I Unclassified 

15. NUMBER OF PAGES 

16. PRICE 

NRC FORM 335 (zag) 
on recycled 

Federal Recycling Program