ML20198H650
| ML20198H650 | |
| Person / Time | |
|---|---|
| Issue date: | 05/31/1986 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | |
| References | |
| NUREG-0040, NUREG-0040-V10-N01, NUREG-40, NUREG-40-V10-N1, NUDOCS 8605300516 | |
| Download: ML20198H650 (158) | |
Text
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NUREG-0040 Vol.10, No.1 l
l LICENSEE CONTRACTOR AND VENDOR INSPECTION STATUS REPORT QUARTERLY REPORT JANUARY 1986 - MARCH 1986 i
l UNITED STATES NUCLEAR REGULATORY COMMISSION p
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I NOTICE Availability of Reference Materials Cited in NRC Publications
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Most dccuments cited in NRC publications will be available from one of the following sources:
- 1. The NRC Public Document Room,1717 H Street, N.W.
Washington, DC 20655
- 2. The Superintendent of Documents, U.S. Government Printing Office, Post Office Box 37082, Washington, DC 20013-7082 1
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- 3. The National Technical Information Service, Springfield, VA 22161 l
Although the listing that follows represents the majority of documents cited in NRC publications, j
it is not intended to be exhaustive.
i Referenced documents available for inspection and copying for a fee from the NRC Publ c Docu-ment Room include NRC correspondence and internal NRC memorands; NRC Office of inspection j
and Enforcement 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 GPO Sales l
Program: formal NRC staff and contractor reports, NRC-sponsored conference proceedings, and N RC booklets and brochures. Also available are Regulatory Guides, NRC regulations in the Code of Federal Regulations, and NucArar 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 j
l Energy Commission, forerunner agency to the Nuclear Regulatory Commission.
i^
Documents available from public and special technical libraries include all open literature items, l
such as books, journal and periodical articles, and transactions. Federal Register notices, federal and 1
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.
j Single copies of NRC draf t reports are available free, to the extent of supply, upon written request l
to the Division of Technical Information and Document Control, U.S. Nuclear Regulatory Com-j mission, Washington, DC 20555.
l Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library, 7920 Norfolk Avenue, Bethesda, Maryland, and are available there for reference 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 j
American National Standards Institute,1430 Broadway, New York, NY 10018.
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i NUREG-0040 Vol.10, No.1 l
LICENSEE CONTRACTOR AND VENDOR INSPECTION STATUS REPORT QUARTERLY REPORT JANUARY 1986 - MARCH 1986 Manuscript Completed: April 1986 Date Published: May 1986 Division of Quality Assurance, Vendor and Technical Training Center Programs Office of Inspection and Enforcement j
U.S. Nuclear Regulatory Commission Washington, D.C. 20555 v" "*%,
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CONTENTS 1
Page 1.
Preface...........................................
iii fi '
2.
Reporting Fonnat (Sample).........................
v I
3.
NRC Approval Status of QA Program Topical i
Report Revisions................................
vii 4.
Sample QA Topical Report Revision l
Approval Letter.................................
ix 5.
Inspector Reports.................................
1 6.
Index of Inspection Reports.......................
137 k
7.
Table of Vendor Inspection Reports Related to Reactor Plants.......................
139 1
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PREFACE l
A fundamental premise of the Nuclear Regulatory Commission's (NRC) nuclear facility licensing and inspection program is that licensees are responsible for the proper construction and safe operation of their nuclear power plants.
The total government-industry system for the inspection of nuclear facilities has been desipned to provide for multiple levels of inspection and verification.
L.icensees, contractors, and vendors each participate in a quality verification process in accordance with requirements prescribed by, or consistent with, NRC rules end regulations. The NRC inspects to determine whether its requirements are being met by a licensee and his contractors, while the great bulk of the inspection activity is performed by the industry within the framework of ongoing 4
quality verification programs.
In implementing this multilayered approach, a licensee is responsible for developing a detailed quality assurance (QA) plan. This plan includes the OA programs of the licensee's contractors and vendors. The NRC reviews the licensee's and contractor's QA plans to determine that implementation of the proposed QA program would be satisfactory and responsive to NRC regulations.
In the :ase of the principal licensee contractors, such as nuclear steam supply system designers and architect engineering firms, the NRC encourages submittal of a description of corporate-wide QA programs for review and acceptance by the NRC. Upon acceptance by NRC, described QA programs provide written bases for inspection on a generic basis, rather than with respect to specific commitments made by a particular licensee. Once accepted by NRC, i
a corporate QA program of a licensee's contractor will be acceptable for all license applications that incorporate the program by reference in a Safety Analysis Report (SAR).
In such cases, a contractors's QA program will not be reviewed by the NRC as part of the licensing review process, provided that the incorporation in the SAR is without change or modification. However, new or revised regulations, Regulatory Guides, or Standard Review Plans affecting QA program cor.trols may be applied by the NRC to previously accepted QA programs. The status of NRC review of QA topical reports submitted by the principal contractors is shown in Table 1.
When design and construction activities were high, firms designing nuclear steam supply systems, architect engineering firms designing nuclear power plants, and certain selected major equipment vendors were inspected on a regular basis by NRC to ascertain through direct observation of selected activities whether these design firms and vendors were satisfactorily implementing the accepted QA program. However, with the substantial decline of new plant design activities, the inspection of QA program implementation has been deemphasized.
Instead, the NRC vendor inspection focus has been shifted to vendor activities associated with nuclear plant operation, maintenance, and modifications.
Inspection emphasis in now placed on the quality of the vendor products including hardware fabrication, licensee-iii
l vendor interfaces, environmental qualification of equipment, and equipment problems found during operation and corrective action.
If nonconformances with NRC requirements and regulations are found, the inspected organization is requested to take appropriate corrective action and to institute preventive measures to preclude recurrence.
If generic implications are identified, NRC assures that affected licensees are expeditiously informed.
In the past, NRC issued confirming letters to the principal contractors to indicate that NRC inspections have confirmed satisfactory implementation of the accepted QA programs. Licensees and applicants could, at their option, use the letters to fulfill their obligation under 10 CFR 50 Appendix B, Criterion VII, that requires them to perform initial source evaluation audits and subsequent periodic audits to verify QA program implementation. However, based on the above described change in nuclear plant design and construction activities, NRC will no longer issue confirming letters to principal contractors since future NRC vendor program inspections will focus on selected areas rather than addressing the implementation of their respective QA programs. Therefore, confirming letters that have already exceeded their three year effective period will not be renewed.
Confirming letters issued less than three years ago will remain in effect until the stated effective period expires.
Therefore, as the confirming letters expire, licensees and applicants will no longer be allowed to take credit for the NRC acceptance of the implementation of a principal contractor's QA program. Licensees continue to be responsible for the conduct of initial source evaluation audits and subsequent periodic audits to verify QA program implementation. The NRC Division of Quality Assurance, Vendor and Technical Training Center Programs will continue to review revisions to principal contractor QA programs when submitted and, when approved, will list the latest approved revision number and date of the approval letter in Table 1 of the next edition of the White Book.
The White Book will continue to be published and will contain copies of all vendor inspections issued during the calendar quarter specified. The vendor inspection reports list the nuclear facilities to which the results are applicable thereby informing licensees and vendors of potential problems.
In addition, the affected NRC Regional Offices are notified of any significant problem areas that may reouire special attention.
The White Book contains information normally used to establish a " qualified suppliers" list; however, the information contained in this document is not adequate nor is it intended to stand by itself as a basis for qualification of suppliers.
Correspondence with contractors and vendors relative to the inspection data contained in the White Book is placed in the USNRC Public Document Room, located in Washington, D.C.
Copier of the White Book may be obtained at a nominal cost by writing to the National Technical Information Service, Springfield, Virginia 22161.
iv
ORGANIZATION: COMPANY, DIVISION CITY, STATE l
REPORT INSPECTION INSPECTION NO.: Docket / Year / Sequence DATE:
DN-SITE HOURS:
CORRESPONDENCE ADDRESS:
Corporate Name Division ATTN:
Name/ Title Address City, State Zip Code ORGANIZATIONAL CONTACT:
Name/ Title TELEPHONE NUMBER:
Telephone Number NUCLEAR INDUSTRY ACTIVITY:
Description of type of components, equipment, or services supplied.
ASSIGNED INSPECTOR:
Name/ Vendor Program Branch Section Date OTHERINSFECTOR(S):
Name/ Vendor Program Branch Section APPROVED BY:
Name/ Chief - Section/ Vendor Program Branch Date INSPECTION BASES AND SCOPE:
A.
BASES: Pertain to the inspection criteria that are applicable to the activity being inspected; i.e.,10 CFR Part 21, Appendix B to 10 CFR Part 50 and Safety Analysis Report or Topical Report commitments.
B.
SCOPE: Summarizes the specific areas that were reviewed, and/or identi-fies plant systems, equipment or specific components that were inspected.
For reactive (identified problem) inspections, the scope summarizes the problem that caused the inspection to be performed.
PLANT SITE APPLICABILITY:
List plant name and docket numbers of licensed facilities for which equipment, services, or records were examined during the inspection.
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ORGANIZATION: ORGANIZATION CITY, STATE REPORT INSPECTION
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N0.:
RESULTS:
3 AGE 2 of 2 A.
VIOLATIONS:
Shown here are any inspection results determined to be in violation of Federal Regulations (such as 10 CFR Part 21) that are a'pplicable to the organization being inspected.
B.
NONCONFORMANCES:
Shown here are any inspection results determined to be in nonconformance with applicable commitments to NRC requirements.
In addition to identifying the applicable NRC requirements, the specific industry codes and standards, company QA manual sections, or operating procedures which are used to implement these commitments may be referenced.
C.
UNRESOLVED ITEMS: Shown here are inspection results about which more information is required in order to determine whether they are acceptable items or whether a violation or nonconformance may exist. Such items will be resolved during subsequent inspections.
D.
STATUS OF PREVIOUS INSPECTION FINDINGS: This section is used to identify the status of previously identified violations, items of nonconformance, and/or unresolved items until they are closed by appropriate action.
For all such items, and if closed, include a brief statement concerning action which closed the iten.
If this section is omitted, all previous inspection findirgs have been closed.
E.
INSPECTION FINDINGS AND OTHER COMMENTS: This section is used to provide significant information concerning the inspection areas identified under
" Inspection Scope."
Included are such items as mitigating circumstances concerning a violation or nonconformance, or statements concerning the limitations or depth of inspection (sample size, type of review performed and special circumstances or concerns identified for possible followup).
For reactive inspections, this section will be used to summarize the disposition or status of the condition of event which caused the inspection to be performed.
F.
PERSONS CONTACTED: Typed, Name, Title
- present during exit meeting SAMPLE PAGE (EXPLANATION OF FORMAT AND TERMIN0 LOGY) vi
TABLE 1 NRC APPROVAL STATUS OF QA PROGRAM TOPICAL REPORT REVISIONS i
DATE OF LATEST NRC i
PRINCIPAL TOPICAL REPORT REVISION REVISION APPROVAL LETTER 4
i Babcock & Wilcox BAW 10096A Revision 4 December 30, 1983 Bechtel BQ-TOP-1 Revision 3A August 28, 1984 i
Black & Veatch BVTR-1-D Revision 0A August 1, 1983 C. F. Braun 21A Amendment #5 July 16, 1980 3
i Brown & Root B&R-002A Revision 3 April 8,1980 1
1 Burns & Roe B& ROE-COM-1-NP Revision 4A March 4, 1986 Combustion Engineering CENPD-210-A Revision 3 October 16, 1984 Ebasco Services, Inc.
ETR-1001 Revision 12 May 4, 1984 Framatome FRA-QP/85 0782 NP Revision 2A March 14,1986 General Atomic GA-A13010A Amendment #8 October 15, 1984 4
General Electric Co.
NE00-11209-04A Revision 5 April 19,1985 j
Gibbs & Hill, Inc.
GIBSAR 17-A Amendment 8 February 27, 1985 Gilbert / Commonwealth GAI-TR-106 Revision 3 August 9, 1984 Ralph M' Parsons P-TOP-QA1 Revision 3A August 26, 1985 j
Sargent & Lundy Engineers SL-TR-1A Revision 6 April 14,1983 i
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Stone & Webster SWSQAP 1-74A Revision E February 6, 1986
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United Engineers &
Constructors UEC-TR-001 Revision 6 September 16, 1982 i
Westinghouse NTO WCAP-8370/7800 Rev. 10/6A August 29, 1984 l
vii.
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R W ASHING TON,0. C. 20555 A...../
SAMPLE QA TOPICAL REPORT REVISION APPROVAL LETTER (ADDRESSEE)
Gentlemen:
Your letter of transmitted Revision of your QA 4
i topical report Quality Assurance Program." The report describes your QA program for design, procurement, fabrication, and testing activities.
1 We have reviewed Revision of the report against the acceptance criteria in Section 17.1 of the NRC's Standard Review Plan for nuclear power plants i
(NUREG-0800, July 1981). Based on our review and evaluation of Revision we find that the criteria in Appendix B to 10 CFR Part 50 are met.
Revision of your QA topical report is, therefore, acceptable, and should be 1
implemented for safety-related applications. Our evaluation is enclosed.
i Should regulatory criteria or regulations change such that our conclusions about this topical report are invalidated, we will notify you. You will be given the opportunity to revise and resubmit it should you so desire. We note the commitment in the Foreward of the report to keep the NRC informed of changes to the QA program description.
4 Please incorporate this acceptance letter and its enclosed evaluation into the topical report, identify the report as Revision A(the"A" indicating that it has been found acceptable by the NRC), and transmit a copy to:
Document Control Desk i
ATTN:
Vendor Program Branch U.S. Nuclear Regulatory Commission Washington, D.C.
20555 By way of letters to the same address, please keep us informed of the nuclear units to which this QA topical report applies.
Should you have any questions regarding our review or if we can provide assistance, please contact us on i
(301)492-Sincerely, i
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Chief i
Vendor Program Branch Division of Quality Assurance, Vendor and Technical Training Center Programs.
I Office of Inspection and Enforcement i
i iX i
ORGANIZATION: ANCHOR DARLING VALVE COMPANY WILLIAMSPORT, PENNSYLVANIA REPORT INSPECTION INSPECTION N0.: 99900053/86-01 DATE:
1/6-10/86 DN-SITE HOURS:
96 CORRESPONDENCE ADDPESS: Anchor Darling Valve company 4
ATTN: Mr. A. E. Caron President 701 First Street i
Williamsport, Pennsylvania 17701 j
ORGANIZATIONAL CONTACT: Mr. George W. Knieser, QA Manager TELEPHONE NUMBER:
(717)327-4824 NUCLEAR INDUSTRY ACTIVITY: Approximately 35 percent of valve sales s
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ASSIGNED INSPECTOR:
T. Conway, lea ive Inspection Section (RIS)
Date OTHER INSPECTOR (S):
J. Harper, RIS L. Vaughan, Program Coordination Section
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APPROVED BY:
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E. J. Herschoff, Chi f,lm, Vendor Program Branch Date INSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR Part 50, Appendix B and 10 CFR Part 21.
B.
SCOPE:
This inspection was made as a result of the receipt of 10 CFR Part 50.55(e), Licensee Event, and 10 CFR Part 21 reports from:
(1)
Public Service Electric and Gas Company and Northeast Utilities on missing lock welds, (2) Washington Public Power Supply System and Cor,monwealth Edison on failure of air check valves to seat, (3) Long Island Lighting Company on components breaking off in swing check valves. and (4) Anchor Darling on unqualified terminal blocks.
PLANT SITE APPLICABILITY: Missing lock welds - Hope Creek (50-354), Millstone Unit 3(50-423); Air check valves - Byron Unit 1(50-454), WNP Units 1/4 (50-460/513); Swing check valves - Shoreham (50-322); Terminal blocks - seven plants.
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ORGANIZATION: ANCHOR DARLING VALVE COMPANY WILLIAMSPORT, PENNSYLVANIA i
REPORT INSPECTION NO.-
99900053/86-01 RESULTS:
PAGE 2 of 14 A.
VIOLATIONS:
Contrary to Section 21.31 of 10 CFR Part 21, a review of documentation packages for Section III nuclear valves revealed that while 10 CFR Part 21 was imposed upon Anchor Darling (AD) by their customers AD referenced Section III requirements but-did not specify that 10 CFR Part 21 requirements would apply on P0s C-432, C-433, A-1770, and A-1778 to Cann & Saul Steel; Y-2303 to Sandvik; Y-2489 and C-970 to Teledyne McKay; C-2348 to General Copper & Brass; A-3015 to R.E.C.;
A-1496 to Dodge Foundry & Machine; and A-664 to Precision Technology.
(86-01-01)
B.
NONCONFORMANCES:
1.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Sections 10.4.5.1 and 10.4.5.2 of the Quality Assurance Manual (QAM), there was no documented evidence of a Rework Ticket for two defects repaired on a Weld Repair Record dated September 4, 1985 for the globe body on Shop Order No. E-6534; and a Pepair Welding and Hardfacing Record dated October 31, 1985 did not contain a sketch for the disc on Shop Order No. E-6534.
(86-01-02) 2.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Section 6.5 of STD No. MQCS-2, a member from QA had not signed, initialed, i
or stamped the MRBA block on Material Rejection Notice (MRN)
No. 9419 dated April 3,1985, relating to the bonnet on Shop Order No. E-6534-001.
(86-01-03) 3.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 Section 5.3.4 of STD No. QAS-9, and Section 9.6.1 of SNT-TC-1A, a review of qualification records for seven nondestructive examination (NDE) personnel (one-Level III and six-level II) revealed that the qualification records did not contain a statement indicating satisfactory completion of training in accordance with STD No.
QAS-9.
(86-01-04) 4.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 Section 9.4.4 of the QAM, and Section 3.0 of STD No. MPDS-7, there was no documented evidence that a file of Final Approved Vendor Procedures existed for Section III valves on Shop Order Nos. E-6534, E-6516, and E-3256, and the file for E-3326 did not contain a copy of Cann &
Saul heat treat Procedure No. 1023, Revision 4 which was used on the bonnet material.
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ORGANIZATION: ANCHOR DARLING VALVE COMPANY WILLIAMSPORT, PENNSYLVANIA REPORT INSPECTION NO - 99900053/86-01 RESULTS:
PAGE 3 of 14 Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Sections 5.
5.4.2, 5.4.3, and 5.4.3.1 of STD No. QAS-10, during an inspection of the shop and the calibration laboratory, it was noted that two spring gages (S/N 0154 and 0395) were not labled properly to indicate that they had been calibrated; and the balance scales, surface plate, and dead weight tester were not correctly labled to indicate that they had been taken out of service.
(86-01-06) 6.
Contrary to Criterion V of Appendix B to 10 CFR Part 50, Section 4.4.4.1 of the QAM, and Section 5.1.1 of STO No. NQCS-2, during an inspection of the shop, it was noted that acceptable seat rings (i.e., 24" - 900 HT E6685-1, 14" - 300 HT P8023-122, 16" - 300 HT P8023-144, and 28" - 150 HT P8157-1) were in a nonconforming material hold area with nonconforming material.
(86-01-07) 4 7.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and STD No, M-240, it was noted that dedicated stainless steel burr wheels located in a cabinet for grinding equipment were not marked in yellow.
(86-01-08) 8.
Contrary to Criterion V of Appendix B to 10 CFR Part 50, Section 10.4.4.3 of the QAM, and Sections 5.4, 5.2.2, and 5.2.3 of STD No.
EMS-1 during an inspection of the manufacturing)and welding material storage area, it was noted:
(86-01-09 An unmarked box of Sandvik electrodes (E308L-16 in three a.
different sizes) was left in the weld area after the completion of a job.
b.
Unmarked stellite rods were not in containers in the welding material storage area.
Bare wire remaining on a wire feeder after usage was not c.
covered.
d.
The temperature in an oven containing E309L electrodes (Lot Nos. 12039-1-1and90194-1-3) was below 225 F on two occasions (150 F in the morning and 180 F in the afternoon of the same day).
9.
Contrary to Criterion V of Appendix B to 10 CFR Part 50, Sections 9.4.2.5 and 9.4.2.1 of the QAM, and Sections 4.4.5 and 4.2.7 of STD No. MQCS-2, a review of the Approved Vendor List (AVL) dated January 6, 1986 and six vendor audit reports for Cann & Saul Steel, Quaker Alloy Coating, Dodge Foundry & Machine, Effort Foundry, R.E.C., and LTV Steel indicated the following:
(86-01-10) 3
ORGANIZATION: ANCHOR DARLING VALVE COMPANY WILLIAMSPORT, PENNSYLVANIA REPORT INSPECTION NO.-
99900053/86-01 RESULTS:
PAGE 4 of 14 l
a.
Vendor corrective action as indicated on VMDRs was not i
audited for implementation by Quality Assurance.
b.
The restriction (i.e., Effort Foundry be re-audited prior to acceptante of any Category 1 material), resulting from the January 1985 vendor audit was not identified in the AVL.
c.
R.E.C. located in Mt. Vernon, NY was not audited within the twelve month frequency.
- 10. Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Section 6.5.4 of STD No. MQCS-2, a review of the QA Department MRN Log Book and purchasing vendor files for 373 VMDRs issued during 1985, noted that 19 were still open having passed the ten working day requirement for a response.
(86-01-11)
- 11. Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Section 1.6 of the QAM, a review of training and indoctrination records for manufacturing personnel indicated no objective evidence of training on the QAM revisions dated November 1, 1984 and December 3, 1985 for the Manufacturing Manager, Engineer, and Planner; Assembly Department Supervisor; Store Supervisor; and Machine Foreman.
(86-01-12) 12.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Section 6.7 of Procedure No. OPER-1, a review of Tech Staff files identified 17 Abnormal Occurrence Reports (AOR) issued in 1985, but only three AORs (Nos. ABW-85-7, ABW-85-10, and ABW-85-14) had an evaluation performed and documented on Form OPER-1-2.
(86-01-13)
C.
UNRESOLVED ITEMS:
None.
D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
1.
(Closed) Nonconformance (Item 8, 84-01): AD, when notified by Namco Controls that certain Model EA-180 limit switches had been assembled with the wrong cover gasket material, did not identify the discrepant gasket material on a Material Rejection Notice, describe the corrective action to prevent recurrence and disposition of the discrepant gasket material, or secure approval of the material review board.
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ORGANIZATION: ANCHOR DARLING VALVE COMPANY WILLIAMSPORT, PENNSYLVANIA INSPECTION REPORT RESULTS:
PAGE 5 of 14 NO.-
99903053/86-01 The NRC inspector verified that AD had notified their customer (i.e.,
Bechtel for Susquetanna, SNUPPS, and Palo Verde nuclear facilities) on Document No.i. CAW-7458, -7453', and -7457 dated February 18, 1980 that they were supplied limit switches with the wrong gasket material.
E.
OTHER FINDINGS OR COMMENTS:
1.
Cracked / Missing Tack Welds During installation of AD swing check valves at Falo Verde in December 1984, it was discovered that tack welds were aissing on the set screws in the hinge support on several valves.,Since that time similar tack weld problems (e.g., cracks / missing welds) have occurred on:
hinge pin bushings at Millstone Unit No. 3 (ref.
Fortheast Utilities letter dated April 25,1985) and at Summer (ref.
morning report dated October 11,1985), and hinge support pieces at Shoreham (ref. morning report dated November 7, 1985).
The NRC inspector examined AD assembly drawings for manufacturing swing check valves. Tack welding was called out on capscrews, hinge supports-and set screws.
The design of the bushings was changed such that tack welding is no longer required, and the disc nut was called to have a stainless steel pin inserted (rather than tack welding) and peened at both ends to prevent backing off.
AD also established a new procedure, No. EPS-159 " Inspection of Valve Assemblies" dated Decembcr 1985.
In the shop area, it was noted that EPS-159 forms were being used.
These forms assure that when tack welding is required, it is completed and inspected by QA. AD feels tnat the design change and Procedure No. EPS-159 will prevent the occurrence of tack welding problems in the future.
2.
Air Check Valves u
The NRC was notified in March 1985 by Commonwealth Edison (Bryon) in a 10 CFR 21 report and ir, June 1985 by Washington Public Power Supply System (WNP Unit Nos.1 & 4) in a 10 CFR.50.55(e) report of air check valves failing to seat properly. The valves, which were m'nufactured a
by Parker-Hannifin (PH) and supplied by 1.0, may degrade the capability for closing mainstream isolation valves or feedwater isolation valves.
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ORGANIZATION: ANCHOR DARLING VALVE COMPANY WILLIAMSPORT, PENNSYLVANIA REPORT INSPECTION NO.-
99900053/86-01 RESULTS:
PAGE 6 of 14 AD notified the NRC of 11 other plants to which similar valves were supplied, and the NRC issued Information Notice No. 85-35 in April 30, 1985. The NRC inspector reviewed nine letters sent to five architect / engineers and four licensees in which AD indicated that testing of the pnuematic check valves by both PH and AD determined that the valves seal reliably under a sudden (guillotine) loss of air supply but do not seal reliably under a gradual loss 'of air supply. AD recommended that an alternate pneumatic check valve (Model No. 8F-C8L-10-55) be used for a slow decay of air supply.
The new valve was tested by PH (ref. March 28, 1985 letter from PH to AD) and qualified to IEEE Standards 323, 344, and 382 (ref.
Environmental Qualification Report dated October 1985).
It was noted that AD sent replacement valves to eight of the affected plants. As of the date of the inspection, South Carolina Electric and Gas (Summer), Louisiana Power and Light (Waterford Unit No. 3),.
and Arizona Public Service (Palo Verde Unit Nos.1, 2, & 3) had not requested replacement valves from AD.
3.
Terminal Blocks On September 5,1985, AD notified the NRC of a potential problem related to terminal blocks supplied as spare parts. The NU2 terminal blocks are manufactured by Connectron out of nylon and polysulfone materials, and only the polysulfone material has been IEEE qualified for application in terminal boxes furnished with AD hydraulic actuators installed on safety related valves. The NRC inspector reviewed AD P0 No. C-2953 dated September 19, 1985 to Connectron for terminal blocks, and it was noted that the P0 did not specify that the material should have been polysulfone.
To prevent recurrence, AD has assigned a code number for the procurement of spare (replacement) NU2 terminal blocks made of polysulfone.
It was also noted that AD notified all the affected customers of this potential problem. The customers included Comonwealth Edison (Byron Unit'Nos.1 & 2 & Braidwood Unit Nos.
1 & 2), Union Electric (Callaway), Kansas Gas and Electric (Wolf Creek), Washington Public Power Supply (WNP Unit Nos. 1 & 3), and South Carolina Electric and Gas (Summer).
4.
Check Valve Failures In November 1985, Long Island Lighting Company (LILCo) described the failure of two steam exhaust line check valves at Shoreham in Licensee Event Report No. 2636. Upon disassembling the valves, l
a bolt and flapper from upstream valve no. 2 were lodged in valve 6
ORGANIZATION: ANCHOR DARLING VALVE COMPANY WILLIAMSPORT, PENNSYLVANIA REPORT INSPECTION NO.-
99900053/86-01 RESULTS:
PAGE 7 of 14 no.1, and a flapper from valve no. 3 located further upstream was found lodged in the internals of valve no. 2.
In July 1982, the NRC issued Informt. tion Notice No. 82-26 pertaining to check valve failures in which, similar to the Shoreham failures, valve components separated and eventually lodged in downstream piping.
With regards to pump turbine exhaust check valves. AD investigated operating conditions to determine the cause of damage to the valves.
It was determined that two forms of periodic testing are potential modes of failure. When the pump turbine is subjected to a fast start, the disc slams into the full open position. Pepeated impact of the disc against the body puts a strain on the hinge assembly and leads to eventual hinge failure. When the pump turbine throttle is partially opened, it causes the disc to flutter on the seat, and this cyclic action causes cracking in both disc and seat ring sealing surfaces. The results of an AC research program to overcome these failyres led to a specia'ily designed lift check valve utilizing a light reight disc and 't dual seat design (primary seat of stellite and secondary seat of a radiation resistant polymer) incorporated in the disc.
In respense to an inquiry, AD notified Stone & Webster in January 1982 of the special lift check valve and enclosed a copy of the informaticn theet entitled " Pump Turbine Exhaust Check Valves."
7he NRC inspector was told that all AD service representatives had copies of this infomation sheet far distribution to their customers. The information sheet dccuments AD findings and recommendations with regards to swing check valve failures.
In an August 1983 letter to Ebasco, A0 recommended that the swing check valves be replaced by l'if t check valves at Waterford Unit No. 3 as had previously been done by Arkansas Power & Light. The NRC inspector was also told that on a verbal recommendation by AD, LILCo purchased lift check valves as replacements for the swing check valves in 1982, but as far as AD could determine, they had not been installed.
5.
Control _ofHeasuringandTestEquipment(E&TEl The NRC inspector reviewed applicable proceduret and calibration status cards on selected M&TE to determine the adequacy and implementation of the calibration control system. The following M&TE were checked for current calibration:
calibration wires, l
charpy impact tester, gage blockt., gas M!1 ding eeg91ator.s, hardness tester, weld rod ovens, micrometers, spring scale, tensile tester, 7
ORGANIZATION: ANCHOR DARLING VALVE COMPANY WILLIAM 5?0RT, PENNSYLVANIA REPORT INSPECTION NO.
__99900053/86-01 RESOLTS:
PAGE 8 of 14 torque wrenches, vernier calipers, depth gages, hydraulic pressure gages, surface plate. pressure gage comparator, and thread plug gages.
All calibrations that were current were verified to be traceable to the National Bureau of Standards.
Personally owned M&TE such as micrometers were verified to have been calibrated by the prescribed procedure.
It was noted that calibration labels were slot being used as intended for thread plug gages which are to be calibrated after 25 uses as documented on a calibration schedule sheet (Form QC-77-1).
Instead of changing all the calibration labels on all of the thread plugs to reflect the currently used calibration schedule (after evey 25 uses),
AD has only changed the labels as they use the thread plug gages.
The other example of calibration labels not being used properly was with equipment that had been taken out of the calibration cycle.
The balance scale, surface plate and dead weight tester are no longer used and were taken out of the calibration cycle. Yet, they have not been labeled "Do Not Use." Upon examination of the spring gages in the shop area, the calibration label did not indicate the correct date of the last calibration and the calibration due date.
(see Nonconformance 86-01-06) 1 6.
C_ontrol of Nonconforming Material During an inspection of the manufacturing area the NRC inspector noted that the nonconforming hold area contained seat rings (24" - 900 heat E6685-1, 14" - 30.0 HT P8023-122, 16" - 300 HT J
P8023-144, 28" - 150 HT P8157-1) that were not identified with nonconforming red hold tags. The acceptaH e seat rings were not segregated from the other nonconforming materials in the area.
In addition, the boundaries of the hold area were not obvious as the hold area was immediately adjacent to an area where acceptable material was present.
(see Nonconformance 86-01-07) l 7.
Control of Special Processes The NRC inspector examined the welding shop area and weld metal dispatcher room.
It was noted that one-half empty box of three different sizes of welding electrodes was found in a welding area l
where welding was not taken place.
Uncovered bare wire was also i
noted on a welding spool at a welding station.
In the weld dispatch room a few bare unmarked stellite rods were left out of their original box on the shelf.
r
ORGANIZATION: ANCHOR DARLING VALVE COMPANY WILLIAMSPORT, PENNSYLVANIA REPORT INSPECTION NO.-
99900053/86-01 RESULTS:
PAGE 9 of 14 An examination of welding rod ovens in the weld material dispatch room indicated that five of six ovens were maintained at the proper temperature. The oven with E309L 5/32" dia (Lot Nos. 12039-1-1 and 90194-1-3) electrodes was operating at approximately 150 F when the NRC inspector checked it at approximately 10:30 am. At approximately 2:00 pm, the same oven was operating at approximately 180 F.
Upon questioning the authorized welding metal dispatcher about the oven temperature, he acknowledged to the QA Manager and the NRC inspector the incorrect temperature reading. With regards to the oven having the low temperature readings, the QA Manager questioned the Welding Engineer who commented that he was aware of the problem, and repairs were scheduled to be made.
(see Nonconformance 86-01-09)
While inspecting the weld area, the NRC inspector observed a case containing burr wheels stored in a stainless steel cabinet with other tools dedicated for stainless steel grinding. The burr wheels were not appropriately marked in yellow.
(see Nonconformance 86-01-08)
The NRC inspector reviewed a document entitled " Record of Welding Qualifications" which is authorized by the Welding Engineer on a monthly basis. This document identifies a welder to a Welding Procedure Specification (WPS) which is cross referenced to an "M" No. which applies to the welder's performance qualifications.
The testing of qualification test specimens for Procedure Qualification Records (PQR) and welder qualification tests are performed by a technician who also does the internal calibration of M&TE. The Welding Engineer, who is also the Inspection Supervisor, signs off the PQRs and welder qualification tests.
Welder performance qualifications for five welders (Nos. 6, 22, 33, 44, and 77) were reviewed against six WPS (including applicable PQRs) to which they had welded.
8.
NDE The NRC inspector reviewed the qualification and certification records of NDE personnel (one-Level III and 11-Level II) to determine whether the individuals performing NDE were certified to SNT-TC-1A. The written practice (STD No. QAS-9) of AD for all phases of training and certifying NDE personnel was also reviewed, and it appeared to be consistent with SNT-TC-1A. With the exception of a missing statement pertaining to training in accordance with
ORGANIZATION: ANCHOR DARLING VALVE COMPANY WILLIAMSPORT, PENNSYLVANIA REPORT INSPECTION NO.-
99900053/86-01 RESULTS:
PAGE 10 of 14 STD No. QAS-9, the records including eye examinations, training, written tests and certifications appeared to satisfy the require-ments of SNT-TC-1A.
(see Nonconformance 86-01-04)
Ar. evaluation of activities at four NDE stations was undertaken to assure that current procedures are being used and all equipment is calibrated. The stations observed were radiography (RT), liquid penetrant (PT), magnetic particle (MT)-wet, and the upgrading station which is MT-dry. All the NDE procedures reviewed at each station were current with the master document retained in the Engineering file.
At the MT station (wet), it was noted by calibration stickers that MT Unit URESCO No. 3505 (S/N 810679) and ultraviolet light No. J-221 were serviced and calibrated by Nutley Equipment on October 21, 1985.
The calibration report from Nutley was reviewed and found acceptable.
Records also indicated that an intensity check of the light was done daily, and a check of the chemical bath was performed twice a shift.
At the upgrading station, the MT Unit No. CRO-10 (S/N 751312) and powder blower No. XB-2A (S/N 76101) were also serviced and calibrated by Nutley Equipment on October 21, 1985. The densitometer (S/N 1402B) at the RT station is calibrated against a density strip No. 762 on a monthly basis as noted on a chart attached to the instrument.
However, it was noted a Density Strip Calibration Report from Dupont for strip No. 762 was dated August 22, 1979, and the reference standard (strip) had not been calibrated since 1979.
9.
Corrective Action The NRC inspector reviewed applicable sections of the QAM, Procedure No. MQCS-2 and the QA department MRN log book. The procedure allows l
purchasing 10 working days to completely process a Vendor Material Discrepancy Report (VMDR) and receive a response from the vendor.
AD issued 373 VMDRs during 1985. At the time of this inspection, the MRN log book listed 39 incomplete VMDRs. Of the 39 incomplete VMDRs, a second letter was sent for 37 VMDRs, and telephone calls were made for eight requesting vendor action / response for the second and third time. A review of purchasing files and the vendor's folder was also conducted, which resulted in the locating of 19 complete VMDRs.
However, as of January 9, 1986, 19 VMDRs were still open pass the required ten working days completion date. The open VMDRs are 9697, 9967, 9972, 10129, 10130, 10171, 10207, 10227, 10229, 10230, 10248, 10258, 10269, 10239, 10279, 10280, 10281, 10282, 10284.
(see Nonconformance 86-01-11) l 10 l
ORGANIZATION: ANCHOR DARLING VALVE COMPANY WILLIAMSPORT, PENNSYLVANIA REPORT INSPECTION NO - 99900053/86-01 RESilLTS:
PAGE 11 of 14 10.
External Audits The NRC inspector reviewed applicable sections of the QAM, one procedure, ten vendor audit reports, two vendor folders, four P0s, and the AVL to verify vender audits and evaluations are being performed as required.
A review of vendor audits conducted in 1984 and compared to those conducted in 1985 noted that a standard checklist is used to conduct vendor audits, but the checklist has not been revised to include verifying the implementation of vendor correction actions as stated in VMDRs.
It was noted that the last audit conducted by AD on R.E.C. was October 1982, at which time R.E.C. located in New Rochelle, NY was an ASME Certificate holder. The company moved from New Rochelle, NY to Mt. Vernon, NY betweeit January 1984 and November 1984.
AD was aware of the move, but had no documented evidence of when the move.
actually took place.
Four P0s(Nos. 610, A-3015, C-922 and A-2875) were issued between November 1984 and April 1985 to R.E.C., Mt. Vernon, NY for the procurement of Category 1 material (material identified as traceable and pressure retaining). However, the AVL identified R.E.C.
as a Category 2 supplier (requires identifying material to specification and grade only). A minimum of 12 months had expired since R.E.C. made the move to Mt. Vernon, NY. Since an ASME Certificate is nontransferable from one location to another, the certificate issued in New Rochelle was void in Mt. Vernon.
R.E.C.
also informed AD in a letter dated October 18, 1985 that they would not be renewing their ASME certificate.
During the inspector's examination of the vendor audit reports and the AVL dated January 6, 1986, it was noted that a restriction was placed on Effort Foundry Co., Bath, PA during the audit conducted in January 1985.
The note called for Effort Foundry to be re audited prior to the acceptance of any Category 1 material. However, his restriction was not listed in the AVL dated January 6, 1986, and Effort Foundry was listed as a supplier of Category 1 material.
(see Nonconformance 86-01-10)
- 11. Training and Indoctrination Tho NRC inspector reviewed applicable sections of the QAM and training and indoctrination records for seven auditors, Purchasing Manager, two buyers, six QA inspectors, four Project Engineers, Manager of Contract Administration, two Contract Administration personnel, Engineering Manager, Manufacturing Manager, Manufacturing Planner, Manufacturing Engineer, Upgrading Manager, Assembly Department l
1 11 l
ORGANIZATION: ANCHOR DARLING VALVE COMPANY WILLIAMSPORT, PENNSYLVANIA REPORT INSPECTION NO.-
99900053/86-01 RESULTS:
PAGE 12 of 14 Supervisor, Store Supervisor and the Machine Foreman. AD could not provide documented evidence that the Manufacturing Manager, Manufacturing Engineer, Manufacturing Planner, Assembly Department Supervisor, Store Supervisor and the Machine Foreman had received indoctrination and training of QAM revisions issued November 1, 1984 and December 3, 1985.
(see Nonconformance 86-01-12) 12.
10 CFR Part 21 The NRC inspector reviewed applicable sections of the QAM, one procedure, and 17 Abnormal Occurrence Reports (A0Rs) issued during 1985. The reports (Nns. ABW-85/1-17) were documented on form OPER-1.
Each A0R is required to be investigated and evaluated to determine whether it is " substantial to safety," "not substantial to safety" or " undeterminable." The evaluations are also required to be documented on form OPER-1-2 and attached to form OPER-1. Of the 17 A0Rs examined, only three (listed below) had been evaluated and documented on form OPER-1-2 and attached to form OPER-1.
(see Nonconformance 86-01-13)
AOR (date)
Item ABW-85-7 (April 27, 1985)
Operator failed and overtorqued the valve bending the discs ABW-85-10(October 17,1985)
Hydraulic actuator accumulators could not be charged hydraulically ABW-85-14 (November 6, 1985)
Disc nut not pinned and hinge supports, hinge support capscrew, and set screws not tack welded It was noted that copies of Section 206 of the Energy Reorganization Act of 1974, 10 CFR Part 21, STD No. CGS-5 " Method for Employees to Report Discrepant Conditions," a'nd Procedure No. OPER-1 " Notification and Resolution of Abnormal Occurrences Report" were posted in the personnel department office. A review of documentation packages which included AD P0s to vendors for components used in the manufacture of safety-related nuclear valves revealed that the
-l requirements of 10 CFR Part 21 were not referenced or identified on the following P0s which all referenced Section III of the ASME Code:
(see Violation 86-01-01) l 12 l
ORGANIZATION: ANCHOR DARLING VALVE COMPANY WILLIAMSPORT, PENNSYLVANIA REPORT INSPECTION NO.-
99900053/86-01 RESULTS:
PAGE 13 of 14 No. (Date)
Vendor Y2303 (September 19,1983)
Sandvik Y2489 (October 12,1983)
Teledyne McKay C970 (March 12,1985)
Teledyne McKay C432(January 23,1985)
Cann & Saul C2348 (July 31, 1985)
General Copper & Brass C433 (January 23,1985)
Cann & Saul A3015 (November 16,1984)
R.E.C.
A1496 (May 29, 1984)
Dodge Foundry & Machine A1778 (June 22, 1984)
Cann & Saul A1770 (June 21, 1984)
Cann & Saul A664 (February 22,1984)
Precision Technology The QA Manager indicated that AD did not impose Part 21 requirements on any P0 prior to January 1985.
- 13. Documentation Packages (DP)
The NRC inspector reviewed DPs (Shoo Order Nos. E-6534, -3326, -5836,
-6516, and -3256) for nuclear valves ordered by Florida Power & Light (Turkey Point Unit Nos. 3 & 4), Bechtel (South Texas), Stone & Webster (Shoreham), Southern California Edison (SONGS), and Pacific Gas &
Electric (Diablo Canyon). DPs consisted of customer P0, Shop Order Instructions (501), Material / Drawing List; drawings; Production Release Tickets (PRT); weld repair records; Material Rejection Notices (MRN); Inspection Records (IR) for components (e.g., body, bonnet and disc); test data; P0s to manufacturers / suppliers; and Certified Material Test Reports (CMTR), Certificate of Conformances, and heat treat charts from manufacturers / suppliers.
All the customer P0s referenced 10 CFR Part 21 and Section III/ Class 1, 2, or 3 of the ASME Code. The 50I identified the customer data, AD 13
ORGANIZATION: ANCHOR DARLING VALVE COMPANY WILLIAMSPORT, PENNSYLVANIA REPORT INSPECTION NO.-
99900053/86-01 RESULTS:
PAGE 14 of 14 data (valve list and procedures), components to be procured, and manufacturing and QA requirements. The PRT identified all activities (e.g., machining, welding, inspection) to be performed on all the components including valve assembly, test, cleaning and packaging.
The activities on the PRT were initiated / stamped and dated by an operator or inspector. A detailed review of weld repair records and MRNs indicated that on Shop Order No. E-6534 two defects were repaired, but there was no documented evidence of Rework Ticket (see Nonconformance 86-01-02), and a member from QA had not signed off on MRN No. 9469 (see Nonconformance 86-01-03).
Although MT and PT examinations are documented on an IR with an inspector's initial and a reference to a procedure, there is no specific report identifying the equipment and/or material used or documenting the test results of NDE. A review of CMTRs from manufacturers indicated that two CMTRs from Cann & Saul did not reference that the item was manufactured in accordance with Section III of the ASME Code. One CMTR dated March 29, 1985 was for disk material for Class 1 and Class 2 valves, and the other CMTR was dated August 29, 1985 for bonnet material for Class 3 valves.
F.
PERSONS CONTACTED:
- J. Chappell, Manager-Engineering
- G. Knieser, OA Manager
- W.
G. Knecht, Technical Director L. Snyder, QA Engineer B. Stannert, QA Engineer D. Wright, QA Engineer T. Wolf, Welding Engineer S. Moon, Welding Material Dispatcher M. Miller, Purchasing Manager C. Gordon, Buyer D. Murray, Buyer H. Patterson, Manager of Contract Administration J. Wingate, Manufacturing Manager F. Bensinger, Product Specialist
- Attended Exit Meeting 14
i ORGANIZATION:
COOPER ENERGY SERVICES MOUNT VERNON, OHIO REPORT INSPECTION INSPECTION N0.-
99900373/85-01 DATE:
10/21-23/85 DN-SITE HOURS:
36 CORRESPONDENCE ADDRESS:
Cooper Energy Services En-Tronic Controls Division i
ATTN: Mr. H. D. Lenz, Manager, Engineering North Sandusky Street Mount Vernon, Ohio 43050 ORGANIZATIONAL CONTACT: Gary W. Mizer, Manager of QA and Technical Services TEtFPHONE NUMRER-(614) 393-8448 NUCLEAR INDUSTRY ACTIVITY:
Engine, compressor and pipeline controls.
Parts and maintenance service only; no current nuclear orders.
ASSIGNED INSPECTOR:
'l8/;,k[6 E. H. Trottier, Reactive Inspection Section (RIS)
Date OTHER INSPECTOR (S):
W. P. Haass, Pro rdination Section APPROVED BY:
/
If paf4 E. W.~Merschoff, Chie RIS, Vendor Program Branch Date INSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR Part 50 Appendix B and 10 CFR Part 21 B.
SCOPE: This inspection was performed to establish the cause of an erroneous engine speed signal that would have prevented a Cooper Energy Services standby diesel generator from starting when required.
PLANT SITE APPLICABILITY:
Byron 1/2 (50-454, 455); Nine Mile Point 2(50-410);
Palo Verde 1/2/3 (50-528, 529, 530).
15 a
i
i l
ORGANIZATION:
COOPER ENERGY SERVICES MOUNT VERNON, OHIO REPORT INSPECTION NO.-
99900373/85-01 RESULTS:
PAGE 2 of 5 A.
Inspection Issues The issue that resulted in this inspection was erroneous RPM indication experienced by a number of standby diesel generators supplied by Cooper Energy Services.
In addition to incorrect display of engine speed, a faulty RPM signal can prevent the diesel generator from starting when required. During a normal start sequence, the starting air supply is removed once the engine reaches approximately 275 RPM.
Thus, with a faulty RPM signal registering engine speed above 275 RPM, no starting air would reach the engine.
In addition, there are sundry other alarm and control functions initiated by engine RPM that would also be disabled.
B.
Inspection Findings Contrary to Criterion XVIII of Appendix B to 10 CFR 50, and Section 18.3 of the En-Tronics Division QA Manual, no records could be found to verify that internal audits had been conducted in 1983.
C.
Other Observations and Comments 1.
Faulty Engine Speed Indication The inspector examined all available documents and conducted several staff interviews to establish En-Tronics part in correcting the erroneous RPM indication that has occurred over the past several years at three nuclear power plants.
l Early correspondence between the Airpax Corporation (subcontract supplier of the engine tachometer) and En-Tronic Controls indicates that the problem was~ initially thought to be in the 21 Vdc power supply for the tachometer.
Specifically, it was believed necessary to completely isolate the power supply unit from " noise" (a small, but undesirable ac voltage component superimposed on the de supply) on the incoming plant 125 Vdc supply by adding a dc to dc converter.
In their letter to En-Tronic Contrdis dated November 9,1983, Airpax Corporation stated that such de to dc conversion for purpose of isolation was not required in the application used by En-Tronic Controls (diesel engine tachometers). Airpax did, however, recommend a change in the ground potential of the tachometer case, such that it be tied to the same potential as the negative side of the incoming 125 Vdc plant power source.
Continued pursuit of the technical solution to this problem by En-Tronic Controls and the affected utilities reveals the following:
16
ORGANIZATION: COOPER ENERGY SERVfCES MOUNT VERNON, OHI0 REPORT INSPECTION NO.-
99900373/85-01 RESULTS:
PAGE 3 of 5 Arizona Public Service, Palo Verde Station, implemented a plant change request that changed the ground potential of their Airpax tachometers to that of the incoming plant.125 Vdc. This modification was shown to be ineffective when the incoming plant 125 Vdc supply was found to have approximately 30 Vac superimposed on it causing
...the Diesel Generator logic to think the engine is starting...."
On August 3, 1985, Palo Verde Station approved a second modification to reduce the amount of noise on the incoming de supply from 30 Vac to less than or equal to 6 Vac. This modification was based on a laboratory test conducted by En-Tronic Controls that demonstrated the system would tolerate an incoming noise signal of at least 15 Vac without causing diesel generator " engine started" signals to appear.
The problem of false diesel generator engine start signals has not occurred at Palo Verde Nuclear Generating Station since this plant change was implemented.
In September, 1984 and July, 1985, Commonwealth Edison's Byron Station standby diesel generators displayed engine RPM indication with the engines shut down, as well as a number of local and control room alarms. Since indicated engine speed was above the starting air cut-out speed (280 RPM), any automatic start signal received would fail to " roll" and start the engines. Analysis by station and corporate personnel revealed a faulty annunciator power supply that was imposing a high frequency ac voltage on the common 125 Vdc supply to the diesel engine control cabinets. This problem was remedied by the installation of de to de converters that effectively separate the incoming 125 Vdc supply from the engine tachometer. No further problems of this type have been reported to date.
Unit 2 at Nine Mile Point has also experienced spurious RPM indication on their standby diesel generators.
Station personnel, believing the problem to be a ground fault in the 125 Vdc supply, ran 1
separate, shielded power supply cables to the engine control panels.
As a precaution, a dc to dc converter was installed to provide additional isolation and prevent unwanted ac from appearing on the de supply to the control panels.
The inspectors noted that while the problems at all three sites were related (faulty 125 Vdc supply causing erroneous engine RPM indication), neither the cause nor cure was consistent. Also, it was learned that in all the standby or on-line diesel engine power systems sold by Cooper Energy Services and En-Tronic Controls, only nuclear standby diesel generator installations seemed to be affected.
17
ORGANIZATION: COOPER ENERGY SERVICES MOUNT VERNON, OHIO REPORT INSPECTION NO.-
99900373/85-01 RESULTS:
PAGE 4 of 5 2.
Internal Audit Program With the exception of the item of nonconformance noted above, the inspector found the En-Tronic Controls Division audit program to be adequate.
Corrective actions to previous audit findings were appropriate.
3.
Procurement Document Control The inspector reviewed applicable sections of the En-Tronic Controls Division QA manual to determine the requirements for vendors to provide En-Tronic with quality products and services.
In addition, the inspector reviewed procurement document files and the report of an audit conducted by an En-Tronic Controls customer. Of the three procurement document files examined, only one was found to indicate that the item to be purchased by En-Tronic was evaluated for its safety significance.
In the remaining two files, no evidence could be found to establish that an evaluation was conducted to establish an appropriate safety classification for the parts supplied. No objective evidence was presented to the inspectors to show that suppliers were providing the appropriate level of quality in their products and services, nor does En-tronic maintain a historical record of vendor general performance (approved vendor list).
- Further, it was not clear that En-Tronic Controls performed any inspection of components on delivery.
Rather, it was left to field failures to point to components that were not adequate for the application.
4.
Design Documentation j
Three nuclear diesel generator design packages were reviewed. The packages included the bid (technical) specifications, drawings, procurement documentation, change notices and miscellaneous letter transmittals between parties to the contract. One significant design change was found and the attendant documentation reviewed.
The change resulted from the addition of high wattage diodes to the engine control panel and required extra ventilation to control the panel's interior temperature.
Review and approval of the design change appeared to be in order.
l i
18
ORGANIZAT10N:
COOPER ENERGY SERVICES MOUNT VERNON, OHIO REPORT INSPECTION NO.-
99900373/85-01 RESULTS:
PAGE 5 of 5 5.
Procedures Internal En-Tronic procedures were reviewed to determine if procedural controls were adequate. The procedures addressed design checklists, quality assurance designations for material, material purchase control, field and shop change orders, engineering responsibilities, and draftsman responsibilities. All procedures appeared to be adequate in content.
l l
19
ORGANIZATION: CORPORATE CONSULTING & DEVELOPMENT COMPANY, LTD RESEARCH TRIANGLE PARK, NORTH CAROLINA REPORT INSPECTION INSPECTION N0.: 99900511/85-02 DATE(S):
12/9-13/85 ON-SITE HOURS:
105 CORRESPONDENCE ADDRESS: Corporate Consulting
& Development Company, Ltd.
ATTN: Dr. J. R. Yow President P. O. Box 12728 Research Triangle Park, NC 27709 ORGANIZATIONAL CONTACT: Mr. Carson Blanton, Jr., QA Manager TELEPHONE NUMBER:
(919) 362-8800 PRINCIPAL PRODUCT: Engineering, consulting, and testing services.
NUCLEAR INDUSTRY ACTIVITY: Corporate Consulting and Development Company, Ltd.
(CCL) provides engineering consulting and testing services to the nuclear industry for seismic analysis, testing, and nuclear environmental qualifications of equipment.
ASSIGNED INSPECTOR:
Odd / J'J. J'2'702iI
/L 13-8/o Jt. N. Moist, Equipm6nt Qualification Inspection Date Section (EQIS)
OTHER INSPECTOR (S):
R. Lasky, EQIS J. Grossman, Sandia National Laboratories
(
El-M Ab M A,6----
APPROVED BY:
U. Potapovs, Chief, EQIS, Vendor Program Branch Date INSPECTION BASES AND SCOPE:
A.
BASES: Appendix B to 10 CFR Part 50 and 10 CFR Part 21.
B.
SCOPE: This inspection consisted of:
(1) a technical evaluation of equipment qualification (EQ) test activities for safety-related equipment, (2) witnessing of inprocess EQ testing, and (3) verfication of implemen-tation of the quality assurance (QA) program, b
PLANT SITE APPLICABILITY: James A. Fitzpatrick (50-333), Cooper (50-298),
Calvert Cliffs 1 & 2 (50-317 and 50-318), Callaway (50-483).
21
ORGANIZATION:
CORPORATE CONSULTING & DEVELOPMENT COMPANY, LTD RESEARCH TRIANGLE PARK, NORTH CAROLINA REPORT INSPECTION NO.-
99900511/85-02 RESULTS:
PAGE 2 of 6 A.
Violations None.
B.
Nonconformances None.
C.
Unresolved Items None.
D.
Other Findings and Comments 1.
Observation of a loss of Coolant Accident (LOCA) test for Job Number (JN) 1953:
General Public Utilities Nuclear Corporation contracted with CCL to conduct EQ test (thermal aging, radiation and LOCA) of eight single conductor and eight multiconductor Rockbestos Firerone R cable specimens. The cable specimens are being qualified for use in the Oyster Creek and Three Mile Isla'nd Nuclear Generating Stations.
The NRC inspectors and.Sandia consultant (NRC inspection team) reviewed data from all previous functional tests.
Post radiation Insulation Resistance (IR) readings taken with a megohmmeter showed five single conductor cable specimens of less than 200,000 ohms and three single conductor cable specimens of zero ohms.
CCL documented the results on a Record of Anomaly.
CCLs customer requested that the eight single conductor cable specimens be included in the LOCA test, however it was stipulated that no electrical potential or current shall be applied and only electrical resistance between conductor and ground will be measured.
One attempt of performing the LOCA test was conducted on December 7, 1985 prior to the NRC inspection.
However, due to steam leaking at the cable penetration, the pressure could not be maintained and the test was discontinued.
Prior to the second attempt of the LOCA test the NRC inspection team reviewed Test Procedure (TP) 1953-4, revision 2, dated December 6, 1985, to verify that; (a) the TP was approved and reviewed by CCL, (b) test instrumentation was adequately described, (c) environmental 22
ORGANIZATION:
CORPORATE CONSULTING & DEVELOPMENT COMPANY, LTD RESEARCH TRIANGLE PARK, NORTH CAROLINA REPORT INSPECTION NO.-
'990006,11/85-02 RESULTS:
PAGE 3 of 6 conditions were established and described (pressure and temperature profiles), (d) all prerequisites for the given test had been met and (e) the simulation test using dummy masses was completed.
The NRC inspection team reviewed the set-up to verify that; (a) instrumentation was calibrated, (b) accuracies of l
instrumentation were consistent with the requirements of the TP, l
(c) CCL Quality Assurance test monitor reviewed test set-up and signed off the test log and (d) functional tests were performed prior to the test. The NRC inspection team observed the initial ramp on December 9, 1985. The ramp time for temperature was specified to be on a best effort basis.
During the test all multi conductor cable specimens were continuously energized at 600 vac and carried 5 amps ac. The NRC inspection team observed that pressure could not be maintained in the test chamber due to steam leaking at the cable penetration. CCLs customer requested that the test be continued through the three hour high temperature plateau and then terminated. The pre LOCA IR readings for three of the conductors of specimen no. 1953-012 were zero ohms and low readings were observed for several other specimens.
IR readings taken twenty minutes prior to the end of the three hour high temperature plateau improved.
Prior to the third attempt of the LOCA test the NRC inspection team reviewed the set-up to verify that; (a) specimens were located in the chamber as specified in TP (single conductor cables were not included in this test) and (b) thermocouples and pressure transducers were located in chamber as specified in the TP.
The NRC inspection team observed the initial ramp on December 11, 1985 and made periodic checks during the remainder of the inspection. Cable specimen 1953-012 was isolated from the circuit during this ramp sin'ce pre-LOCA IR readings were low.
Since the testing was not completed, the final test report and test results will be reviewed during a future inspection.
2.
Observation of an accident simulation test for JN 1927.07 Patel Engineers (Patel) contracted with CCL to conduct an accident simulation -test of four 8 point Buchanan 0241 terminal blocks (two pre-1970, two new) and four Raychem WCSF-N cable splices (1" overlap). The environmental qualification of the 23
ORGANIZATION:
CORPORATE CONSULTING & DEVELOPMENT COMPANY, LTD RESEARCH TRIANGLE PARK, NORTH CAROLINA REPORT INSPECTION NO.-
99900511/85-02 RESULTS:
PAGE 4 of 6 terminal blocks was for Nebraska Public Power District's Cooper Nuclear Power Plant and the cable splices for New York Power Authority's James A. Fitzpatrick Nuclear Power Plant. The terminals and splices were tested under identical temperature and pressure profiles simultaneously in the same chamber.
Accident test requirements for the terminal blocks were defined in paragraph 3.7 of Patel Final Qualification Plan PEI-TR-860100-01 Rev. B dated November 18, 1985. Test requirements for the splices were defined in paragraph 3.4 of Patel Test Procedure PEI-TR-82-4-203 dated September 30, 1985.
The NRC inspection team reviewed the Qualification Plan & Test Precedure to verify that (a) environmental conditions were established (temperature and pressure profiles were identical) and (b) test acceptance criteria was established. Prior to the start of th( test the NRC inspection team reviewed the test set-up to verify that instrumentation was calibrated.
The' NRC inspection team observed the initial ramp on December 10, 1985.
Steam leaks at the penetration were observed. The steam condensed directly on unsealed terminal connections outside the test chamber. The terminal connections were used to connect cabling between the test specimens and test instrumentation. CCL's customer requested that the test be terminated and that the unsealed terminal connections be coated with Patel Engineers' Conformal Coating (PECC) prior to retest. The NRC inspection team observed the inital ramp (second attempt) on December 10, 1985 and made periodic checks during the inspection.
Early in the test the fuse on the new uncoated 600 vac terminal block circuit shorted out.
On the second day of the test the 250 vdc power supply failed.
The two terminal blocks were subsequently reenergized using 125 vdc. The old uncoated 125 vdc terminal blocks repeatedly had fuse failures with erratic leakage current readings. On the third day one Raychem splice powered at 600 vac failed causirig the protective fuse to fail and was subsequently disconnected from the circuit. Leakage current measurements were for information purposes only.
The final test report and test results will be reviewed by the NRC inspection team during a future inspection.
3.
Technical Evaluation The NRC inspection team performed a technical evaluation arid review of previous testing conducted on two test programs for 24
1 ORGANIZATION:
CORPORATE CONSULTING & DEVELOPMENT COMPANY, LTD RESEARCH TRIANGLE PARK, NORTH CAROLINA REPORT INSPECTION NO.-
99900511/85-02 RESULTS:
PAGE 5 of 6 qualification of safety related electrical equipment. The following table summarizes the test programs examined including equipment type, type of test, plant and documents reviewed.
Test Equipment Documents Program Type Test Plant Reviewed 1927.05-01 anti-wicking accident Callaway Patel Qualifi-splices simulation cation plan, purchase order, Patel test procedure, test report 1927.05-02 electrical accident Generic Patel test connectors simulation procedure, purchase order, test report The NRC inspection team reviewed the EQ process prescribed in each test plan / procedure and reviewed test results.
Each test plan / procedure and related engineering documents were examined for the following:
a)
Adequate test instrumentation and their accuracies were described and used to meet the requirements of IEEE-STD-323/1974.
b)
Equipment interfaces were addressed.
c)
Test acceptance criteria were established as described in the test specification or in the design engineering I
documents, such as calculations and engineering letters to meet the requirements of IEEE-STD-323/1974.
d)
Same equipment was used for all phases of testing and represented a standard production item.
1 e)
Enviromental conditions were established and described (e.g., pressure and temperature profiles),
f)
Test results were adequately reduced and evaluated against established acceptance criteria described in customer test specifications or purchase orders.
25
ORGANIZAT10N: CORPORATE CONSULTING & DEVELOPMENT COMPANY, LTD RESEARCH TRIANGLE PARK, NORTH CAROLINA REPORT INSPECTION NO.-
99900511/85-02 RESULTS:
PAGE 6 of 6 g)
All prerequisites for the given tests as outlined in the test specification had been met.
h)
Test equipment included a description of all materials, parts, and subcomponents.
i)
Notice of Anomaly Records were properly documented.
j)
Appropriate margins were applied.
With respect to (f), no acceptance criterion was established in Patel test procedures PEI-TR-841203-03 and PEI-TR-841203-04 for functional testing during the accident simulation. However, post accident functional tests did specify acceptance criteria in the two Patel test procedures, but discussions with CCL personnel indicated that post accident functionals were not required to be performed by CCL. Therefore the NRC inspection team was unable to evaluate the test results against established acceptance criteria. The NRC inspection team will evaluate the post accident functional test results during a future inspection at Patel Engineers.
26 l
ORGANIZATION:
CORPORATE CONSULTING & DEVELOPMENT COMPANY, LTD RESEARCH TRIANGLE PARK, NORTH CAROLINA REPORT INSPECTION 1/6-7/86 INSPECTION N0.: 99900511/86-01 DATE(S):
2/3-4/86 ON-SITE HOURS:
28 CORRESPONDENCE ADDRESS:
Corporate Consulting & Development Company, Ltd.
ATTN:
Dr. J. R. Yow, President Post Office Box 12728 Research Triangle Park, North Carolina 27709 ORGANIZATIONAL CONTACT: Mr. Carson Blanton, Jr., QA Manager TELEPHONE NUMBER:
(919) 362-8800 NUCLEAR INDUSTRY ACTIVITY:
Corporate Consulting and Development Company, Ltd.
(CCL) provides engineering, consulting, and testing services to the nuclear industry for seismic analysis, testing, and nuclear environmental qualifications of equipment.
ASSIGNED INSPECTOR:
/b(s.,/ 6 /
.t.. /
sN v/N 1
R. H. Lasky, Equipment QuaH fication Inspection Date Section (EQIS)
OTHER INSPECTOR (S):
M. Jac c, Sandia National Laboratory in APPROVED BY:
d.2 hbc ko S2~y U. Potapovs, Chief, EQJS, Vendor Program Branch ate INSPECTION, BASES AND SCOPE:
A.
BASES: Appendix B to 10 CFR Part 50 and 10 CFR Part 21.
i B.
SCOPE: There were two (2) visits made to Corporate Consulting &
Development Company, LTD (CCL). The first on January 6-7, 1986 and the second on, February 3-4, 1986.
Both visits were related to the qualification of KXL 780 insulation compound Rockbestos cable for (continued on page 2)
+
PLANT SITE APPLICABILITY:
Sequoyah 1 & 2 (50-327/328); Watts Bar 1 & 2 (50-390/391).
I 27
ORGANIZATION:
CORPORATE CONSULTING & DEVELOPMENT COMPANY, LTD RESEARCH TRIANGLE PARK, NORTH CAROLINA REPORT INSPECTION NO.-
99900511/86-01 RESULTS:
PAGE 2 of 5 B.
SCOPE:
(continued)
Tennessee Valley Authority (TVA). The inspections consisted of:
(1) a technical evaluation of equipment qualification (EQ) test activities for safety-related equipment; (2) witnessing of inprocess EQ testing; and (3) verification of implementation of quality assurance (QA) program.
A.
VIOLATIONS:
None.
B.
NONCONFORMANCES:
None.
C.
UNRESOLVED ITEMS:
None.
D.
OTHER FINDINGS AND COMMENTS:
1.
Observations of Insulation Resistance and Elongation Test for Job Number (JN) 1963:
(1/6-7/86 inspection)
The Tennessee Valley Authority (TVA) contracted with Corporate Consulting and Development Company, LTD (CCL) to conduct testing to establish similarity between two formulations of Rockbestos chemically cross-linked polyethylene cable, the old formulation KXL 780 and the newer formulation KXL 760-D. TVA intended to qualify XXL 780 type cable by proving similarity to the previously qualified KXL 760-D type cable.
The CCL testing was intended to show similarity between the XXL 780 and KXL 760-D type cables, both before and after the cables were irradiated. The testing consisted of insulation resistance checks, thermogravimetric analysis and elongation testing of the wire insulation.
The test setup consisted of thirty, six inch samples of both the KXL 780 and KXL 760-0 wire insulation.
Fifteen of each sample were irradiated. There were also two, approximately fifteen foot lengths of each the KXL 780 and the KXL 760-D cables.
Cables were two conductor cables.
28
ORGANIZATION: CORPORATE CONSULTING & DEVELOPMENT COMPANY, LTD RESEARCH TRIANGLE PARK, NORTH CAROLINA REPORT INSPECTION NO.-
90Q00511/86-01 RESULTS:
PAGE 3 of 5 The NRC inspector and the Sandia consultant (NRC inspection a.
team) examined the CCL procedures for the insulation resis-tance testing, elongation testing and the purchase order from CCL to Georgia Tech for the irradiation of the samples. No deficiencies were noted in these documents.
b.
The thermogravimetric analysis (TGA) calculations were examined by the Sandia consultant. The data was still in rough form and had not been formalized by CCL. CCL indicated that the calcu-lated activation energies between the KXL 780 and KXL 760-D were within three percent of each other. No deficiencies were noted in the TGA.
c.
The insulation resistance (IR) data was examined by the NRC inspection team. The IR data was the same for both formulations of cable. The cables had the same IR unirradiated as they did.
after they were irradiated.
All IR meter readings were greater than the range of the instrument used (1000 Mohms). No deficiencies were noted in the IR testing.
d.
The NRC inspection team observed the elongation testing of seven of the sixty samples.
Four of the samples were the unirradiated KXL 760-D and the other three samples were the unirradiated KXL 780. All of the KXL 780 samples stretched to the maximum travel of the test machine (approximately 400 percent elongation).
It was noted that the KXL 780 samples tended to stretch uniformly over the length of the sample. Of the four unirradiated KXL 760-D samples observed, only the black sample elongated to the maximum travel of the machine. The three white KXL 760-D samples all broke with only a 15-20 percent elongation, measured at the center two inches of the sample.
It was also observed that while the KXL 780 tended to elongate uniformally when stretched, the KXL 760-D tended to neck down at the ends. There were no procedural deficiencies noted while the testing was being observed.
2.
Post LOCA/HELB Testing of Rockbestos Cable:
(2/3-4/86 inspection)
TVA had determined that similarity between KXL 780 and KXL'760-0 a.
Rockbestos cable could not be proved and contracted with CCL to perform a Design Basis Accident (DBA) simulation on the fifteen foot piece of KXL 780 Rockbestos cable previously irradiated for IR testing.
i l
29 i
ORGANIZATION: CORPORATE CONSULTING & DEVELOPMENT COMPANY, LTD RESEARCH TRIANGLE PARK, NORTH CAROLINA REPORT INSPECTION NO.-
99900511/86-01 RESULTS:
PAGE 4 of 5 b.
The NRC inspection team arrived the day before the KXL 780 cable was to be removed from the test chamber for post LOCA/HELB testing, to review test procedures and test data.
Review of radiation test data from Georgia Tech Hot-Cell Operations, Neely Nuclear Research Center showed that data sheets were not signed by Georgia Tech.
CCL had previously recognized the lack of signatures and had documented that they were requesting a signed data sheet from Georgia Tech.
No deficiencies were noted in the thermal aging procedure.
No calculations were required since TVA had provided the time and temperature to be used for aging.
The DBA simulation procedure was examined. Although the written text recognized the removal of the load resistors during the leakage current checks, figure 2 of the procedure did not.
CCL changed the figure to recognize that the load resistors would be disconnected during the leakage current checks. This was considered a clarification of the procedure and not a deficient condition.
The chemical spray work sheet was examined and no deficient 4
conditions were found.
The Instrument Calibration Log was examined to ensure all instrur;ientation used was within calibration requirements.
No deficient conditions were found.
Selected insulation resistance and leakage current measurements that had been taken during the LOCA/HELB simulation were reviewed and verified to be within the specified values.
The LOCA/HELB simulation strip chart recordings were examined and found to contain all pe'rtinent information.
c.
The NRC inspection team observed the post LOCA/HELB functional test of the Rockbestos KXL 780 sample.
The test chamber was filled with water prior to the measurements.
Leakage current measurements were made at 24 Vac, 48 Vac, 120 Vac, and 480 Vac. Leakage currents at 24 Vac and 48 Vac were too i
30 i
P
ORGANIZATION:
CORPORATE CONSULTfNG & DEVELOPMENT COMPANY, LTD RESEARCH TRIANGLE PARK, NORTH CAROLINA REPORT INSPECTION NO.-
99900511/86-01 RESULTS:
PAGE 5 of 5 low to measure. The leakage current at 120 Vac was approximately 0.04 mA and at 480 Vac was approximately 0.18 mA. All insulation resistance readings were greater than 500 Mohms.
The KXL 780 cable was removed from the test chamber, inspected and then subjected to a 40 times the cable diameter bend. A voltage withstand test at 2000 Vac (based on 80 Vac/ mil and 25 mils insulation of the cable) was conducted. Both conductors passed the five minute 2000 Vac test.
The cable appeared to be in good condition when removed from the chamber. Some yellow deposits were on the cable, but no apparent physical damage.
3.
CCL audit report on Georgia Tech's Hot-Cell Operations Center, Neely Nuclear Research Center, was examined during the February 3-4, 1986 inspection.
Report was examined to verify that CCL had approved the procedure by which Georgia Tech documents nonconformances.
Section 15.2 of report concurred on Georgia Tech's procedure for documenting nonconformances.
4.
Elongation test data was reviewed during the February 3-4, 1986 inspection. The final results of the test were as follows:
(see Section D.le for earlier inspection of this test) a.
Twelve of the KXL 780 non-irradiated samples exceeded the machine travel.
b.
One of the KXL 780 irradiated samples exceeded the machine travel.
c.
Four of the KXL 760-D non-irradiated samples exceeded the machine travel.
d.
None of the KXL 760-D irradiated samples exceeded the machine travel.
Two reports of anomalies were written concerning the number and size of test samples. The reports had not been approved at the time of the inspection.
No deficiencies were noted.
31
i 4
ORGANIZATION:
THE F0XB0R0 COMPANY FOXBOR0, MASSACHUSETTS REPORT INSPECTION INSPECTION NO.-
99900225/85-01 DATE(Sl-12/9-13/85 ON-SITF H0tlRS-84 i
CORRESPONDENCE ADDRESS: The Foxboro Company ATTN: Mr. J. W. Graham Vice President i
38 Neponset Avenue Foxboro, Massachusetts 02035 ORGANIZATIONAL CONTACT: Mr. M. J. Berberian, Corporate QA Manager TFt FPHONF NilMRFR-(617) cdLA7RO PRINCIPAL PRODUCT: Industrial Control Systems & Instruments NUCLEAR INDUSTRY ACTIVITY:
Current nuclear in house orders are for the following utilities: Arizona Public Service, Florida Power & Light, Duke Power, Power Authority of the State of New York, Northern States Power, GPU Nuclear Corporation (Three Mile Island), and Gulf States Utilities.
d
/
i ASSIGNED INSPECTOR:
, 4_ w -
/,/__ www L u.-sc
,/J.~JfPefosino,ReactiveInspectionSection(RIS) Date
/
OTHER INSPECTOR (S):
E. Yachimiak, RIS T. Keck, Brookha N tional Laboratories Mu[sc_,
)
APPROVED BY:
E. W. MerschofV/ Chief, RIS, Vendor Program Branch Date INSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR Part 21 and Appendix B to 10 CFR Part 50 B.
SCOPE:
This inspection was made as a result of a Haddam Peck Licensee Event Report concerning the degradation of electrical conductor insulation on The Foxboro Company's E-line" electronic controllers.
PLANT SITE APPLICABILITY:
See next page.
33,
ORGANIZATION: THE FOXBORD COMPANY FOXBOR0, MASSACHUSETTS REPORT INSPECTION NO.-
99900225/85-01 RESULTS:
PAGE 2 of 9 Plant Site Applicability:
Arkansas 1 & 2 (50-313/368); Beaver Valley 1 & 2 (50-334/412); Bellefonte 1 & 2 (50-438/439);Braidwood 1 & 2 (50-456/457): Browns Ferry (50-454/455);
1, 2 & 3 (50-259/260/296); Brunswick 1 & 2 (50-324/325); Byron 1 & 2 Callaway (50-483); Calvert Cliffs 1 & 2 (50-317/318); Catawba 1 & 2 (50-413/414); Clinton 1 (50-461); Cook 1 & 2 (50-315/316); Cooper Station (50-298); Crystal River (50-302); Davis-Besse (50-346); Dresden 2 & 3 (50-237/249); Duane Arnold (50-331); Enrico Fermi (50-341); Farley 1 & 2 (50-348/364); Fitzpatrick (50-333); Fort Calhoun (50-285); Fort Saint Vrain (50-267); Ginna (50-244); Grand Gulf (50-416); Haddam Neck (50-213); Hatch 1 & 2 (50-321/366); indian Point 2 & 3 (50-247/286); Kewaunee (50-305);
Lacrosse (50-409); LaSalle 1 & 2 (50-373/374); Limerick 1 & 2 (50-352/353);
Maire Yankee (50-309); Millstone 1, 2 & 3 (50-245/336/423); Monticello (50-263);
Nine Mile Point 1 & 2 (50-220/410); North Anna 1 & 2 (50-338/339); Oyster Creek (50-219); Palisades (50-255); Palo Verde 1, 2 & 3 (50-528/529/530);
Peach Bottom 2 & 3 (50-277/278); Perry (50-440); Pilgrim (50-293); Point Beach 1 & 2 (50-266/301); Prairie Island 1 & 2 (50-282/306); Quad Cities 1 & 2 (50-254/265); Rancho Seco (50-312); River Bend (50-458); Robinson (50-261); San Onofre 1, 2 & 3 (50-206/361/362); Seabrook (50-443); Sequoyah 1 & 2 (50-327/328); South Texas 1 & 2 (50-498/499); Saint Lucie 1 & 2 (50-335/389); Surry 1 & 2 (50-280/281); Three Mile Island 1 & 2 (50-289/320);
Trojan (50-344); Turkey Point 3 & 4 (50-250/251); Vermont Yankee (50-271);
Vogtle 1 & 2 (50-424/425); Washington Nuclear 2 & 3 (50-397/508); Waterford (50-382); Watts Bar 1 & 2 (50-391/390); Wolf Creek (50-482); Yankee Rowe (50-029); and Zion 1 & 2 (50-295/304).
A.
Inspection Issues 1.
The purpose of this inspection was to:
a.
Review the conductor insulation degradation problem which was identified on LER #84-017, dated October 4, 1984 from the Haddam Neck Nuclear Station. Additionally, to evaluate The Foxboro Company's (TFC) determination that 10 CFR Part 21 does not apply to the insulation degradation, b.
Determine the adequacy of the implementation of TFC's quali_ty assurance program at the Highland facility and the corporate office.
34
(
t
l ORGANIZATION: THE FOXB0RD COMPANY FOXB0R0, MASSACHUSETTS REPORT INSPECTION Mn -
QQQnn??G/AG nl RF9ttl T9 -
Pact 7 nf.o 8.
Inspection Findings Violations 1.
Contrary to Section 21.21, " Notification 'of failure to comply or existence of a defect," of 10 CFR Part 21:
a.
TFC failed to notify the Commission of a defect in the conductor insulation of an interconnection coil cord cable set'(cable set),
part number N0101NW or N0101PW. This was identified to TFC by the San Onofre Nuclear Station in tne fall of 1978. TFC's Component Failure Analysis Report # 1979-CFA-28, dated June 11, 1979, indicates that a defect was found and identified at an NRC licensee, wh'ich involves "... crumbling and disintegration' of insulation....". However, no evaluation for reportability was performed or notification made to the Commission or the applicable licensees concerning the insulation degradation defect until October 18, 1985 (85-01-01).
This is a Severity Level III violation (Supplement VII).
b.
TFC procedures which were adopted as required by 10 CFR Part 21 failed to adequately provide for evaluating ceviations for reportability or informing the licensee of. deviations in order that the licensee could cause the deviation to be evaluated.
The following examples are deviations which TFC was_ cognizant of but were not evaluated for reportability or transmitted to 4
the licensees for their evaluation,s, specifically (85-01-02):
1.
TFC Report # 1984-CFA-08, " Leaking / Corrosion on Electrolytic Capacitors" - (Sangamo-generic) 2.
TFCReport#1985-CFA-019(S1)," Transistor / Diode / Relay Failures" - (UCM Power. Supply).
3.
Highland # 11364/11365, " Failure of Pen Drive (PWA)
Cards-N227S Recorders" - (Peach Bottom).
~
4.
High1and # 1102-1104 " FAT /AI intermittent output supply *
- (Northeast Utilities).
4 35 1
.I.
ORGANIZATION: THE FOXB0R0 COMPANY FOXB0R0, MASSACHUSETTS REPORT INSPECTION 44400??5/85-01 RESULTS:
PAGE 4 of 9 NO 5.
Highland # 4112, " Power Supply Fuse Problem" - (Pilgrim).
6.
Highland #5169, "62H Electronic Controllers - Erratic Outputs" - (End of Life).
This is a Severity Level IV violation (Supplement VII)
Nonconformances 1.
Contrary to Criterion II, " Quality Assurance Program," of Appendix B to 10 CFR Part 50 and Section 2.7 of TFC Highland Quality Program
, Manual, no indoctrination program was established or implemented for any personnel performing activities that affect nuclear quality with the exception of the Q.C. inspectors and assembly line personnel (85-01-03).
2.
Contrary to Criterion V, " Instructions, Procedures, and Drawings,"
of Appendix B to 10 CFR Part 50 and Section 4.1, of TFC's Corporate Quality Assurance Procedure 3.18.3 (85-01-04):
a.
Procedures were not adequately established to provide instructions for personnel to notify the Foxboro Nuclear Safety l
Sub-Committee (NSS) of "... conditions adverse to quality...."
b.
Inadequate procedure implementation was evidenced by the lack of NSS records to indicate the following deviations were evaluated for 10 CFR Part 21 reportability:
1.
Peach Bottom letter to TFC, dated July 22, 1985
" Failed printed circuit cards," utilized on safety related model No. N 2275 recorders.
l 2.
Report No.1984-CFA-08 "Sangamo Electrolytic Capacitors, Series 139R, Leaking and Corrosion found in Three Lots."
3.
Report No. 1985-CFA-019(S1)
" Transistor diode and relay failures in the UCM Power supplies," (this was identified as a "Long-standing problem").
4.
Highland No. 1102-1104 "2A0VAI, intermittent output supply" - (Northeast Utilities).
5.
Highland No. 4112
" Power supply fuse problem" - (Pilgrim).
36 l
l
W ORGANIZATION: THE FOXE0R0 COMPANY f0XB0R0, MASSACHUSETTS
=
REPORT IN5'ECTION I
RESULTS:
PAGE.5 of.9 NO. _ 99900225/85-01 l
6.
Highland No. 5169 "628 Electronic Controllers - Erratic Outputs" - (End of life susceptibility).
3.
Contrary to Critert;in VII, " Control of Purchased Material, Equipment, and Services," of Appendix B to 10 CFR Part 50, an outside talibration laboratory was on TFC apprGwed vendors list, without any documented audit results in TFC files q
(85-01-05).
a 4.
Contrary to Criteriors XII, " Control of Measuring and Test Equipment," of Appendix B to 10 CFR Part 50 and TFC Highland Plant Quality Procedure 56.01, instrument numbers 204700-1 (cal. due date 08/18/85), and 20425-1 (cal. due date 11/16/85) were found to have calibration due dates which were past due.
These instruments were found in the Highland manufacturing areas where they could be utilized (85-01-06).
5.
Contrary to Criterion XVIII, " Audits," of Appendix B to 10 CFR Part 50 and Section 18 of TFC Highland Plant Quality Program Implementation Plan, in-process internal audits were not performed quarterly as required for "all stages of operatitins" by the Highland Plant quality control audit schedule as indicated below (85-01-07):
One out of the 37 scheduled audits was not performed in the first calendar quarter of 1985.
30 out of the 37 scheduled audits were not performed in the second calendar quarter of 1985.
37 out of the 37 scheduled audits were not performed in the third calendar quarter of 1985.
37 out of the 37 sche 6uled audits were not performed in the last caleridar quarter of 1985, as of December 13, 1985.
C.
Background Inforcation TFC manufactures industrial electrical and pneumatic control systems, devices, and instruments. Several of TFC manufacturing locations in the U.S., Canada, and England, supply components to the U.S. nuclear industry.
The insulation degradatior, problem was identified with TFC E-Line controllers, and is found on the " cable set," TFC Part Number N0101NW or N0101PW. The " cable set" is a 15 conductor coiled cord cable with a 37
t ORGANIZATION: THE FOXBORO COMPANY FOXBOR0, MASSACHUSETTS REPORT INSPECTION H0 - 99900225/85-01 RESULTS:
pAGE 6 of 9 molded plug at one end and a terminal block at the other end. The coiled cord cable was a procured item and was supplied by two manufacturers.
The manufacturers were Whitney-Blake Company and Gavitt Wire and Cable.
The E-line controllers and associated cable sets were manufactured and assembled by TFC during the approximate period of 1960 - 1974. Until 1984, cable set replacement units were offered by TFC.
TFC was first informed of the degradation problem by the San Onofre r
Unit i nuclear station in the fall of 1978.
The conductor insulation disintegrated on all conductors while the cable jacket remained intact.
1FC was again informed of the same problem by the Haddam Neck Plant in November of 1984.
LER #50-213/84-017 was generated by Connecticut Yankee i
(CY),
i The University of Connecticut performed an analys's of two of the j
degraded ceble sets for CY and concluded in a letter to CY and Northeast i
utilities, dated August 2, 1985, that:
a.
... an inadequate amount of stabilizers or poorly chosen stabilizers, would contribute to the severe... breakdown, in this case.,. lack of effective stabilizers may explain the degradation under apparently mild service conditions...."
b.
... the type of insulation is not known exactly, but may be synthetic rubber... possibly of the butadiene-stycene (Bona-S) or butadiene-acrylonitrile (Buna-N) type...."
~
c.
"... insu1ation... is badly embrittled, both inside and outside the jacket... touching the insulation caused it to crack and fall off... "
d.
... the conclusion seems clear that the insulation used in these units has poor stability, whether for reasons mainly of stabilizer j
choice and amount or of the rubber type being inappropriate for this application...."
1 In response tc this problem, Foxboro stated in an October 18, 1985 letter i
to CY that the cable has a design life of ten years and insulation should be examined for deterioration at least annually and that TFC "does not believe that 10 CFR Part 21 applies because the instruments pre-date the i
regulation."
However, the notification requirements of 10 CFR Part 21 still apply if a deviation amounting to a defect, as defined by Part 21 is discovered 38
ORGANIZATION: THE FOXBOR0 COMPANY FOXBORO, MASSACHUSETTS REPORT INSPECTION NO.-
99000225/85-01 RESULTS:
PAGE 7 of 9 after the effective date of 10 CFR Part 21.
If it was not possible for TFC to determine whether the deviation could create a significant shfety hazard for the instal ~ led equipment application, then all pertinent information should have been provided to the end-user for evaluation and deterrnination of whether a reportable defect exists.
TFC was notified of the insulation $legradation by the San Onofre Nuclear Plant, Unit 1, in 1978.
In the t'all of 1978, San Onofre informed TFC that they were having failures with the cable sets in which the conductor insulation disintegrated on all conductors while the cable. jacket remained intact. San Onofre stated that the insulation "... exhibited ozone type attack and failure (crumbling and disintegration of insulation)."
During the ti!ne period when TFC was first notified in 1978 and the November 1984 notification by CY, no actions were taken by TFC to:
1.
Notify 'the Comission of a defect.
2.
Notify the licentee for them to cause an evaluation to be performed.
3.
Perform a 10 CFR Part 21 reportability evaluation (reference violations 85-01-01/02).
D.
Other Corrents 1.
Insulation Degradation TFC has stated that the cable set which was utilized in their E-Line controllers has a design life of ten years after which, "end of life susceptibility" is reached.
Further, they have stated the insulation should be inspected "at least annually." Hcwever, no informa tion could be produced which indicated that a specific ten year design life was factored into the performance of the cable set, or that the 10 year product' life span has been transmitted to the end users.
An analysis was performed by the University of Connecticut (UC) on two " cable sets" from the Haddam Neck Plant. Their conclusion states:
"... the conclusion seems clear that the insulation used in these units has poor stability, whether for reasons mainly of stabilizer choice and amount or of the rubber type being inappropriate for this application." They go on to say that the key i
factor in the degradation app' ars to be "... material selection...."
e Consequently, it appears that the maximum life of the cable sets is 10 years and the cable sets should be inspected at least annually as
)
t l
suggested by Foxboro. Discussions with the cable manufacturers have determined that the coil cord sets typically have a life expec-tancy of 31-10 years under mild service conditions and environment.
i 39
ORGANIZATION: THE FOXBORO COMPANY FOXBOR0, MASSACHUSETTS REPORT INSPECTION NO_-
99900995/AR-01 RF9tHT9-Par # R nf 0 2.
10 CFR Part 21 During discussions with TFC personnel from many different departments, it was observed that a lack of cognizance of 10 CFR Part 21 was apparent. Two departments which independently evaluate failures and deviations appear to do an adequate job of analyses and reaching the root cause of failures or deviations. However, it was apparent that the personnel were not aware of any 10 CFR Part 21 considerations.
Because of the lack of awareness of the 10 CFR Part 21 procedure the completed analyses appear to stop at a department section supervisor level. Discussions with the department section supervisor personnel indicate that they too are unfamiliar with the Part 21 requirements, (reference nonconformance 85-01-04).
3.
Product "End-of-Life" Foxboro control equipment is used widely throughout the nuclear industry in commercial nuclear applications. The Haddam Neck LER #50-312/84-017 described the degradation of cable insulation.
The cause of this deterioration was stated as, "... obviously a function of age...", in the LER.
On October 18, 1985, Foxboro issued letters to all applicable customers identifying "End-of-Life" susceptibility on model-E cable sets. These coil cord cable sets were stated as having a
" design" life of ten years, after which the insulation of the individual wires would deteriorate. Additionally, possible "End-of-Life" failures of electrolytic capacitors and silver plated switch contacts were mentioned.
However, this correspondence has been the only one that Foxboro has made concerning "End-of-Life" failures. During discussions with Foxboro management, concerns were expressed as to the " design" life of other Foxboro components. Since it could not be demonstrated that " design" life was information known by the customer, Foxboro components may not be receiving appropriate maintenance checks to ensure their continued operability by the applicable licensees.
Further discussion over this item revealed that Foxboro will be evaluating the need to inform nuclear customers of "End-of-Life" susceptibility in other Foxboro components.
l 40
0RGANIZATION: THE FOXBOR0 COMPANY FOXBOR0, MASSACHUSETTS REPORT INSPECTION NO.-
99900225/85-01 RESULTS:
PAGE 9 of 9 l
4.
Highland Facility Tour:
A tour was conducted of the Highland Plant operations which included the calibration laboratory, wave soldering, shipping /
receiving, PC card assembly and general product repair areas.
A sample of 18 instruments were reviewed during the tour to verify that required recalibrations of instruments were being performed.
Two out of 18 had not been calibrated as required. They were identified as instrument # 204700-1 (calibration due date 8/18/85),
and instrument # 20425-1 (calibration due date 11/16/85) (reference nonconformance #85-01-06).
E.
Persons Contacted J. W. Grahm Vice President **
M. J. Berberian QA Manager-Corporate **
R. M. Grahn CQA**
W. F. Griffin QC Engineering (QCE)
D. R. Scamman QCE C. O. Wilson QCE Marion Tiehle Inspection Support Surinder Kumar QCE M. Faria QCE-Customer Support G. Hill QCE H. Rizvi Component and Reliability R. La Rose Component and Reliability D. W. Moon Reliability L. Hewey CQA Labs C. Lundstedt Lab R. Vaillancourt Senior Buyer C. Stevens QCE W. Tripp Test Equipment
- Atter.ded exit meeting.
l 41
ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR FIELD SERVICES DEPARTMENT KING OF PRUSSIA, PENNSYLVANIA REPORT INSPECTION INSPECTION N0.: 99901001/86-01 DATE:
1/27-31/86 3N-SITE HOURS: 60 CORRESPONDENCE ADDRESS: General Electric Company Nuclear Field Services Department ATTN: Mr. G. C. Nelson, Manager 1000 First Avenue King of Prussia, Pennsylvania 19406 ORGANIZATIONAL CONTACT: Mr. Vincent P. Kenney, QA Manager TELEPHONE NUMBER:
(215)962-6007 NUCLEAR INDUSTRY ACTIVITY: The General Electric (GE) Nuclear Plant Services Department offers NRC licensees the following services:
(1) ASME Section III Code Work; (ASME Stamps "N", "NA", and "NPT" are held by GE) and (2) 10 CFR Part 50 Appendix B modification, repair, maintenance, and replacement services. The area of services may be mechanical, electrical, civil, or structural.
l ASSIGNED INSPECTOR: [
F
(
,a J.
dtrosiino, Reactive Inspection Section (RIS) ' [ fate OTHERINSPECTOR(S):
E. H. Gray, Region I, Engineering Programs Branch K. G. Aspinwall, Brookhaven National Laboratory h
uiu d
APPROVED BY:
/
E. W. Merschoff, Chief ( RIS, Vendor Program Branch D' ate INSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR Part 21, Appendix B of 10 CFR Part 50 and ASME Section IX.
B.
SCOPE:
This inspection was made to:
(1) verify the Quality Assurance TQKT program implementation in the areas of welder performance qualifi-cations and welding procedure qualification; (2) review and obtain copies of specific GE " Service Advice Letters;" and (3) discuss LER "50-219/85-14 concerning low oil level in two GE transformers at the Oyster Creek station.
PLANT SITE APPLICABILITY:
Recent GE services were provided at: Hope Creek 1 & 2 (50-354/355); Indian Point 2 (50-247); Limerick (50-352); Oyster Creek (50-219);
i Peach Bottom 2 (50-277); Perry (50-440); Pilgrim (50-293); Salem (50-272);
l Susquehanna 1 & 2 (50-387/388); andWolfCreek(50-482).
l 43
ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR FIELD SERVICES DEPARTMENT KING 0F PRUSSIA, PENNSYLVANIA REPORT INSPECTION NO.: 99901001/86-01 RESULTS:
PAGE 2 of 6 A.
VIOLATIONS:
None.
B.
NONCONFORMANCES:
1.
Contrary to Criterion V, " Instructions, Procedures, and Drawings,"
of Appendix B to 10 CFR Part 50, and Article III, " Welding Performance Qualifications," of Section IX of the ASME Boiler and Pressure Vessel Code, no procedures or instructions had been written or implemented to assure that full supervision and control of the ASME Section IX welder performance qualification testing is achieved.
(85-01-01) 2.
Contrary to Criterion XVIII, " Audits," of Appendix B to 10 CFR Part 50 and Section 18, " Audits," of GE's QA Manual PIA-AE-11, GE failed to adequately carry out a planned and periodic management audit system, and verify compliance with all aspects of the quality assurance program, specifically:
(85-01-02) a.
Internal audits have not been scheduled or performed at the Devon, Pennsylvania welder performance qualification testing facility.
b.
The 1985 " Management Audit of the Program" has not been performed.
c.
The scheduled 1984 " Management Audit of the Program" was not performed as scheduled.
However, a make-up audit for 1984 was performed in March of 1985.
C.
UNRESOLVED ITEMS:
None.
D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
Not applicable.
E.
OTHER FINDINGS CR COMMENTS:
1.
Welder and Welding Procedure Qualifications The areas of welder performance qualification and welding procedure qualification were examined for compliance to the ASME Code Section IX requirements as applicable to the nuclear safety related systems.
44 1
ORGANIZATION:
GENERAL ELECTRIC COMPANY NUCLEAR FIELD SERVICES DEPARTMENT KING 0F PRUSSIA, PENNSYLVANIA I
9EPORT INSPECTION N0.: 99901001/86-01 RESULTS:
PAGE 3 of 6 Welder performance qualification testing for the Limerick nuclear station was in progress during the NRC inspection.
Different portions and aspects of the performance testing were observed, as follows:
a.
ASME Code Section IX requirements.
b.
10 CFR Part 50, Appendix B requirements.
c.
Welder identification verification.
d.
Documentation of test observation, e.
Pipe position, dimensions, and materials.
f.
Supervision of testing, g.
Documentation of acceptability of test coupon.
h.
Documentation of ASME Section IX form for qualification.
i.
Coupon orientation marking.
j.
Utilization of procedures and instrumentation for assuring i
acceptable fit-up and tack welding, visual inspection of root pass, etc.
The inspection determined that the requirements of the ASME Code,Section IX were being met with the exception of welder performance qualification test control. Verification of the welders identity was being performed by GE, Devon, Pennsylvania personnel. This i
verification was included as a part of the welder performance l
qualification initial check in process.
I However, the Nuclear Field Services Department (NFSD) had not written or used any procedures or instructions to conduct and control the welder performance qualification testing activities (reference Nonconformance 86-01-01). This was evidenced by several inconsistencies which were observed and noted:
One test pipe assembly (TPA) for a 6G position test did not have the weld procedure process number stenciled on it.
(Different weld processes were being performed at this time.)
One TPA had the " Top" reference position (0 degrees) identified at two separate locations around the circum-ference of the TPA.
(This was identified during an in-process welder performance qualification test.)
Welders were being qualified for the Limerick project using site specific welding procedures.
However, the visual inspection procedure which was being utilized on January 29, 1986, was for the Peach Bottom site (No. PB-85-7.0), not the Limerick site.
45
ORGANIZATION:
GENERAL ELECTRIC COMPANY NUCLEAR FIELD SERVICES DEPARTMENT KING OF PRUSSIA, PENNSYLVANIA REPORT INSPECTION N0.: 99901001/86-01 RESULTS:
PAGE 4 of 6 Another welder's performance qualification test was allowed to start without having the " Top" reference location on the TPA identified. This test started and was observed by the NRC on January 28, 1986. Subsequently, the test was started again with a new TPA which was correctly marked.
2.
Audits The NFSD audit schedule was reviewed for the required system of planned and periodic audits "to determine the effectiveness of the program." Audit reports were reviewed in the areas of quality assurance; ASME Section III, Division I; ANSI-N 626.0; and NFSD program management audits.
A system of planned and periodic audits is controlled by the quality assurance department and is adequate. However, there were two areas in which NFSD failed to implement their program.
Periodic audits were not performed at or planned for the Devon facility for the nuclear safety related activities. The Devon facility is the location where the site specific welder performance qualification testing is performed (reference Nonconformance 86-01-02).
In the area of the annually scheduled NFSD management audits, a review of activities for the last three years was performed.
Annual audits for the years 1983, 1984, and 1985 were requested.
The review revealed that as of January 30, 1986 (reference Nonconformance 86-01-02):
a.
The 1983 management audit was performed in May of 1983, as scheduled.
b.
The 1984 management audit was not performed witnin the calendar year as scheduled.' However, an audio was performed in March of 1985 which stated it was the "make-up" 1984 audit, c.
The 1985 management audit was not performed as scheduled.
NFSD indicated that a "make-up" audit was tentatively scheduled for April of 1986.
46
ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR FIELD SERVICES DEPARTMENT KING 0F PRUSSIA, PENNSYLVANIA REPORT INSPECTION N0.: 99901001/85-01 RESULTS:
PAGE 5 of 6 3.
GE Product Department Service Advice Letters Discussions were conducted with NFSD personnel concerning GE
" Service Advice Letters" (SALs).
The NRC inspector requested to review and obtain copies of specific safety related SALs which the King of Prussia office may have in their files. The purpose of obtaining the copies 6f certain safety related SALs is to aid the NRC in their licensee inspection efforts.
l Copies of the requested safety related SALs were not obtained.
However, a listing was developed by NFSD of all of the SAL numbers which were in their files, along with a subject title for each. The listing of the SAL number and subject titles was provided to the NRC inspection team.
4.
Oyster Creek Substation Transformer Oil Level The Oyster Creek station issued LER #50-219/85-19, dated 8/9/85, concerning low oil levels in two GE safety related transformers.
Discussions were conducted with the cognizant NFSD personnel regarding the areas of receiving, storing, installation practices, liquid level gauge, liquid temperature indicator, pressure-vacuum gauge, required maintenance of oil levels, and oil sampling practices. Additionally, specific information on the Oyster Creek problem and NFSD's subsequent thermographic and visual inspection work was discussed.
The information obtained will be transmitted to and discussed with the cognizant NRC staff to determine if additional inspections or other actions are required.
5.
NRC Staff Concern The NRC staff he:i a concern regarding safety related activities performed at locations other tha'n plant sites.
The concern was that the safety related activities could not be easily inspected for conformance to codes, standards, and regulatory requirements.
This inspection has concluded that the activities which were observed and examined at the General Electric facilities meet the intent of ASME Code Section IX, and Appendix B to 10 CFR Part 50.
47
ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR FIELD SERVICES DEPARTMENT KING 0F PRUSSIA, PENNSYLVANIA REPORT INSPECTION NO.: 99901001/86-01 RESULTS:
PAGE 6 of 6 F.
PERSONS CONTACTED:
G. C. Nelson Manager - NFSD
- V. P. Kenney QA Manager
- R. Taron Quality Support Manager
- C. Logan Document Control Specialist R. E. Cameron NDE Specialist E. J. Hirtzo Electrical Service Manager D. J. DiFilippo QC Supervisor D. W. Diefenderfer Service Project Manager T. Byrum Service Supervisor J. F. Plantz Service Supervisor J. E. McClure Welding Supervisor W. S. Fingrutd Welding Engineer S. Ebneter Director, RI, NRC
- H. Gray Engineer, RI, NRC
- K. Aspinwall Consultant, BNL
- J. J. Petrosino Inspector, VPB, NRC
- present at exit meeting 48
ORGANIZATION:
GEORGIA INSTITUTE OF TECHNOLOGY ATLANTA, GEORGIA REPORT INSPECTION INSPECTION N0.: 9990C903/85-01 DATE:
12/20/85 ON-SITE HOURS: 4 CORRESPONDENCE ADDRESS:
Georgia Institute of Technology Neely Nuclear Research Center ATTN: Mr. Radib Karam Director 900 Atlantic Avenue, N.W.
Atlanta, Georgia 30322 ORGANIZATIONAL CONTACT:
Steve Chester, Hot Cell Operator TELEPHONE NUMBER:
(404) 894-3600 NUCLEAR INDUSTRY ACTIVITY:
Georgia Institute of Technology (GIT) is actively engaged in irradiation of components for nuclear power generating plant The service is provided to various sectors of the nuclear applications.
industry including utilities, test laboratories and equipment vendors.
O
_dm f/p/fp ASSIGNED INSPECTOR:
d Date S. Alexander, Equipment Qualification Inspection Section (EQIS)
OTHER INSPECTOR (S):
(
/
APPROVED BY:
t (,( d, _
N (' bu O n a 2 -26 SD U. Potapovs,' Chief EQIS,, Vendor Program Branch Date INSPECTION BASES AND SCOPE:
A.
BASES:
SCOPE: This inspection consisted of observation of irradiation of Rockbestos chemically cross-linked polyethylene (CXLPE) insulated Firewall III cable specimens (of the KXL-780 formulation) from the Sequoyah Nuclear Plant (TVA) and review of related documentation.
PLANT SITE APPLICABILITY:
Sequoyah 1 and 2/50-327 and 328 49
ORGANIZATION: GEORGIA INSTITUTE OF TECHNOLOGY ATLANTA, GEORGIA REPORT INSPECTION NO.-
99900903/85-01 RESULTS:
PAGE 2 of 4 A.
VIOLATIONS:
None.
B.
NONCONFORMANCES:
1.
Contrary to the requirements of Criterion V of Appendix B to 10 CFR Part 50, GIT performed activities affecting quality, i.e.,
determination of specimen irradiation dose rate, exposure-time and total dose, without documented instructions or procedures.
2.
Contrary to the requirements of Criterion V of Appendix B to 10 CFR Part 50 and the requirements of paragraph IV of the GIT Quality Assurance (QA) Manual for Hot Cell Operations, there was no documentation of the calibration of the clock normally used by GIT to measure sample radiation exposure time and thus, with dose rate, l
to measure radiation dose.
C.*
UNRESOLVED ITEMS:
None.
D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
1.
(0 pen) Nonconformance [82-02 Item B]:
GITs test records did not identify the inspector or data recorders. This item will be reviewed programatically in a future inspection.
2.
(0 pen) Nonconformance [82-01 Item B]:
GIT had not established a documented QA program. The NRC inspector noted that GIT's QA manual had been completed and also contained some implementing procedures.
Review of the GIT QA program will be completed in a future inspection.
E.
OTHER FINDINGS AND COMMENTS:
1.
With respect to nonconformance B.1. above, GIT determines the sample dose rate by placing the dose rate instrument's probe at the samples' location with respect to the exposed Co-60 array and recording the digital meter reading.
Lab barometric pressure (assumed to be the same as the hot cell) is recorded and a hot cell temperature of 29'C is assumed.
These data are reduced on a pre-printed worksheet to convert the meter reading to 50
ORGANIZATION: GEORGIA INSTITUTE OF TECHNOLOGY ATLANTA, GEORGIA REPORT INSPECTION NO.-
99900903/85-01 RESULTS:
PAGE 3 of 4 Reentgens/hr which is then converted to Rads /hr.
Rads /hr are corrected for ambient hot cell temperature (assumed) and pressure to normalize to Standard Temperature and Pressure (STP) and account for a pressure effect on the dose rate probe. The NRC inspector observed the taking of a demonstration reading and calculation of a dose rate for the cable specimens and noted that no procedure is used which describes how to obtain the raw data and perform the calculations. GIT had informally established a contention for rounding or truncating numbers and a standard number of significant ficures to be used in these calculations.
The purpose of this was, in part, to ensure that the final dose rate figure represented a conservative, minimum value. This adopted practice was not, however, prescribed by documented instructions or procedures. The inspector also noted that the calculation worksheet did not specify or provide for recording identification of the instrumentation used or the data recorder or inspector /checke. Elements of the calculation such as scale factors or probe / instrument calibration or correction factors etc. were not related in the lab record to any particular instrument. While the accumulated error resulting from not adhering to the standard practice might be less than that introduced by the maximum expected instrument errors or specified instrument accuracies, the entire procedure, including setup of the instrument, and taking and converting the readings on a form which is only partially self explanatory, constitutes an activity affecting quality, i.e., the actual value and accuracy of the result, and therefore must be controlled by procedures including appropriate acceptance criteria adequate to determine that the activity has been properly performed.
2.
With respect to nonconformance B.2. above, GIT determines or measures the total sample dose by determining the actual air equivalent dose rate as described in paragraph E.1 above, then exposing the sample for a time period calculated to achieve the desired total dose. The dose rate is chosen to fall within test plan requirements and yield a reasonable exposure time. The exposure time is measured in practice, GIT stated, by the electric wall clock in the lab adjacent to the hot cell.
(Mote that no dosimetry to measure total dose directly is used by GIT.) The test documentation and GIT procedures did not specify any particular time piece.
It is recognized that the measurement affecting quality here is elapsed exposure time rather than real or absolute time and that the accumulated error for this type of clock (assuming this same clock is always used) would likely be very small over the typical exposure period.
However, the GIT irradiation facility QA manual specifically 51
ORGANIZATION: GEORGIA INSTITUTE OF TECHNOLOGY ATLANTA, GEORGIA REPORT INSPECTION NO.-
99900903/85-01 RESULTS:
PAGE 4 of 4 required obtaining and maintaining documentation of NBS traceability of the calibration of equipment used to measure radiation dose; whereas contrary to this, there was no documentation addressing at all the accuracy or error or calibration of any time piece which might be used to measure the exposure period and hence (with dose rate) measure the total radiation dose.
3.
The NRC inspector reviewed the sponsor-prepared test plan and test documentation including the " official" dose rate work sheet, viewed the test setup and observed a demonstration of dose J
rate determination. The inspector noted that specimen handling, actual dose rate and source used were consistent with that recorded and with the test plan requirements.
No nonconformances with the test plan were identified.
52
ORGANIZATION: JOSEPH OAT CORPORATION
~
CAMDEN, NEW JERSEY REPORT INSPECTION INSPECTION NO.: 99900251/86-01 DATE:
1/13-17/86 ON-SITE HOURS:
114 CORRESPONDENCE ADDRESS: Joseph Oat Corporation ATTN: Mr. Maurice Holtz Vice President 2500 Broadway Camden, New Jersey 08104 ORGANIZATIONAL CONTACT: Mr. John Renckert, QC Manager TELEPHONE NUMBER:
(609) 541-2900 i
NUCLEAR INDUSTRY ACTIVITY: Heat exchangers, fuel storage racks, and defueling canisters bI AA YJ/ff
[R.L.Cirimberg,SpecfflProjectsInspection ASSIGNED INSPECTOR:
D' ate Section (SPIS)
V OTHERINSPECTOR(S):
C. Abbate, SPIS J. Thomas, TMI Program Office B. Brown, EG&G R. Pettis, SPIS APPROVED BY:
M
/3 k
,Cohn W. Craig, Chief, SPIS,VendorProgramBranch ae 4
I INSPECTION BASES AND SCOPE:
A.
BASES:
10 CFP Part 21 and 10 CFR Part 50, Appendix B.
B.
SCOPE: Review fabrication activities, the Joseph Oat QA and QC programs and their implementation, and compliance with the requirements of 10 CFR Part 21 and 10 CFR Part 50, Appendix B during the fabrication of fuel and krockout canisters to be used during the TMI-2 defueling effort. The QA (continued on page 2)
PLANT SITE APPLICABILITY:
TMI-2(50-320).
53 i
ORGANIZATION: JOSEPH 0AT CORPORATION CAMDEN, NEW JERSEY REPORT INSPECTION NO.-
99900251/86-01 RESULTS:
PAGE 2 of 12 B.
SCOPE:
(continued) requirements and applicability of 10 CFR Part 21 were specified in Bechtel purchase order TC-022111, Rev. O dated August 7, 1985, and Bechtel Technical Specification 15737-2-M-101A(Q), Rev. 2, dated June 18, 1985.
A.
VIOLATIONS:
None.
B.
NONCONFORMANCES:
None.
C.
UNRESOLVED ITEMS:
None.
D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
1.
(Closed) Violation (85-01, Item A.1):
Contrary to Sections 21.21(a)(1) and (2) of 10 CFR Part 21, the Joseph Oat procedure SP-1552, " Procedure for Reporting Defects and Noncompliances" dated January 15, 1979, does not adequately address a procedure to be followed by Oat personnel to inform the licensee or purchaser of a defect, and does not address the issue of informing a director or responsible officer of deviations reportable to the NRC, purchaser or licensee.
The NRC inspectors determined that procedure SP-1552 " Standard Procedure for Reporting Defects and Noncompliances" Rev. I dated December 24, 1985, the latest revision, includes a procedure to inform the licensee or purchaser of a defect and a procedure to inform a responsible officer of deviations reportable to the NRC, purchaser or licensee. This item is considered closed.
2.
(Closed) Nonconformance (85-01, Item B.1):
Contrary to Criterion V of Appendix B to 10 CFR Part 50, and Section 4.2.2 of the Oat Quality Assurance Manual, Revision 8, dated June 26, 1985, a description did not exist in the traveler for the installation and removal of lifting lugs during the fabrication of the fuel canisters.
54
ORGANIZATION: JOSEPH OAT CORPORATION CAMDEN, NEW JERSEY REPORT INSPECTION NO.-
99900251/86-01 RESULTS:
PAGE 3 of 12 During this inspection it was observed that the shop travelers for the remaining fuel canisters to be fabricated have been revised to address the installation and removal of lifting lugs.
In addition, the traveler references " Procedure for Placement of Temporary Lift Lugs on J-2476," Rev. O dated September 26, 1985.
This item is considered closed.
3.
(Withdrawn) Nonconformance (85-01, Item B.2):
Contrary to Criterion V of Appendix B to 10 CFR Part 50, and Section 4.2.2 of the Oat QAM, the shell of canister 2476A2 was formed with a steel bar and a hammer to achieve a concentric fit-up between the shell and the lower head without a description in the traveler for this operation step in the manufacturing process.
As discussed in section E.3, discussions with welders and welder /
fitters during this inspection confirmed that the term " fit-up" in the traveler is the written description for forming with a steel bar and hammer. This activity was determined to be within the skill of the individuals assembling the canisters. This item of nonconformance is withdrawn.
4.
(Closed) Nonconformance (85-01, Item B.3):
Contrary to Criterion V of Appendix B to 10 CFR Part 50, Section 4.3.3.1 of Bechtel Technical Specification 15737-2-M-101A, Revision 2, dated June 19, 1985, Turco Dy-Check Remover No. 3 was being used for cleaning canister parts and welds before obtaining Bechtel's approval for the procedure which permits the use of this cleaner.
Oat Job Procedure JP-2476-1, " Cleaning Procedure for Fuel and Knockout Canisters," was revised (October 24,1985) to permit the use of Turco Dy-Check Remover No. 3; the revised procedure was approved by Bechtel.
This item is considered closed.
5.
(Closed) Nonconformance (85-01, Item B.4):
Contrary to Criterion XVI of Appendix B to 10 CFR Part 50, Special l
j Condition 19 to Bechtel P0 TC-022111, and Section. 9.1 of Oat QAM, conflicting documentation existed which left the quality of the I
recombiner elements in the lower heads in question.
55
ORGANIZATION: JOSEPH 0AT CORPORATION CAMDEN, NEW JERSEY REPORT INSPECTION NO.-
94900251/86-01 RESULTS:
PAGE 4 of 12 The inspectors reviewed a letter to Oat from the supplier, NES Manufacturing, dated September 30, 1985, and a letter to Oat from the buyer, Bechtel, dated October 4, 1985, which provided assurance that the issue had been reviewed.
Bechtel stated that the required quantity of catalyst had been installed in the recombiner elements and that the quality of the catalyst was sufficient to function as designed. This item is considered closed.
6.
(Closed) Unresolved Item (85-01, Item C.1):
A chemical reaction between water and the concrete resin was observed to take place within one hour of the concrete resin being placed in water. This reaction weakened the concrete resin and led to questions concerning the suitability of the material in the fuel canisters when exposed to water.
The inspectors reviewed a letter dated November 21, 1985, from Bechtel to Oat. This letter contained Babcock & Wilcox (B&W) test results which stated that the concrete resin met design specifications after being in contact with water. This item is considered closed.
7.
(Closed) Unresolved Item (85-01, Item C.2):
Upon inspection of the catalyst in the lower heads supplied by NES, there were indications of overheating and weight discrepancies.
(The resolution of the weight discrepancies was discussed in nonconformance 85-01, Item B.4.)
The inspectors reviewed a letter from Bechtel to Oat dated November 21, 1985 which stated that catalyst discoloration from heating was not expected to be a concern. As discussed in Item E.7, this area has been evaluated. The conclusion of the evaluation is that the catalyst was acceptable for use. These two items are considered closed.
8.
(Closed) Unresolved Item (85-01, Item C.3):
After shipping canister shells from NES to Oat, it was determined that 40 longitudinal welds had one radiograph view missing, and one loagitudinal weld had no radiographs.
The NRC inspectors verified that Oat was reradiographing the sections of pipe with the missing radiographs as the individual canisters were needed for fabrication. This item is considered closed.
56
ORGANIZATION: JOSEPH OAT CORPORATION CAMDEN, NEW JERSEY REPORT INSPECTION NO.
99900251/86-01 RESULTS:
PAGE 5 of 12 9.
(Closed) Unresolved Item (85-01, Item C.4):
After reviewing the concrete resin pour, it was determined that the procedure did not address whether continuous pouring is required.
The inspectors reviewed Oat procedure QC-82, Rev. 3, dated January 8, 1986, " Concrete Pouring Procedure" which states that pouring will occur within one hour after mixing. This item is considered closed.
10.
(Closed) Unresolved Item (85-01, C.5):
The applicability of Joseph Oat invoking the provisions of 10 CFR Part 21 upon Air-Oil Systems Incorporated was an unresolved item.
After discussion with Oat staff members, it was determined that these items were standard catalogue listed fittings and that the modifi-cations, although listed on the purchase order, were made by Oat and not by Air-Oil Systems Incorporated. This item is considered closed.
E.
INSPECTION FINDINGS AND OTHER COMMENTS:
1.
10 CFR Part 21 During the inspection the NRC inspectors reviewed procedure SP-1552,
" Procedure for Reporting Defects and Noncompliances," Revision 1 dated December 24, 1985. This revision included the delineation of responsibilities of Oat personnel when a nonconformance or defect is found at various stages of the manufacturing process. The procedure also described the steps to be followed when a defect or noncon-formance is identified after shipment including customer notification.
Records of such notifications will be kept by the QC Manager.
As a result of this review, Oat's procedure was determined to be adequate; Violation 85-01-01 is considered closed.
2.
Lifting Lugs The NRC inspectors reviewed shop travelers for fuel canisters at various stages of fabrication and found that a revision had been made to include the steps for installation and removal of lifting lugs.
The traveler also referenced " Procedure for Placement of Temporary Lift Lugs on J-2476," Revision 0 dated September 26, 1985. The procedure included an outline of the welding procedure to be used, and a drawing which indicated where the lifting lugs are to be placed.
57
ORGANIZATION: JOSEPH 0AT CORPORATION CAMDEN, NEW JERSEY REPORT INSPECTION NO.-
99900251/86-01 RESULTS:
PAGE 6 of 12 Standard Procedure SP-1609 dated July 6, 1978, is used for the removal of the lugs. A liquid penetrant examination of the area where the lifting lug had been welded to the canister is then performed.
As a result of this review, Nonconformance 85-01-01 is considered closed.
3.
Control of Special Processes The NRC inspectors reviewed 14 Oat procedures to determine whether special processes were being conducted by qualified personnel using qualified procedures and equipment. A review of 12 travelers relating to fabrication of 10 fuel canisters and two knock-out canisters revealed that all individual operations were properly initialed and dated. The traveler package contained a weld and heat sketch and record sheet which identified bill of material item numbers, heat numbers, weld numbers, welding procedure specification numbers, welder stamp, and filler metal heat numbers. Operation descriptions in the traveler described fabrication steps and identified the approved drawings which contained weld notes and joint details.
While observing the fit-up of lower heads to several canister shells the NRC inspectors questioned the Oat welders and welder / fitters concerning what description in the traveler described the use of a steel bar and hammer on the canister shell. All Oat personnel stated that the term fit-up, referenced in the travelers, allows the necessary forming to achieve the proper dimensions specified on approved drawings. Also, the inspectors noted that this process is consistent with fabrication requirements and within the skill of the individuals assembling the canisters. Based upon the above, Nonconformance item 85-01-02 is withdrawn.
4.
Canister Cleanliness Canister shells are stored indoors at the Oat facility under clear plastic sheets with rubber end caps on the empty shells. The cold weather has resulted in split end caps which are being taped on an as needed basis.
Cleaning operations are described on fabrication travelers by reference to Job Procedure No. JP-2476-1, Rev. 3.
Cleaning is being performed in accordance with written procedures.
The NRC inspectors reviewed JP-2476-1, Rev. 3 and determined that Turco Dy-Check Remover No. 3 is permitted for use by this procedure which is on the list of procedures that have been approved by Bechtel.
Nonconformance 85-01-03 is considered closed as a result of this review.
58
ORGANIZATION: JOSEPH OAT CORPORATION CAMDEN, NEW JERSEY REPORT INSPECTION NO.-
99900251/86-01 RESULTS:
PAGE 7 of 12 5.
Recombiner Catalyst During the previous inspection a nonconformance was identified involving conflicting documentation supplied to Oat from both Bechtel and NES concerning the recombiner catalyst. After identification of this nonconformance, Joseph Oat issued Deviation Notice 2961 dated September 27, 1985. This documented the condition which was adverse to quality and initiated actions to resolve the discrepancy resulting from the conflicting documents.
The inspectors reviewed a letter to Oat from NES Manufacturing, dated September 30, 1985, and a letter to Oat from Bechtel dated October 4, 1985. These letters provided assurance to Joseph Oat that the issue had been reviewed and certified that the required quantity of catalyst had been installed in recombiner elements.
In addition, this issue was reviewed by the NRC Three Mile Island Cleanup Project Directorate (TMICPD) which has lead regulatory responsibility for safety evaluation of the defueling canister design; and GPU Nuclear Corporation, the end user of the defueling canisters.
Acceptable resolution of the catalyst loading issues is discussed in NRC letter NRC/TMI-85-086 dated November 8, 1985 to GPU Nuclear Corporation, docket 50-320, and GPU Nuclear letter 4410-85-L-0229 dated November 12, 1985 to NRC:TMI Program Office. This nonconformance is considered to be an isolated incident of personnel error, thus no further corrective action is deemed necessary to prevent recurrence.
Nonconformance 85-01-04 is considered closed.
6.
Durability of Concrete Resin The inspectors reviewed a letter from Bechtel to Joseph Oat dated November 21, 1985 which was in response to a Joseph Oat letter of October 7, 1985 which addressed the issue of concrete resin durability.
The letter transmitted a test report by the canister designer (Babcock
& Wilcox) and confirmed that the refractory cement is acceptable for use and meets the design specifications. Testing performed by the designer determined that the cement exhibits considerable reduction in compressive strength after immersion in water, but still retains sufficient strength to exceed the design specifications for the. fuel canister.
l In addition, the NRC TMICPD staff reviewed the results of l
additional testing of the refractory cement performed by Rockwell l
Hanford Operations which verified that any material leached from the cement by immersion in water would not adversely affect catalytic recombiner operation. Based on the above, Unresolved Item 85-01-01 is considered closed.
59
ORGANIZATION: JOSEPH OAT CORPORATION CAMDEN, NEW JERSEY REPORT INSPECTION NO.-
99900251/86-01 RESULTS:
PAGE 8 of 12 7.
Inspection of Recombiner Catalyst in Lower Heads The inspectors reviewed a letter from Bechtel to Joseph Oat dated November 21, 1985 which informed Oat that catalyst discoloration from heating had also been observed at other canister vendors. This letter stated that discoloration was not expected to be a concern but was being evaluated, and Oat should proceed with catalyst installation. This issue was subsequently reviewed by the NRC TMICPD technical staff with GPU Nuclear and Rockwell Hanford Operations, who designed the recombiner elements.
It was determined that catalyst heating due to welding of the retention screens would not adversely affect the Englehard catalyst.
Significant heating of the AECL catalyst would degrade the catalyst but would also cause an obvious charred appearance to the silicone coating which had not been observed. Based on the above, Unresolved Item 85-01-02 is considered closed.
8.
Nondestructive Examination (NDE)
Nondestructive examinations of the TMI-2 Defueling Canisters must meet the requirements set forth in the 1983 Edition of ASME Code Section VIII, Subsection UW (lethal), paragraphs UW-50, UW-51, and UW-53 as required by Bechtel Specification 15737-2-M-101A(Q).
a.
Since the previous inspection conducted on September 23-27, 1985, several NDE procedures were revised and/or added to the list of Bechtel approved Oat procedures.
The following NDE procedures were reviewed and found to meet the applicable Code requirements:
1.
" Ultrasonic Inspection Procedure for Cire. Pipe Weldments,"
QC-2476-70, Rev. 2, dated January 15, 1986. This procedure was revised and added to the list of approved procedures.
Revision 2 now contains ultrasonic instrument gain settings for scanning, search unit overlap of at least 10%, and the rate of search unit movement for examination not to exceed six inches /second unless calibration is verified at increased scanning speeds.
2.
" Ultrasonic Examination Procedure, Addenda #1, to QC-2476-70,"Rev.1, was written and approved to govern the ultrasonic examination of the base materials for the bulkheads on the fuel canisters and the top heads on the knockout canisters.
60
ORGANIZATION: JOSEPH OAT CORPORATION CAMDEN, NEW JERSEY REPORT INSPECTION NO.-
99900251/86-01 RESULTS:
PAGE 9 of 12 3.
" Procedure for Qualification of Liquid Penetrant Examination Techniques at Nonstandard Temperatures," dated September 14, 1976. This procedure was used to qualify QC-2476-10,
" Liquid Penetrant Examination of Welds," Rev. O, dated July 25, 1985, for liquid penetrant examinations with component and material temperatures down to 40 F.
With the penetrant procedure certified for 40*F, it was still necessary to preheat the components and materials before penetrant examinations could be performed at the fabrication area of Oat during this NRC inspection, b.
Ultrasonic examinations are either performed under subcontract by Eastern Testing and Inspection, Inc. (ETI) of Pennsauken, NJ, or by an Oat Level II ultrasonic examiner.
Individual training and certification records, including eye examination results, were reviewed for one Oat Level 11 examiner and one ETI contractor.
During discussions, Oat personnel stated that these are the only two Level 11 examiners which will be used for ultrasonic exami-nation of canisters. The certifications for each examiner were found to be in compliance with SP-1576, " Requirements for Qualification and Certification of NDE Personnel" (Oat), Rev. 4, dated. September 1,1983; CP-101, " Procedure for Qualification and Certification of NDE Personnel" (ETI), Revision 3, dated June 1, 1981; and the guidelines of SNT-TC-1A.
c.
Radiographs, technique sheets, and inspection reports for the following bottom-head welds were reviewed: J-#2476-C61, J-#2476-C62, J-#2476-C63, J-#2476-C64, J-#2476-C65, J-42476-C66, and J-#2476-C68.
It was reported that these were the only bottom-head weld radio-graphs available since the previous radiographs had been shipped with the completed canisters.
No rejectable indications were noted. The inspection reports, the radiograph technique, quality and densities were determined to meet the applicable ASME Code requirements.
d.
An in-progress liquid penetrant examination on canister pipe 116P-2 at the Oat fabrication area was observed. Although the temperature at the fabrication area was less than 40 F, the materials and parts were preheated to a temperature greater than 40 F during examination.
Liquid penetrant examinations on the weld prep areas of knockout canister top-heads D-61, D-36, D-32, D-13, and D-45 were also observed in the inspection area.
All of 61
ORGANIZATION: JOSEPH OAT CORPORATION CAMDEN, NEW JERSEY l
REPORT INSPECTION NO.-
99900251/86-01 RESULTS:
PAGE 10 of 12 the above liquid penetrant examinations were done by certiffed Level II personnel using approved procedures and certified:
materials.
During the inspection, interviews were held with Oat personnel e.
regarding the missing radiographs on canister pipe longitudinal welds.
Each full length of pipe was radiographed with twenty location markers. The pipes were then subsequently cutin half.
NES or B&W received one half of the pipe and maintained the radiographs applicable to their piece. Apparently, this resulted in one view missing since the cutoff point on the pires is at the midsection of the 10-11 location markers. Dat is reradiographing the sections of pipe with missing radiographs as the individual canisters are fabricated. The following replacement radicgraphs were reviewed and found acceptable:
34P-2, 47P-2, 48P-1, 63P-1, 20P-2, 30P-2, 40P-1, 118P-2, 121P-1, 56P-1, 83P-2, 125P-1, and 140P-1.
Unresolved Item 85-01-03 is considered closed as a result of the above discussions.
9.
Concrete Resin Pouring The NPC inspectors reviewed information sent to Oat from Bechtel concerning the concrete resin.
This information included a B&W test l:'
report dated October 10, 1985.
The report outlines the curing process and states that one day curing at room temperature forms the t
hydraulic bond and sets the shape. Curing continues 20-30 days after pouring at which time the bond is at full strength.
Dat procedure QC-2476-82, " Concrete Resin Pouring Procedure," Rev. 3 dated January 3, 1986 was reviewed and was determined to be acceptable.
Section 4.1 states that pouring will begin 15 minutes after mixing and I that the mixture will be used within one hour of mixing. Thus, the procedure allows a delay in pouring (less than one hour) which would r.ot result in deterring the curing process.
Unresolved Item 85-01-04 is considered closed as a result of this review.
10.
Procurement Document Control At the time of the inspection, no purchase orders (P0s) had been written for material needed for the fabrication of fuel or knockout canisters. However, two P0s had been written for machining services.
P0 19643 to Permutit Co. was for machining work to be done on material d
62
ORGANIZATION: JOSEPH DAT CORPORATION CAMDEN, NEW JERSEY REPORT INSPECTION NO.-
99900251/86-01 RESULTS:
PAGE 11 of 12 supplied by Oat. Bechtel Specification 6400, Rev. 2 and 10 CFR 21 were imposed on Permutit and certificates of conformance were required.
P0 19518 to Beaver Tool was a change notice to the original P0 dated September 9, 1985. The change notice called for machining to be l
performed on the bulkhead weld prep and pipes supplied by Oat. All requirements on the original P0 applied to the change notice. These two suppliers are on the Oat Active Qualified Suppliers List.
The inspectors reviewed Bill of Material (B0M) M-7949 for the fabri-cation of 107 knockout canisters. This material was shipped from NES Manufacturing's facility in Greensboro, NC to Oat. The inspectors verified that the B0M had been prepared by Oat from the Bechtel purchase order, technical specification, and drawings.
The inspectors discussed the issue of Oat invoking 10 CFR Part 21 on Air-Oil Systems Incorporated with Oat staff.
Purchase order 19515 indicated that the Hansen pipe fittings were to be modified and therefore 10 CFR Part 21 would be applicable since these commercial grade items were subject to specification requirements unique to the canister application.
However, Rev. I to P0 19515 deleted the requirement for modification of the part, and materials actually supplied to Oat were standard catalogue fittings. The modifications were subsequently made by Oat in accordance with the QA plan.
Unresolved Item 85-01-05 is considered closed as a result of the above discussions.
- 11. Lower Heads On January 6,1986, Joseph Oat received a letter from NES Manufacturing indicating that ten canister lower heads may have been sent to Oat without the required welding and NDE traceability, or contained catalytic recombiners that had been contaminated with liquid penetrant inspection materials.
The letter identified the subject heads by NES ID numbers. Oat notified Bechtel of the potential defect that same day.
By January 9, 1986, Oat had completed a review of records and determined that four of the heads had been used in fabri-cation, notified Bechtel, and requested further instruction on disposition of the matter. Oat also issued a deviation notice that would initiate action to prevent further use of any of the identified heads with potential defects.
The inspectors reviewed the correspondence relating to this matter and determined that Joseph Oat's actions were prompt, proper, and in full conformance with the QAM.
63
ORGANIZATION: JOSEPH OAT CORPORATION CAMDEN, NEW JERSEY REPORT INSPECTION NO.-
99900251/86-01 RESULTS:
PAGE 12 of 12 F.
PERSONS CONTACTED:
Joseph Oat Corporation
- M. Holtz
- J. Benckert C. Leonard B. Badaili J. Boyer J. Muschek B. Hutchinson M. Madera i
R. Ewing Bechtel R. Emrich T. McKearney Hartford Steam _ Boiler p
A. Mcdonald GPU J. Campbell s
c.
'i i
~
- attended exit meeting 64 l
ORGANIZATION: LIMITORQUE CORPORATION LYNCHBURG, VIRGINIA REPORT INSPECTION INSPECTION NO.: 99900100/85-01 DATE(S):
11/4-8/85 ON-SITE HOURS:
116 CORRESPONDENCE ADDRESS: Limitorque Corporation ATTN: Mr. F. K. Denham Executive Vice President 5114 Woodall Road Lynchburg, Virginia 24502 OPGANIZATIONAL CONTACT: Mr. P. McQuillan, QA Administrator TELEPHONE NUMBER:
(804) 528-4400 PRINCIPAL PRODUCT: Motor and manual operated valve actustors.
NUCLEAR INDUSTRY ACTIVITY: Approximately 5% of sales.
Ne f ASSIGNED INSPECTOR: & k. Li\\um,
1 - li-86 T. Conway, Rea tive Inspection Section (RIS)
Date OTHER INSPECTOR (S):
E. Yachimiak Jr., RIS L. Vaughan, Program Coordination Sectlon G. Hubbard, Equipment Qualification Inspection Section T. Tinkle, Consultant
,\\,. k h44rb
/
2 - li -E.C APPROVED BY:
Ef] Merschoff, Ch f, RIS, Vendor Program Branch Date INSPECTION BASES AND SCOPE:
A.
BASES: 10 CFR Part 50, Appendix B and 10 CFR Part 21.
B.
SCOPE: This inspection was made as a result of the receipt of 10 CFR Pirt 50.55(e), Licensee Event (10 CFR 50.73), and 10 CFR Part 21 reports
~
l from:
(1) Gulf States Utility Company and Niagara Mohawk Power Corporation relating to the loss of torque of Reliance motors containing i
magnesium rotors, (2) Washington Public Power Supply System pertaining l
(continued on page 2) l PLANT SITE APPLICABILITY:
Loss of torque-River Bend (50-458/459) and Nine Mile Point 2 (50-410); Valve-actuator weight discrepancies-WNP 1&4(50-460/513);
(continued on page P) 65
ORGANIZATION: LIMITORQUE CORPORATION LYNCHBURG, VIRGINIA REPORT INSPECTION N0.: 99900100/85-01 RESULTS:
PAGE 2 of 21 INSPECTION BASES AND SCOPE:
(continued)
B.
SCOPE:
to valve / actuator weight discrepancies, (3) Public Service Electric and Gas Company on replacement actuators having different spring packs, and (4) Commonwealth Edison and Tennessee Valley Authority on qualification of internal wiring.
PLANT SITE APPLICABILITY:
(continued)
Replacement actuators-Salem 2 (50-311); Valve Operator Wiring Quali'ication-Sequoyah 1&2 (50-327/328), Zion 1&2 (50-295/304), Dresden 2&3(50-237/249),
Quad Cities 182 (50-754/265) and any other plant with limitorque operators.
A.
VIOLATIONS:
1.
Contrary to Section 21.21(b) of 10 CFR Part 21, Limitorque did not notify the NRC that unqualified Baldor motors were used in safety related valve actuators installed in the WNP-2 and Salem nuclear power plants as noted by the following (85-01-01).
Limitorque discovered on March 28, 1983 that unqualified Baldor motors were installed in actuators for four Crane valves at the Public Service Electric and Gas Salem Nuclear Power Plant.
The Limitorque Bill of Material indicates that the actuators were safety related.
Limitorque was advised by Washington Public Power Supply System (WPPSS) on May 27, 1983 that unqualified baldor motors were installed in actuators for four Velan valves at the WNP-2 Nuclear Power Plant.
The Limitorque Bill of Material indicates the actuators were safety related.
Following the January-February 1984 NRC inspection, Limitorque initiated a record review to determine the number of actuators shipped with unqualified Baldor motors.
In August 1984, Limitorque discovered unqualified Baldar motors were also installed in actuators for four Borg-Warner valves at WNP-2.
The Limitorque Bill of Material indicates the actuaters were safety related.
2.
Contrary to Section 21.51(b) of 10 CFR Part 21, limitorque was unable to provide documented evidence that required reviews and evaluations were performed for possible 10 CFP, Part 21 repceting on the following safety related actuators with noted problems (85-01-02).
i l
l i
66
i ORGANIZATION:
LIMITOROUE CORPORATION LYNCHBURG, VIRGINIA REPORT INSPECTION NO.: 99990100/85-01 RESULTS:
PAGE 3 of 21 a,
Fotor pinion key fallures described in Section E.4 of NRC Inspection Report No. 99900100/84-01 (ref. Limitorque letter dated September 27, 1982 to the NRC).
b.
Worm shaft gear failures (ref. Limitorque letter dated August 13, 1985 to the NRC).
c.
Potential field maintenance problem on type H3BC actuators (ref. Limitorque letter dated January 10, 1983 to the NRC),
d.
Cracked / broken limit switch rotors (ref. Limitorque letter dated February 21, 1984 to the NRC).
3.
Contrary to Sections 21.6 and 21.21 of 10 CFR Part 21, it was noted that (85-01-03):
a.
Section 206 and a " Notice" were not posted in the shop fabrication area to be read by the manufacturing personnel, and a " Notice" which was posted in an adjacent building did not describe the regulation or state where Part 21 reports could be examined.
b.
Procedure QCP-22 " Reporting Defects for Safety Related Equipment..." did not provide for informing the licensee or purchaser and a responsible officer of a deviation /
defect in a basic component.
B.
NONCONFORMANCES:
1.
Contrary to Criterion V of Appendix B to 10 CFR Part 50, Subsection A.1 of Section I of the Quality Assurance Manual (QAM), and Section 5 of ANSI N45.2, a review of 13 P0s to vendors (Ryerson Steel-two, Copperweld Steel-one, Walker Machine-two Southern Centrifugal-two, Lynchburg Foundry-two, foster Electric-three, and Wesco-one) indicated that quality requirements (e.g., QA Program) were not passed on to these suppliers and manufacturers of compcnents destined i
for actuators purchased by nuclear customers as safety relateditems(85-01-04).
2.
Contrary to Criterion V of Appendix B to 10 C'FR Part 50 and Section 2 of ANSI N45.2, a review of the QAM, Revision 2 dated January 4, 1984 indicated that Limitorque did not update the QAM to address training of non QC personnel in disciplines such as engineering, parts and shipping, purchasing, order processing, and field service (85-01-05),
67
ORGANI?ATIONi LIMITOROUE CORFORATION LY.NCliBURG, VIRGINIA y
REPORT INSPECTION NO.: 99900100/85-01 RESULTS; PAGE 4 of 21 3.
Contrary to Criterion y of Appendix B to 10 CFR Part 50, Subsect:fon B-1.b of Sectico VI 6f the QAM, Section 11.A of Procedure No. QCP.-11, acid Sect.ic~ns 5.2, 9.6.1 ond 9,7.3 of SNT-TC-1A, a review cf Limitorcue% writt6n pr.actice for nondest~ructive examination (NDE) and qualifi?ation rec 0rds for three NDE personnel revealed that (26-01-06):
)
a.
Procedure No. QCP-11 did not describe the responi;ibility of each level of certification or rules covering the
. duration of interrupted service requir169 re-examiiation i
and recertifichtien, b..
The qualification rec 6rds of the three NOE personeel did not i:oritain a statem6nt indicating satisfactory ccTletion of training in accordar.ce with Proce6.ure No. OCP-11, 4.
Contrary to Criterion V of Apperdix.C to 10 CFR Part 50, subsc-ction W.1 of Section I of the QAM, and Section 14,A.2 of Procedure No.
QCP-14, the inspettar reviewed checklist fonii No. QCD019 for internal audits conducted in 1063 and 1984, and it was noted that the ares of QA Records (i.e.,Section XI pf the QAM) was not in the checklist and was not audited ( 85-01 07)i I
A Contrary to Crit.erton V of Appendix B to 10 CFR Part 50, Subsection BJ ef Section IH of the QAM, Sectio.n 3.E of ' Procedure ND. QCP-3, and Section 25. A of Procedure No. QCP-25, it was noted (85-61-08):
a, llendor Evsiuetion Reports (VI'R) were not generated semi-annually for the following Vendors:
gndor VEP. Peried*
Petriest-Vinsmith D)
Relience Electric Co.-
(3)
Electric Apparatus Cc.
(1)(2)(3)
Dreyer Co.
(1)(2)(3)
Wesco (35 Philadelphia Bear Corp.
(2)(3 l
Sovereign 14etal Corp.
(1)(2(3)
Valley Fastene'es (1.)(2 (3) n 1
ORGANIZATION: LIMITORQUE CORPORATION LYNCHBURG, VIRGINIA REPORT INSPECTION N0.: 99900100/85-01 RESULTS:
PAGE 5 of 21 Vendor VER Period *
(1)((2)(3)
Tubular Steel Inc.
(1) 2)(3)
NES/Selamco Foster Electric (1)(2)(3) b.
Vendors listed below with an evaluated rating below limitorque's quality standard were not removed from the AVL.
Vendor Rating VER Period
- Advance Pressure Casting 20% reject (3) 100% reject (2) 100% reject (1)
Bronze & Plastics Specialty 70% reject (3)
Duer Spring & Manufacturing Co.
45% reject (3)
International Spring 43% reject (3)
Newcomb Spring 56% reject (3) 40% reject (2)
Southwire Machine Div.
95% reject (3)
,The Spring Work 56% reject (3)
Guilford Foundry Co.
55% reject (3) 8% reject (2)
Handwheel Inc.
50% reject (3)
American Spring & Wire 91% reject (2)
Rocky Mountain Casting 96% reject (2)
- (1) - 1/84-6/84 (2) - 7/84-12/84 (3) - 1/85-6/85 69
ORGANIZATION: LIMITORQUE CORPORATION LYNCHBURG, VIRGINIA REPORT INSPECTION N0.: 99900100/85-01 RESULTS:
PAGE 6 of 21 c.
Audits were not performed on Page-Wilson Corporation and Hartzog Granite Company who calibrated the hardness testers and surface plates, respectively.
6.
Contrary to Criterion V of Appendix B to 10 CFR Part 50, Procedura No. QCP-22, and Procedure No. QCP-13, Limitorque is not effectively implementing QA program requirements for reporting defects in safety related equipment as noted by the following (85-01-09):
A field service report was prepared in May 1983 when Limitorque field service personnel installed qualified Reliance motors in place of unqualified Baldor motors in safety related actuators for four Crane valves at Public Service Gas and Electric's Salem Nuclear Power Plant. Also in May 1983, a customer problem report was made when WPPSS advised Limitorque service and special project personnel that unqualified Baldor motors were installed in safety related actuators for four Velan valves at WNP-2. The Technical Manager did not see the field service report or the problem report until he reviewed the Order File as part of the record review initiated after the unqualified Baldor motor problem was raised as an issue during the January-February 1984 NRC inspection.
The record review was completed in August 1984 which is approximately 15 months after the problem was reported to Limitorque personnel.
The Quality Assurance Administrator could not provide a copy of his letter to the Design Review Committee concerning the unqualified Baldor motor problem nor a copy of any letter from him to the Design Review Committee for any problems for evaluation and investigation as required by Procedure No.
QCP-22.
The Quality Control Manager could not provide copies of any documentation showing that the Material Review Board evaluated the circumstance concerning the installation of unqualified Baldor motors in the actuators for four Crare valves at the Salem Nuclear Power Plant. This problem was identified by the QC Manager during a review of QA records for the particular order.
7.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Section 17 of ANSI N45.2, Limitorque discovered on March 28, 1983 that unqualified Baldor motors were installed in four safety related actuators for Crane valves at Public Service Electric and Gas' Salem Nuclear Power Plant.
- Further, Limitorque was advised by WPPSS on May 27, 1983 that 7U
ORGANIZATION: LIMITORQUE CORPORATION LYNCHBURG, VIRGINIA REPORT INSPECTION NO.: 99900100/85-01 RESULTS:
PAGE 7 of 21 1
i unqualified Baldor motors were installed in four safety-related actuators for Velan valves at WNP-2.
For the two cases in point, Limitorque could not provide any documented evidence that measures were taken to identify the cause of the condition, the extent of the problem, and necessary corrective action to preclude repetition, or that the condition was reported to appropriate levels of management (85-01-10).
C.
UNRESOLVED ITEMS:
Limitorque was requested to provide the Order Files for the blocking valve actuators involved in the " Salem Unit 2 Depressurization Event" discussed in NRC memo AE0D/E505 dated Jdly 25, 1985. The QA Administrator stated that actuator unit (P0 875064, S/N 355494) documentation should be in Orcer File 3H8897. However, the QA Administrator stated the file could not be found. The QA Administrator further stated the d0cumentat' ion for the actuator unit on P0 927970 could not be identified without either the order file number or the unit serial number and ?. hat Limitcrque could not retrieve this information. This itert uill be evaluated during a future NRC inspection.
D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
e 1.
(Closed) Nonconformance (Itect A, B4-01); Limitorer.:e ccrrective action commitments ctade in a letter dated Novenber 15, 1962, in response to Item A of the Notice of Nanconformance of NPC Inspection Report No. 99900100/R2-02, were found to be iradequate in that the training of men-QC personnel did not include:
a.
All applicable personnel having quality functiont (e.g., the n:anagers bf Special Processes and Industrial e gineering).
n i
l b.
All applicable Quclity Control Procedures (QCPs) were l
not addressed e.9., OCP-3 ter purchasing persennel t t
l QCP-13 for ent,iceering personrel: and OC?-23 for shipping l
personnel.
c.
Any of the applicable sections of the QA manual for any of the personnel.
71
ORGANIZATION: LIMITORQUE CORPORATION LYNCHBURG, VIRGINIA REPORT INSPECTION N0.: 94900100/85-01 RESULTS:
PAGE 8 of 21 The NRC inspector reviewed these areas and found that corrective action documented in L$miterque's April 2, 1984 letter was adequately implernented.
However, limitorque has not revised their QAM to address training of non-QC personnel (see Nonconformance 85-05-01).
a.
The Special Processes Manager was certified to his understanding of QCP-12 on April 3,1984.
The Industrial Engineer was not certified because his presence is not quality related.
b.
QCP-3 is written for inspection personnel and defines the Insoection Sampling Plan, writing of Variation Poports, etc.
end is not related to purchasing managers.
QCP-13 is written for the Quality Control Manager to assist him in preparing the Material Review Board report and has no affect on the functions of the Engineering Department.
QCP-23 is writien for shipping personnel thus the Shipping Department Supervisor and the Stockroom Supervisor were formally certified to this procedure on April 3, 1984.
c.
Limitorque's respons,e was that the QCPs address "how" quality related functions are accompitshed, whereas the QA manual addresses "what" quality related functicns are carried out by the company.
Formal training in the QCPs for applicable personnel is conducted.
2.
(Closed) Nenconformance (Ctem G, 84-01): Limitorque corrective action commitments made in a letter dated May 23, 1983 stated that key material
"...is presently procured with certified mateiial test reports...".
Contrary to this statement AlSI 1018 3/32" keys were found in the stock room without their respective certified material test report on file.
The NRC inspector reviewed this item and fcund that the certified m4terial test reports for the 1018 3/32" keys are now on file.
3.
(Closed) Nonconformance (Item C, 84-01): Limitorque corrective action cpaituents made in a letter dated November )$.1982 stated that they would "... revise QA manual with requiremnts for Quality Control to monitor safety-related material contracts." Contrary to the above, the revisityps identified had not been accomplished.
1 72
ORGANIZATION: LIMITORQUE CORPORATION LYNCHBURG, VIRGINIA REPORT INSPECTION N0.: 99900100/85-01 RESULTS:
PAGE 9 of 21 Limitorque has been monitoring safety-related contracts since October 2, 1982.
The formal inclusion of this into a procedure to insure the monitoring of safety-related material contracts has been done and is documented in Procedure No. QCP-21.
4.
(0 pen) Nonconformance (Item D, 84-01): Contrary to Criterion XVI of Appendix B to 10 CFR Part 50, measures were not taken to assure conditions adverse to quality were identified and corrected.
Limitorque corrective action commitments made in a letter dated April 2,1984 in response to the Nonconformance indicates that a review of records would be accomplished to substantiate that the shipment of unqualified motors in actuators for Velan valves for WPPSS was an isolated occurrence. The results of the record review which was completed around August 1984 indicated that the shipment of unqualified motors to WPPSS for Velan valves was not an isolated incident and that two additional occurrences existed.
This item will be reviewed during a future NRC inspection.
5.
(Closed) Nonconformance (Item A, 83-01): Commitments have not been accomplished for establishment of an inspection procedure to identify the RPM acceptance tolerance, and QA manual revision to include a requirement that inspection procedures identify accept / reject criteria.
Limitorque's response of April 2,1984 stated that Form LP-11 was being revised to include the acceptance criteria for operator motor output speed. This was verified by the NRC inspector as having been completed. A generic requirement, where applicable, for accept / reject criteria was also included into the QCPs.
6.
(Closed) Nonconformance (Item F, 83-01): NDE personnel were not qualified in accordance with SNT-TC-1A, the proper liquid penetrant (PT) dwell and development times were not implercented, and the guidance of a minimum 10-minute development time was not followed.
A review of NDE records indicated that one Level III and two Level II personnel were qualified to perform PT examination in June 1983.
Training and examination were performed by Pittsburgh Testing Laboratories in accordance with SNT-TC-1A (1980 Edition).
PT examination of components for nuclear actuators was not being 73
ORGANIZATION:
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PAGE 10 of 21 conducted on two occasions when the inspector witnessed activities performed at final inspection station No. 1.
Revision 9 to PT Procedure No. QCP-11 was on file, and a " Flow Chart for PT" was attached near the PT station.
The chart, which was signed by the Level III and the QC Manager and dated April 30, 1985, identified penetrant dwell and development times.
7.
(Closed) Nonconformance (Item G, 83-01): Hardness testers were checked only once every three months and unacceptable readings were obtained when calibration checks were performed at the request of the NRC.
During a walkthrough, calibration checks of three hardness testers were performed at the request of the NRC inspector. Acceptable readings were obtained on the microhardness tester in the Special Process Department (SPD) and two Rockwell testers, one in SPD and the other in Station No. 1.
E.
OTHER FINDINGS OR COMMENTS:
1.
Qualification of Electric Wiring in Limitorque Valve Operators The NRC inspector reviewed documentation and held discussions with Limitorque personnel to determine the basis for the environ-mental qualification of internal wiring for Limitorque valve operators. These activities were conducted as a followup to a 10 CFR Part 21 report from Commonwealth Edison, concerning the discovery that four Limitorque valve operators at the Zion nuclear plant had jumper wires different than the wires which Limitorque claimed were qualified for use in their valve operators.
Additionally, the Tennessee Valley Authority (TVA) notified the NRC l
in October 1985 that they were unable to adequately document the l
environmental qualification of internal jumper wires on 214 j
Limitorque valve operators at Sequoyah Nuclear Plant (SNP).
l Therefore, TVA was going to replace all jumper wires in Limitorque i
valve operators with wires containing documanted environmental qualification.
In conjunction with this inspection at Limitorque, a vendor inspector and one NRC Region II inspector visited TVA offices t
in Knoxville, Tennessee on October 29, 1985 and SNP in Daisy, Tennessee on October 30, 1985 to obtain information concerning TVA's valve operator rewiring activities and to followup the 1
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ORGANIZATION: LIMITORQUE CORPORATION LYNCHBURG, VIRGINIA REPORT INSPECTION N0.: 99900100/85-01 RESULTS:
PAGE 11 of 21 information at Limitorque. As part of the effort at TVA, three valve operators for SNP valves (2-FCV-67-146, 1-FCV-63-175, and 1-FCV-3-87) were selected and nhysically inspected at SNP to determine what jumper wires were installed in the operators.
The physical inspection determined that Rockbestos Firewall SIS (switchboard) wire was installed in operators for valves 2-FCV-67-146 and 1-FCV-3-87 as well as Carolina Wiring Cable Company (CWCC) wire. The wiring in the operator for valve 1-FCV-63-175 was identified as Teledyne Thermatic and AMP L-79F 125.
Shop order numbers and operator serial numbers for the three operators were recorded to trace through Limitorque's documentation to determine what wiring had been installed in the operators at Limitorque.
At Limitorque, the inspe'ctor reviewed Bill of Materials (BMs),
drawings, and final inspection and test reports for the three SNP operators and determined that Rockbestos Firewall SIS wire had been installed by Limitorque in the operator for SNP valves 2-FCV-67-146 and 1-FCV-3-87.
Limitorque stated that the CWCC wires were added by someone else.
During the TVA visit, the inspectors determined that TVA had probably installed the CWCC wires. Limitorque had no record of the type of wires they had installed in the operator for valve 1-FCV-63-175 since it had been purchased as a commercial operator (non nuclear) by Walworth Valve Company in mid-1973.
Limitorque does not maintain records that would identify the type of wire installed in commercial (non nuclear) operators.
Limitorque further stated that they were unfamiliar with the Teledyne Thermatic or AMP wires.
Qualification of the operator for valve I-FCV-63-175 will be reviewed during a future NRC inspection at SNP.
Based on discussions with Limitorque personnel and review of Limitorque test reports, the inspector determined that Limitorque does not specifically address wiring or wiring qualification in their qualification test reports'.
They consider their valve operators sold as safety related nuclear to be qualified (including internal wiring), if either Raychem Flametrol or i
Rockbestos Firewall SIS (Rockbestos specification RSS-4-002) jumper wires are installed in the operators. Qualification of these wires for use in Limitorque operators is based on individual qualification test reports for the wires rather than the Limitorque operator qualification program. The qualification of the Firewall SIS was based on a Rockbestos l
l 75
ORGANIZATION: LIMITORQUE CORPORATION LYNCHBURG, VIRGINIA REPORT INSPECTION N0.: 99900100/85-01 RESULTS:
PAGE 12 of 21 Qualification Test Report for Firewall III/ SIS dated February 1, 1977; however, the validity of all Rockbestos qualification tests conducted prior to June 1983 were questioned in IE Information Notice (IN) 84-44.
Based on this information notice, the qualification of Firewall SIS jumper wires is considered unsupported when qualification is based solely on the February 1, 1977 Rockbestos report; however, Rockbestos has recently provided Limitorque with results of its requalification tests conducted during the past year. The Rockbestos requalification tests are being conducted to resolve the qualification concerns identified in IN 84-44. The new test reports provided Lim torque support qualification of Firewall irradiation cross-linked i
polyethylene (XLPE) and the presently produced chemically XLPE (760-D formula) wires. Both of these XLPEs are covered by the Rockbestos specification RSS-4-002. Analysis of the cable test reports need to be performed and documented to assure that qualifi-cation of the wires is clearly established for specific applications.
The inspector's review of P0s established that Limitorque generally installed Raychem wire in valve operators from 1973 until approximately October 1978 at which time they started using Rockbestos wire.
Additional discussions with Limitorque personnel established that even though Limitorque may provide or reference documentation to support qualification of jumper wires they have installed in their nuclear supplied valve operators, valve manufacturers, licensees, or others may have added additional wires which are not qualified by the Limitorque data. The physical inspection at TVA's Sequoyah nuclear plant verified the above to be true.
2.
Valve / Actuator Weight Discrepancies Babcock & Wilcox (B&W) was notified in December 1984 of discrepancies in valve actuator waights listed in the valve vendor's drawings and seismic analyses. This problem, which involved valves manufactured by Anchor / Darling, Rockwell Inte~rnational, Copes-Vulcan, and ACF Industries, was identified by TVA at Bellefonte Units Nos.1 and 2.
The NRC inspector reviewed Limitorque records and iaentified P0 No.
048072LU from B&W to Limitorque to perform actual weight calculations for 18 actuators supplied by the above valve vendors. The calcu-lations identified the valve vendor, Limitorque's order number, total actuator unit weights, motor size, handwheel size and 76 l
ORGANIZATION: LIMITORQUE CORPORATION LYNCHBURG, VIRGINIA REPORT INSPECTION N0.: 99900100/85-01 RESULTS:
PAGE 13 of 21 mounting configuration and the actuator type. These calculations were completed March 18, 1985 and supplied to B&W on April 1, 1985. This area will be further reviewed at a future inspection of B&W to obtain the results of their evaluation / investigation.
l 3.
Reliance Electric Motors - Magnesium Alloy Rotors During qualification testing by General Electric, a Limitorque AC motor operator exhibited a substantial loss of torque. The failure was traced to the Reliance Electric motor and attributed to the use of a magnesium alloy rotor. This problem was identified by Niagara
{
Mohawk Power Corporation at Nine Mile Point Unit 2 and Gulf States I
Utility Company at River Bend.
At the request of Stone & Webster Engineering (S&W), Limitorque is currently pursuing a feasibility study on the development of motor designs utilizing alternate materials for the magnesium alloy rotor historically used in the standard, medium, and large frame nuclear motors (i.e., 180 frame and above). Limitorque stated that since the failure was at conditions exceeding the qualification parameters, no further action will be taken by Limitorque.
4.
10 CFR Part 21 Limitorque's process for evaluating and reporting Part 21 deficiencies was reviewed.
Procedure No. QCP-22, relating to the reporting of defects and failures describes responsibilities for the Design Review Committee (DRC) but does not ider.tify the chairman or members of the DRC. The procedure does not contain requirements for a committee quorum nor require written minutes to be kept of the committee's evaluations, course of corrective actions, and actions relative to NRC notification for 10 CFR Part 21 reporting.
(Note: Procedure No. QCP-15 specifies a composition
{
for a DRC for design control matters.) The procedure assigns the Technical Manager with the responsibility for reviewing all field services and customer problem reports. However, no formal procedure exists that requires personnel who handle these reports to submit them to the Technical Manager so that he may perform the required reviews.
A file of DRC minutes provided by the Technical Manager was reviewed. The general nature of the topics discussed in these minutes was product development or production oriented. No reference was found concerning any items discussed for 10 CFR 77
ORGANIZATION:
LIMITORQUE CORPORATION LYNCHBURG, VIRGINIA REPORT INSPECTION N0.: 99900100/85-01 RESULTS:
PAGF 14 of 21 Part 21 reporting. The minutes in the file covered meetings which were held on November 11, 1983; January 4, March 8, April 19 and 27, September 4 and 20, and October 18, 1984; and February 21, March 11, and September 13, 1985.
Some confusion apparently exists within Limitorque concerning what specifically constitutes a 10 CFR Part 21 report.
The QA Administrator verbally stated that Limitorque has never submitted a 10 CFR Part 21 report, although they have submitted 10 CFR Part 21 Maintenance Advisories (e.g., Limitorque letter to the NRC dated January 10,1983).
Yet, the Limitorque Interoffice memo dated February 10, 1983 that distributed NRC IN 83-02 dated January 28, 1983 referred to the Limitorque January 10, 1983 letter as a Part 21 notification.
Violation 85-01-02 was identified in this area of the inspection.
The content of Procedure No. QCP-22 was reviewed and the implementation of the procedure in regard to 10 CFR Part 21 posting requirements was assessed by inspecting the engineering and shop fabrication areas. The procedure does not identify the responsible director or corporate officer per Part 21 requirements and does not assure that such a director is informed if a basic component fails to comply with regulations or contains a defect.
Rather, procedures assign this responsibility to the DRC. Also, the procedure does not provide for informing the licensee or purchaser of a deviation or defect.
A copy of the regulation and a Notice were posted in the engineering areas, but there was no posting in the shop fabrication area.
In addition, the Notice did not contain all the information required by the regulations (see Violation 85-01-03).
5.
Actuator Spring Pack Design Changes NRC memorandLm AEOD/E509 dated July 25, 1985 discusses a depres-surization event that occurred at Salem Unit 2.
Although unrelated to the cause of the event, the subsequent investigation identified that one of the block valves involved had its Limitorque actuator replaced in April 1984.
The replacement actuator contained a spring pack with a slightly higher spring constant than the spring pack on the original actuator.
It was determined that Limitorque had incorporated small design changes between the original 78
ORGANIZATION: LIMITORQUE CORPORATION LYNCHBURG, VIRGINIA REPORT INSPECTION NO.:
99900100/85-01 RESULTS:
PAGE 15 of 21 procurement and the procurement of the replacement actuator. The end result was that the torque limit switch' setting on the replace-ment actuator had to be set at a different value than the one on the original actuator to obtain the same torque valve.
Limitorque's practice relating to changing part numbers when design changes are made is to assign new part numbers when a design change affects form, fit, or function. According to Limitorque, design change for a spring pack may or may not affect form, fit or function. A design change that modifies the spring constant but not the torque setting range is not considered by Limitorque to affect form, fit, or function, and a part number change is not made in this case.
Further investigation revealed that Limitorque does not routinely advise customers they are receiving an actuator which contains a spring pack with a different spring constant, although Limitorque will provide this information if requested by the customer. One exception to this practice occurs if a customer purchases a replacement spring pack.
In this case, a torque limit calibration tag is attached to the replacement spring pack which identifies the torque limit setting and corresponding torque.
The Limitorque practice that permits replacement spring pack changes to be made without either changing part numbers or providing notice of the change to a customer when replacement parts are ordered is of concern. This could lead to confusion by customer maintenance personnel who might set the torque limit switch on a replacement actuator at the same number as the original actuator, believing this will yield the same torque.
6.
Procurement Document Control The NRC inspector reviewed applicable sections of the QAM, and 22 P0s issued to 9 vendors to assure that quality requirements (e.g., ANSI N45.2 and 10 CFR Part 21) were passed on to suppliers / manufacturers of compor,ents for actuators destined for nuclear service. As noted below,13 of the 22 P0s did not impose ANSI N45.2 and 10 CFR Part 21 requirements on the vendor, and 3 of the P0s did not identify the 4
l latest revision of drawings and electrical quality control (EQC) documents (see Nonconformance 85-01-04).
It was also noted that quality personnel do not review P0s for inclusion of quality requirements prior to the issuance of P0s to vendors.
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ORGANIZATION: LIMITORQUE CORPORATION LYNCHBURG, VIRGINIA REPORT INSPECTION N0.: 99900100/85-01 RESULTS:
PAGE 16 of 21 P0 Date Vendor Item DIL-89921 7/12/85 Walker Machine Drive Sleeve DIL-87878 2/22/85 Walker Machine Drive Sleeve BSL-91090-MT 10/3/85 Southern Centrifugal Drive Sleeve &
Gear BSL-90260-MT 8/15/85 Southern Centrifugal Drive Sleeve &
Gear CIL-90997 9/24/85 Lynchburg Foundry Co.
Housing DIL-85391 2/5/85 Lynchburg Foundry Co.
Housing Cover LS-07?6 6/12/84 Foster Electric Rockbestos 14 AWG Wire LS-1046 9/4/84 Foster Electric Rockbestos 14 AWG Wire LS-1540 1/10/85 Foster Electric Rockbestos 14 AWG Wire LS-8034 6/17/82 Wesco Inc.
Rockbestos 14 AWG Wire SSL-91312 Ryerson Steel Bar Stock SSL-89633 Ryerson Steel Bar Stock 10/1/85 Copperweld Steel Bar Stock P0 (Item)
Vendor DWG/EQC (Rev) 3P3400 Item A Peerless-Winsmith EQC-1 (F)*
3P3673 Item S Peerless-Winsmith DWG #21-497-0014-1 (B)**
3P3673 Item G Peerless-Winsmith EQC-1 (F)*
3P3673 Item G Peerless-Winsmith DWG #21-497-0014-1 (B)**
- Latest revision is "G" dated June 11, 1983
- Latest revision is "C" dated July 10, 1985 7.
NDE The NRC inspector reviewed the qualification and certification records of NDE personnel (one-Level III and two-Level II) to I
determine whether the individuals performing PT examinations were certified to SNT-TC-1A. The written practice (Procedure No. QCP-11) of Limitorque for all phases of certifying NDE personnel was also reviewed.
The procedure indicated that training and examination was to be in accordance with SNT-TC-1A and also identified the standard methods for PT examination which included cleaning, application of dye, application of developer, and the acceptance / rejection criteria.
The procedure did not address the responsibility of each level of certification or duration on interrupted service requiring j
reexamination and re-certification.
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ORGANIZATION: LIMIT 0RQUE CORPORATION LYNCHBURG, VIRGINIA REPORT INSPECTION N0.: 99900100/85-01 RESULTS:
PAGE 17 of 21 Training and examination had been given to the three NDE personnel by a Level III examiner from Pittsburgh Testing Laboratories in June 1983. With the exception of a statement showing satisfactory completion of training in accordance with Procedure No. QCP-11, the qualification records of the three NDE personnel appeared to be in accordance with SNT-TC-1A. Nonconformance 85-01-06 was identified in this area of the inspection.
8.
Audits The NRC inspector reviewed applicable sections of the QAM, two Quality Control Procedures (QCP), four internal audit reports for 1983 and 1984, five vendor audits, 32 evaluations, and 10 appraisal reports to assure tha' item and services were being purchased from qualified noted that the area of QA Records was not covered in vendors.
,~
either the 1 C. or 1984 audits (see Nonconformance 85-01-07).
Vendor evaluations are not completed every six months as required by a QCP.
It was also noted that vendors who maintained a rating below Limitorque's quality standards were still listed in the AVL (quality standard is based on a 95% acceptance rate of reccipt inspection).
Two vendors who performed calibration services were listed in the AVL, but they have not been audited by Limitorque.
Nonconformance 85-01-08 was identified in this area of the inspection.
9.
Calibration of M&TE The NRC inspector reviewed applicable sections of the QAM, one procedure, records for M&TE, and certifications for reference standards used by service vendors to calibrate M&TE. An observation of M&TE at various work stations was also performed to assure that M&TE are properly identified, controlled and calibrated at specified intervals.
All of the M&TE were identifico with a S/N and contained a sticker from '.imitorque and/or the outside calibrator. The calibration dates an the sticker were cross-checked against gage record cards in tN calibration lab and were found to be satisfactory. The M&TE checPcJ included:
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LIMITORQUE CORPORATION LYNCHBURG, VIRGINIA REPORT INSPECTION N0.: 99900100/85-01 P.ESULTS:
PAGE 18 of 21 Station S/N Item Calibrator (date)
No. 1 QC 593 0-1" micrometer L (11/85)
QC 493 2-3" micrometer L (11/85)
QC 760 6" vernier L (10/85)
QC 267 surface plate Hartzog (10/84)
QC 743 hardness tester Page/ Wilson (9/85)
Lathe QC 651 3-4" micrometer L(10/85)
QC 703 inside micrometer L (4/85)
Automatic QC 863 height gage L (9/85)
QC 289 surface plate (3'x6')
Hartzog (10/84)
QC 761 depth vernier L(5/85)
Gear QC 610 surface plate (12"x18") Hartzog(10/84)
QC 634 dial depth gage L (11/85)
QC 27 4-5" micrometer L (11/85)
Receipt QC 949 vernier L (10/85)
Inspection QC 750 vernier L (7/85)
Special QC 861 hardness tester Page/ Wilson (9/85)
Process QC 792 micro-hardness tester Page/ Wilson (9/85)
Heat Treat QC 769 chart recorder Honeywell (10/85)
QC 770 temp recorder Honeywell (10/85)
QC 768 chart recorder Honeywell (10/85)
QC 766 chart recorder Honeywell (10/85)
QC 767 chart recorder Honeywell (10/85)
It was noted that the calibration of the three service plates by Hartzog was one month overdue, but the inspector was told that Limitorque was contacted by Hartzog, and the delay was due to a personal problem at Hartzog.
In the Special Process Department, the manager checked out both hardness testers using gage calibration blocks on the 30N scale on the Rockwell tester and KHN 746/750 on the microhardness tester.
It was noted that calibration samples No. 501-506 from Leco Corporation were used to calibrate the " Carbon Determinator - EC 12" instrument just prior to use.
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1 ORGANIZATION: LIMITORQUE CORPORATION LYNCHBURG, VIRGINIA REPORT INSPECTION l
NO.: 99900100/85-01 RESULTS:
PAGE 19 of 21 At Station No.1, the Level II examiner checked out the hardness tester using tw; gage calibration blocks (i.e., 3 x w/Rc and 1 x w/Rn.)
It was noted that the master setting disc (S/N QC 1038) from REB Ind. was not calibrated, but the QC Manager indicated that it was only used for reference purposes.
In the Gage Lab, a list of reference / transfer standards were reviewed, and calibration certifications from a number of service vendors were reviewed as noted:
S/N Item Calibrator QC 610 12" x 18" surface plate Hartzog QC 289 36" x 72" surface plate Hartzog QC 606 gage blocks (reference)
L QC 406 gage blocks (master)
Webster Gage Hardness tester (6)
Page Wilson Shinatzo NT-M001 Micro tester Page Wilson All certifications showed traceability to the National Bureau of-Standards. The Rockwell hardness testers were certified to ASTM E-18 and the microhardness tester to ASTM E384.
- 10. Motor Operator Failure - Feedwater Isolation Valve During testing in December 1984 at the River Bend job site of a feedwater system valve, the drive sleeve for the valve operator failed. The valve operator was returned to Limitorque for determination of the cause of failure, replacement of damaged parts, specified testing and inspection prior to return to the job site. Meetings held at the job site between responsible piping engineers and test engineers failed to identify the cause of the failure. A review of the condition of the operator by Limitorque was documented in an Inter-Office Correspondence dated September 25, 1984, but it did not show the cause of the failure.
The inspector reviewed:
S&W P0 No. 28530 dated April 13, 1984 issued to Limitorque for repair and retest of one actuator (S/N 321320); letter dated May 24, 1984 from Limitorque to S&W identifying parts replaced (one torque switch, two geared limit switch finger assemblies and one drive sleeve) and test performed; and a letter dated September 27, 1984 from Limitorque to Velan l
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ORGANIZATION: LIMITORQUE CORPORATION LYNCHBURG, VIRGINIA REPORT INSPECTION NO.: 99900100/85-01 RESULTS:
PAGE 20 of 21 Engineering (valve manufacturer) stating that the cause of the failure was due to several stall thrust loads.
The inspector also reviewed S&W's Certificate of Compliance dated July 3, 1984 which indicated the actuator was tested and inspected by S&W representative prior to shipping.
- 11. Documentation Packages The NRC inspector reviewed six documentation packages and QC-Nuclear files for safety-related actuators ordered in November 1983 to July 1985. The six P0s from Public Service Electric &
Gas, Niagara Mohawk Power Corporation, Velan Valve, Westinghouse, Northeast Utilities, and Iowa Electric Light & Power Company referenced specific technical specifications,10 CFR Part 21, and IEEE Standards. The documentation packages consisted of calibration sheets, Bill of Materials, P0s to rnotor manufacturers, Certificate of Compliances and test reports and performance curves from motor manufacturers.
The P0s to two motor manufacturers, Reliance Electric and H. K.
Porter, referenced standard EQC-1 and drawing No.21-497 which documented technical and quality requirements. The QC-Nuclear file for each order consisted of a station No. 1 inspection repcrt, final inspection and test report, liquid penetrant examination report and a coating certification.
It was noted that quality requirements (e.g., QA Program and 10 CFR Part 21) were only imposed upon motor tranufacturers. The imposition of quality requirements was imposed upon cable manufacturers starting in March 1985. Traceability is maintained only on the motor and the completed actuator to specific tests, but there is no traceability to individual components.
- 12. QA Program Deficiencies In reviewing the previous inspection findings, it was noted that the deficiency in nonconformance (Item A, 84-01) was originally identificd by the NRC during an April 1982 inspection.
In a letter dated November 15, 1982 to the NRC, limitorque stated, "Limitorque Corporation to revise QA manual to include training of non-QC personnel." Approximately three years have gone by, and Limitorque has still failed to incorporate this activity into their QA manual (see Nonconformance 85-01-05).
84 i
ORGANIZATION: LIMITORQUE CORPORATION LYNCHBURG, VIRGINIA REPORT INSPECTION N0.: 99900100/85-01 RESULTS:
PAGE 21 of 21 i
TVA had reported a number of QA program deficiencies in a combined deficiency report (10 CFR Part 50.55(e)) and a defect report (10 CFR Part 21) dated December 15, 1983, for valve operators to be supplied for Watts Bar and Bellefonte Units 1 and 2.
Three deficiencies were closed out by TVA in a March 6,1984 letter (QDBVA-84-5) which was reviewed by the NRC inspector.
- 13. Control of Purchased Material P0s for bar stock to Ryerson Steel called for ASTM A304 heat treatable 8620 steel, but the CMTR specified that ASTM-A-322 heat treatable 8620 steel had been supplied. The QC Manager's explanation of this was that the two specifications were the same if a dominy test was performed on the material. This test provides hardness values which can be used to determine the heat treatability of the steel.
In the CMTR mentioned above and others reviewed, the QC Manager stated that the dominy values were compared to the required ASTM A304 standard and found to be adequate. When asked as to whether an internal deficiency reported was filed by the receiving inspector, the QC Manager said no because the acceptability of ASTM-A-322-82 standard 8620 steel was verified verbally with the engineering department.
i i
85
ORGANIZATION: MINNES0TA MINING AND MANUFACTURING COMPANY SAINT PAUL, MINNES0TA REPORT INSPECTION INSPECTION N0.: 99901038/85-01 DATE(S):
11/20-22/85 ON-SITE HOURS: 44
)
CORRESPONDENCE ADDRESS: 3M Company Ceramic Materials Department ATTN: Mr. Joseph T. Bailey Department Manager 225-4N-07 3M Center Saint Paul, Minnesota 55144-1000 ORGANIZATIONAL CONTACT: Mr. W. P. Nitardy TELEPHONE NUMBER:
(612) 733-6262 NUCLEAR INDUSTRY ACTIVITY: 3M currently offers the nuclear industry a one hour UL rated fire barrier wrapping system for protecting electrical cables.
p il I
((
ASSIGNED INSPECTOR:
,u je
'J.' J.~fefrosino,' Reactive Inspection Section (RIS
~
Date f
0THER INSPECTOR:
E. Yachimiak, RIS g
APPROVED BY:
Date E. W. Merschof#g Chief, RIS, Vendor Program Branch INSPECTION BASES AND SCOPE:
A.
BASES:
SCOPE: The purpose of this inspection was to review the circumstances surrounding an NRC concern which was identified during a recent I
Ft. Calhoun outage inspection.
PLANT SITE APPLICABILITY:
Enrico Fermi 2 (50-341); Ft. Calhoun 1(50-285);
Haddam Neck (50-213); Hope Creek 1(50-354); Nine Mile Point 1 (50-220);
(continued on next page) 87
ORGANIZATION: MINNES0TA MINING AND MANUFACTURING COMPANY SAINT PAUL, MINNES0TA REPORT INSPECTION NO.-
99901038/85-01 RESULTS:
PAGE 2 of 4 PLANT SITE APPLICABILITY:
(continued) Palisades (50-255), Perry 1 (50-440);
Quad Cities 1 & 2 (50-254/265); Rancho Seco 1 (50-312); Salem 1 & 2 (50-272/311); and Watts Bar 1 & 2 (50-390/391).
A.
Violations None.
B.
Nonconformances 1.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Section 1 of the 3M quality assurance manual (85-01-01):
a.
The 3M Company did n'ot perform or control ampacity derating computer code activities, in accordance with documented instructions.
These activities included engineering, research, data processing and document control, b.
Documented procedures or instructions to control activities affecting quality were not initiated by the 3M Company for its cable ampacity derating computer code development, in-process program activities, or modifications thereof.
2.
Contrary to Criterion II of Appendix B to 10 CFR Part 50 and Section 2 of the 3M QA manual, three employees out of a sample of ten had not received the required QA program indoctrination (85-01-02).
3.
Contrary to Criterion I, of Appendix B to 10 CFR Part 50 and Section I of the 3M QA manual, the QA Manager does not report to a management level which would enable him to achieve the required authority and organizational freedom, including sufficient independence from cost and schedule.
Examples of this as of November 22, 1985, are (85-01-03):
a.
The QA Manager reported to the manager of production instead of the 3M Ceramic Department Manager, as indicated on the organizational chart within the QAM.
b.
The annual 3M personnel performance evaluation for the QA Manager is initiated and completed by the 3M Production Manager.
C.
Unresolved Items None.
88
r ORGANIZATION: MINNES0TA MINING AND MANUFACTURING COMPANY SAINT PAUL, MINNES0TA REPORT INSPECTION NO.-
99901038/85-01 RESULTS:
PAGE 3 of 4 D.
Background Information During a recent NRC inspection at the Omaha Public Power District (0 PPD)
Fort Calhoun Station, concerns were identified relative to cable ampacity derating values that the 3M Company provided to OPPD. The derating values were requested by 0 PPD for use with the 3M "Interam" fire wrap system. The specific NRC concerns are:
OPPD did not verify the validity of the ampacity derating values supplied to them by the 3M Company.
Failure of OPPD to assure the validity of the derating values is identified as a deficiency, item #D.5.2-1, in NRC report #50-285/85-22.
A misinterpretation by'0 PPD or other NRC licensees could occur, because of the manner in which the derating values were stated and supplemented with percentages for quick reference.
Additionally, a lack of standards for the acceptance or rejection criteria of " fire wrapped" cables and raceways, was revealed.
The lack of this criteria has prompted fire wrapping manufacturers to calculate and provide their own derating values.
E.
Other Findings or Comments Ampacity Derating Methods The 3M Company is taking additional measures to assure the accuracy of its derating values.
The methods which 3M are utilizing to obtain the cable ampacity derating values are discussed below:
a.
Two methods of determining ampacity derating factors are
]
currently utilized by 3M.
One involves a mathematical computer model, and the other involves 3M's experimental test data
- results, b.
The 3M Company documented and correlated separate computer code run results with specific National Electrical Code (NEC) copper conductor ampacity tables.
In addition, they expressed their derating value in a percentage. This derating percentage was based on the computer generated ampacity data, which when compared to the NEC ampacity l
value, resulted in a less conservative derating factor.
i Currently, there are no industry wide standards that manufactures or licensees could utilize for guidance when using the "Interam" fire wrap system or other similar products.
89
ORGANIZATION: MINNESOTA MINING AND MANUFACTURING COMPANY SAINT PAUL, MINNES0TA REPORT INSPECTION NO.-
99901038/85-01 RESULTS:
PAGE 4 of 4 Part 21 The 3M Company is currently evaluating the computer code and derating percentage values as potentially reportable 10 CFR Part 21 items. All affected licensees are identified on page 1 of this inspection report.
An evaluation of a previous problem was reviewed by the NRC inspectors and subsequently discussed with 3M QA personnel. This item concerned fire barrier ceramic cord which had been utilized to secure the "Interam" system once it was wrapped around the cable or raceway.
Degradation of the cord could occur at certain humidities. The 3M Company is reevaluating this situation following discussions with the NRC inspector.
10 CFR 50 Appendix R Implications on Cable Ampacity Derating During the recent Outage Inspection at Fort Calhoun Nuclear Station by the NRC a deficiency was identified concerning Appendix R compliance modifications.
These modifications consisted of the wrapping of electrical conduit and cable trays with 3M's Interam fire protection wrapping system.
The " engineering" analysis performed by 0 PPD staff members to determine if cable ampacity derating.was required consisted of comparing the FSAR cable rating design value (125% of design load) with a 3M supplied ampacity derating factor.
No documented records could be produced to assure that the information supplied by 3M was acceptable and within the original design parameters for their proposed modifications. Since there are no standard tables available to determine cable ampacity derating factors for insulated conduit or cable trays, an analysis should have been performed to verify the validity of 3M's derating factors. Therefore, deficiency item #D.5.2-1, on NRC inspection report #50-285/85-22 for the Fort Calhoun station was identified.
E.
Persons Contacted William Nitardy QA Manager **
Randall Koza Sr. Product Development Engineer **
Richard Licht Production Development Supervisor **
l Joseph T. Bailey Ceramic Department Manager **
I Thomas Sheehan Marketing Manager **
Robert Topness Production Manager **
Diane Trumble Document Control Coordinator
- Exit meeting attendees.
i 90 l
ORGANIZAT10N:
PAClFIC VALVES LONG BEACH, CALIFORNIA l
REPORT INSPECTION INSPECTION NO.: 99900075/86-01 DATE: 2/10-14/86 ON-SITE HOURS: 28 i
CORRESPONDENCE ADDRESS:
Pacific Valves ATTN: Mr. Reid Armstrong Vice President, Engineering l
i 3201 Walnut Street Long Beach, California 90807 ORGANIZATIONAL CONTACT: Mr. Robert Argent, QA Manager TELEPHONE NLHBER:
(213) 426-2531 NUCLEAR INDUSTRY ACTIVITY:
Approximately 2 percent of valve sales.
ASSIGNED INSPECTOR:
8\\.hv74 blT-E4 U.)T. Conway, React ve Inspection Section (RIS)
Date 1
1 V
OTHER INSPECTOR (S):
O' 2
i APPROVED BY: b
& M4///b
^
E. W. Merschoff, Chief,' RIS, Vendor Program Branch te INSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR Part 21 and 10 CFR Part 50, Appendix B.
i I
B.
SCOPE: The purpose of this inspection was to review manufacturing records and gather data on design changes, recommended maintenance programs and vendor / licensee interface pertaining to swing-type check l
valves.
PLANT SITE APPLICABILITY:
Failed swing-type check valves - San Onofre Unit 1 (50-206); Failed stop check valves - Turkey Point Units 3 and 4 (50-250/251).
I 91 i
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ORGANIZATION:
PACIFIC VALVES LONG BEACH, CALIFORNIA a
REPORT INSPECTION NO.-
99900075/86-01 RESULTS:
PAGE 2 of,11 A.
VIOLATIONS:
1.
Contrary to Section 21.6 of 10 CFR Part 21, copies of Section 206 of the Energy Reorganization Act and Procedure No. QSP-0004
" Identifying and Reporting Defects Under 10 CFR 21" were not posted, and an outdated copy of 10 CFR Part 21 was posted.
(86-01-01) 2.
Contrary to Section 21.31 of 10 CFR Part 21, a review of documen-tation packages for Section III swing-type check valves revealed that while 10 CFR Part 21 was imposed on Pacific Valves (PV) by their customers, PV did not specify that 10 CFR Part 21 require-ments would apply on purchase orders 46330N and 46360N to Pacific Southern Foundries (PSF), 50396 to Sun-Ray Testing (SRT), 49956N to A&G Engineering, 56698N to Poly Cast, and 58159 to Jorgenson Steel.
(86-01-02)
B.
NONCONFORMANCES:
1.
Contrary to Criterion V of Appendix B to 10 CFR Pa'rt 50 and Sections 5.7.1.1 and 5.7.1.2 of the Quality Assurance Manual (QAM), there was no documented evidence that NDE procedure QCS 300, Revisions K and N; FSP-0900, Revision 0; FSP-0901, Revision 1; and heat treat procedure FSP-0950, Revisions 0 and 2 from PSF were reviewed and approved by PV.
(86-01-03) 2.
Contrary to Criterion V of Appendix B to 10 CFR Part 50, Sections 7.2 and 9.2 of STD No. QAS-6, and Sections 7.4.1 and 7.4.2.1 of the QAM, a review of qualification records for nine NDE personnel, three each from PV, PSF, and SRT revealed the following:
(86-01-04) a.
There was no documented evidence that M. Hess from SRT was qualified to a Level III examiner when he qualified PV's J. Sewell to a Level III-liquid penetrant testing (PT) in February 1984.
b.
Qualification records were missing for R. Nielson from F3F who performed magnetic particle testing (MT) in May 1982.
c.
The qualification records of the three PV examiners did not contain a statement indicating satisfactory completion of training in accordance with STD No. QAS-6.
92
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ORGANIZATION:
PACIFIC. VALVES LONG BEACH, CALIFORNIA L_
REPORT INSPECTION N0.-
99900075/86-01 RESULTS:
PAGE 3 of 11 3.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Sections
~
5.3.8 and 11.5.2 of the QAM, it was noted that vendor audits were conducted by M. Merrill in September 1983 and K. Cranek in April 1985, but they were not qualified as auditors until July 1984 and October 1985, respectively.
(86-01-05)
C.
OPEN ITEMS:
s None.
D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
Not addressed during this inspection.
E.
OTHER FINDINGS OR COMMENTS:
1.
San Onofre Unit No. 1 (SONGS-l)' Incident The NRC inspector discussed the five swing-type check valves that failed in the feed water system at SONGS-1 with the V-P Engineering, the Chief Design Engineer and the QA Manager. Upon learning of the failed valves, PV sent a Customer Service Representative to the reactor site to get the Serial Number (S/N) of the affected valves.
The tag containing the S/N for the 4" valve was missing, but the S/Ns for the three 10" valves were 47754, 47755, and 47756; and the 12" valve was identified with S/N 48345.
The inspector was told that manufacturing records did not exist for commercial valves manufactured p prior to the first Section III valve fabricated in 1975. The only record was PV's Valve Serial Number Log dated 1957,to 1966. The log contained model numbers along with S/Ns and the date that a' valve underwent hydro testing. A review of this log indicated.that thes,
'j three 10" valves were hydrotested on December 7, 1965, and the 1?"
c, valve was tested on December 30, 1965-However, there is no s
indication where the valves, which were manufadtured as commercial grade WCB, were shipped. The NRC Incident l'nvestigation Team (IIT) that evaluated the ~ event at SONGS-1 was told that the valves were ~
purchased by Bechtel from Atlantic Richfield Hanford Company.
Neither Southern California Edison nor Bechtel conucted PV 'regdr'd-ing check valve deteriorations (i.e., discs, hinge supports,.and hinge pins were replaced in'the failed valves at SONGS-1 in 1975 and 1977) until January 1986 when Bechtel called PV requesting spare parts for
.eight check valves and a proposal for design enhancements for the swing-type check valves.
Initially, Bechtel was satisfed with s
gs 93 s
~
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ORGANIZATION: PACIFIC VALVES LONG BEACH, CALIFORNIA REPORT INSPECTION NO - 99900075/86-01 RESULTS:
PAGE 4 of 11 PV's proposed changes which included adding stellite to the wear surfaces of the disc and hinge support, enlarging the diameter of the hinge pin, and changing the configuration of attaching the disc to the hinge support.
However, Bechtel later told PV that they were going to replace the failed valves with Atwood Morrill swing-type check valves which have a one piece disc / hinge assembly.
NUREG-1190 documented the I!T's findings at SONGS-1, and it was noted that the three 10" main feedwater regulator check valves were inspected in 1975 and 1977, and new internals were installed in all three check valves. The NRC inspector was told by IV management that records did not exist to show that spare parts for PV swing-type check valves were ever ordered or shipped to SONGS-1.
2.
Swing-Type Check Valves In 1975, PV received their "N" and "NPT" stamps, and to date approximately 226 swing-type check valves have been manufactured to the requirementsSection III, Clacs 1, 2 & 3 of the ASME Code.
Since the initial nuclear check valve was shipped in January 1976, PV has not made any changes in material for individual components or design configuration. Tack welds are not used, and the attachment of the disc to the hinge support is with a nut that is pinned to the threaded stem of the disc.
Since 1970, a copy of PV's Maintenance Manual has been sent to customers with the finished valves.
Section C " Pacific Swing Check Valves" of the bnual addresses Installation, Operating Instructions, Maintenance, and Preventive Maintenance. There is a caution not to exceed the pressure and temperature limits of ANSI B16.34, but specific examination periods for maintenance activities are not identified in the Manual.
Starting in 1978, PV cautioned owners in Catalog No. 400 regarding the use of swing check valves in certain environments. Under
" Service Recommendations" it was noted that service in systems involving rapid and frequent flow reversals, pulsation or excessively turbulent flow should be avoided; and the location o.f swing-type check valves away from fittings within the piping system could minimize or eliminate potential problems.
It was also recom-l mended that suspect problem systems be reviewed with PV before selecting and purchasing a particular swing-type check valve.
For all the nuclear orders of Section III swing-type check valves, PV could not recall any discussions or correspondence with customers 94 l
I
ORGANIZATION: PACIFIC VALVES LONG BEACH, CALIFORNIA REPORT INSPECTION NO - 99900075/86-01 RESULTS:
PAGE 5 of 11 pertaining to operational environments within 6 piping system. The purchase orders (P0) and technical specifications did not identify the specific flow conditions, and the P0 only listed a valve size, model number and class.
For best performance, the catalog also recommended that swing-type check valves operate within a flow range sufficient to hold the valve fully open, but not so high that it produces excessive turbulence. To correctly size a valve, the seat port velocity at the operating flow rate should be calculated, and it should fall within the velocity range noted by the following formula:
V (fps) 240 h
[$
55 Where [ equals the square root of specific volume of flow medium at operating conditions.
The NRC inspector was told that PV service representatives normally distribute the catalog to their customers.
Feedback from customers pertaining to problems with PV valves is directed to the Field Services organization who maintains a file of customer inquiries or complaints and subsequent action taken by PV.
PV could not identify in their files any incidents where nuclear or commercial customers had notified PV of swing-type check valve failures and in particular any that were similar to those failures at SONGS-1.
It was also noted that PV does not perform any trend analysis on failure data received from their customers.
3.
Welding The NRC inspector reviewed the qualification records of seven welders and seven welding procedures used for the manufacture of nine nuclear check valves and related spare parts.
A review of Welder Performance Continuity Records and Qualification Test Records of Welder Nos.10, 12,13, 21, 30, 32, and 34 indicated that all the welders were qualified to weld with the applicable Welding Procedure Specification (WPS). A review of seven Procedure Qualification Records indicated that the WPSs used to weld the seal rings into the body, to apply the hard-facing on the sealing surfaces and for repair welding were all qualified, and test specimens had been satisfactorily tested by either Continental Testing or Dickson Testing.
l l
95
ORGANIZATION: PACIFIC VALVES LONG BEACH, CALIFORNIA 1
REPORT INSPECTION NO.-
99900075/86-01 RESULTS:
PAGE 6 of 11 The inspector evaluated the weld material control area. Electrodes for nuclear work were retained in two furnaces. One furnace (S/N 23916) contained la different type rods and was designated for the "1st shift." A calibration sticker was on the front door and indicated that Golden State Calibration & Service Company had calibrated the temperature recorder (current reading of 240 F) on December 10, 1985.
The second furnace (S/N 23978) was for "2nd shift' and contained a similar calibration sticker. The second furnace contained six different size rods. Both furnaces were padlocked, and the keys are controlled by the Weld Shop Supervisor.
4.
10 CFR Part 21 Procedure QSP-0004 relating to the identification and reporting of defects and failures was reviewed, and the implementation of the procedure in regard to posting requirements was evaluated by inspecting the shop areas.
It was noted that an outdated copy of 10 CFR Part 21 was the only document posted in several areas (see Violation 86-01-01). A review of PV P0s to vendors for ccmponents used in the manufacture of safety-related nuclear valves revealed that the requirements of 10 CFR Part 21 were not referenced or identified on the following P0s which were stamped " Nuclear:"
(See Violation 86-01-02)
- 46330N (December 1, 1981) and 46360N (January 5, 1982) to PSF for castings (body, bonnet, and disc)
- 50396 (October 6, 1982) to SRT for radiographic testing (RT)
- 49956N (August 10,1982) to A&G Engineering for fasteners
- 56698N (September 10,1984) to Poly Cast for hard-coating plasma powder
- 58159 (March 5,1985) to Jorg'ensen Steel for bar and plate 5.
NDE The NRC inspector reviewed the qualification and certification records of nine NDE personnel, three each from PV,' PSF, and SRT, to determine whether the individuals performing NDE were certified to SNT-TC-1A.
The written practice (STD No. QAS-6) of PV for all phases of training i
and certifying NDE personnel wa's also reviewed, and it appeared to be consistent with SNT-TC-1A.
96 1
ORGANIZATION: PACIFIC VALVES 1
LONG EEACH, CALIFORNIA
)
i - - - _ _,
REPORT INE?ECTION I
NO - _99900015485-0,J _
RESULTS:
lPAGE 7 of 11 One Level III examiner from SRT performd RT on nuc' lear valve components in Octobee and December 1982. The other two examiners administered and graded the examinations of PV's J. Sewell to a 1.evel IZ-RT and F.T and a level III-PT. The qualtficatien records a
on file at FV for the examiner who performed RT ond the Level Ill-RT and MT indicated that they bcth met the requiren.ents of SNT-TC-1A.
It was noted that J. Sewell was certif1ed in February 1984 to a l
Level Ill-PT, but the records on fife at,PV indicated that Hess was l
not certified to a Level IIT-PT until August 1984 -(See Nonconformance i
86-01-04a).
Tne three examiners from PSF perTermed RT and MT on valve coripenents (body, bonnet ar.d disc) used on Section 711 ning-type check valves.
With the exception of r& cords fcr R. Neilsen sne cerforme:.' MY in May 1982 of a bonnat on shop job (S/J) No.10000ii6H, PV had sufficient records to verify that the other exanjiners :neet the qualification requirements of MT-TC-1A when they performed a specific OE (See Nonconformance 86-01-04b).
The NDE examiners frcm PV included two Level IIIs and one Level II certified to PT. The current Level III at PV was also certiffed to
~
a level II in MT and RT in Fet'ruary 198a. 14f th r.he exception of L missing statement pertaining to training in accordance with STD No.
QAS-6,, the records including eye examinatint.s, training, w"itten tests and certifications appeared to satisfy the requirements of SNT-TC-1A (see Nonconformance S6-01-0a ).
c Five NDE procedures were reviewed. Two procedures from FV addressed RT (QSP-0045) and FT (QSP.0G47) of valve components. The F.T (QOS-300), PT (FSP.-0900), and MT (FSP-0901) of steel castings was ccvered in PSF procedures, it was cated that the PSF procedures wera oot i
reviewed and approved by PV's QA departunt (see Nonconformance
]
86-01-03).
During a walk-throu0h of the RT station, it was observed that the
]
densitometer is calibrated cgairist a censity strip (tlo.1498) on an i
annual basis, but the reference standard (strip) had not been calibrated since 1981 as documented on a Denslty Strip Calibrattor.
Report from Dupont.
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97 1
ORGANIZATION:
PACIFIC VALVES LONG BEACH, CALIFORNIA REFORT INSPECTION NO.
99900075/86-01 RESULTS:
PAGE 8 of 11 6.
Audits The NRC inspector reviewed applicable sections of the QAM, two procedures, seven external audit reports and the qualification records of four auditors to assure that items and services were being purchased from qualified vendors, and PV personnel were trained and qualified to perform audits.
PV's activities pertaining to five manufacturers who supplied items for safety related swing-type check valves were evaluated.
Audits were conducted by PV of Poly Cast (one) who supplied hard-coating material, PSF (three) who supplied castings, and SRT (three) who performed NDE.
Checklists were used on all the audits which were conducted in 1983, 1984, and 1985, and the results were documented in a Vendor Quality System Evaluation report. The other two manufacturers - E.M. Jorgenson (wrought products) and A&G Engineering (fasteners) - were Quality System Certificate (Materials) holders and were not audited by PV.
The QA records for the four personnel who conducted the external audits consisted of Auditor Qualification Records, a written examination addressing material contained in ANSI /ASME NQA-1 and related codes and regulations, and a Record of Lead Auditor Qualification.
It was noted that SRT was audited by M. Merrill in September 1983 and K. Cranek in September 1985, but QA records indicated that they were not qualified to lead auditors until July 1984 and October 1985, respectively (see Nonconformance 86-01-05).
7.
Stop Check Valve Failures - Turkey Point 1
Between November 1985 and January 1986, numerous failures were experienced in stop check valves located in the steam supply system to the auxiliary feedwater pumps at Turkey Point Unit Nos. 3 and 4.
The failure mode was degradation of the disc and disc nut assembly due to low stream flow conditions caused by a leaking motor-operated valve (M0V) in the system.
In seven valves, the vibration and chattering of the disc assembly resulted in a broken disc guide stud which caused the disc to become locked in the valve thus preventing full closure and full opening of the valve. The broken stud was also free to travel causing further damage.
98
ORGANIZATION:
PACIFIC VALVES LONG BEACH, CALIFORNIA REPORT INSPECTION NO.-
99900075/86-01 RESULTS:
PAGE 9 of 11 The failed valves were ordered by Florida Power & Light (FP&L) on P0 No. 65380-27088P dated May 24, 1983 to be fabricated in accord-ance with specification PTP-1000-4.55.5 and Section III/ Class 2 requirements of the Code. Twelve carbon steel valves (3" 660S WE-80-X) were manufactured by PV on S/J No. 3A0038N, and the valves were sent to FP&L in October 1983.
With regards to the failures in November 1985, PV's Manager, Field Services (MFS) supervised the replacement of damaged parts in five valves in Unit No. 3.
The MFS told the NRC inspector that the spare parts (e.g., disc assembly, seat ring) were identical (material and configuration) to the original components in the valve. When the reactor went operational, the MFS observed the re-worked valves and noted that the internals were still chattering, l
and the valves were very noisy.
It is PV's opinion that the valves are failing because of the turbulence problem caused by the leaking M0V.
8.
Nuclear Orders - Check Valves The NRC reviewed the QA records for three nuclear orders and five spare part orders of swing-type check valves.
Records consisted of P0s and technical specifications from the customer and PV Data i
Packages (DP). Eight P0s, five of which were for spare parts, referenced technical specifications and the requirements of Section III of the Code and 10 CFR Part 21. A DP consisted of Data Reports for the pressure retaining parts (i.e., body, bonnet, and disc), Production Work Order (i.e., traveler), hydrostatic test report, certified material test reports (CMTR), heat treat reports with time / temperature charts, casting repair reports, welding and hard facing records, NDE reports, wall thickness measurements, and visual examination checklists. The CMTRs were for pressure retaining castings, welding material, and fasteners; and the NDE reports were for MT, PT and RT.
The three most recent nuclear orders were from:
(a) Stone &
Webster (S&W) on P0 No. 2362.050-651 dated May 7, 1981 for Millstone Unit No. 3; (b) S&W on P0 No. 2282.050-676 dated September 17, 1981 for Millstone Unit No. 3; and Gibbs & Hill (G&H) P0 No. 5021-1-6.6-044 dated October 23, 1984 for Susquehanna.
The May 1981 order from S&W was for carbon steel valves, and the other order was for stainless steel valves. The G&H order was for valves to be installed at the Diesel Generator "E" facility.
I 99 l
ORGANIZATION: FACIFIC VALVES LONG BEACH, CALIF 0P.NIA REPORT INSPECTI0li Nn
- 99900075/86-01 RESUI.TS:
PAGE 10 of 11 Data obtained on the three nuclear orders is as follows:
S/J No.
Quality-Model No.
Approx. Date Shipped IW0066N 2-10" 180 '7-WE-X 10/82 l
IW0163N 10-2s" A-160-14-FF-X 11/82 I
2-3" A-180-14-FF-X 11/82 2-8" G-680-7-WE-X 12/82 4WO340H 2-10* 380-15-WE-40-X 10/85 Spare part orders for nuclear valves included itenis.such as seat ring, wedge, stem, eyebolt, gasket (bonnet), hinge pin, disc, nut washer (disc), and segment ring. The five orders were from the following customers and were manufactured under the noted S/J No.:
1 Customer P0 No.. (date)
S/J No.
S&W 2362.050-651-30 4WO284ti (9/19/85)
S&W 2282.050-676 4W00321t (7/17/84)
Omaha Public 536874 4B02691 Power District (9/17/84)
Ouke Power M25624-73 SWO143N (4/25/85)
Kansas Gas &
53239 6H0369Q Electric (11/7/85)
The S/J Instructions for both valve and spare part ordert originated in the Project Engir.eering Department and included requirements for pressure retaining and non-pressure retaining components, and weld filler metal; list of procedures; and documentation required with the shipment.
100 l
i OAGANIZATION: PACIFIC VALVES LONG BEACH, CALIFORNIA 1
REPORT INSPECTION HD.-
99900075/86-01 RESUI.TS:
PAGE 11 of 11 F.
PERSONNEL CONTACTED
- R. Argent, QA Manager
- R. Armstrong., V-P, Engineerirg J. Sewell, NDE - Level III F. Hernandez, Gage Technician G. Brown, Chief Design Engineer d
F. O'Brien, Manager - Design Engineering R. McConnell, Managar - Field Services G. Hay, held Shop Supervisor
- attended exit meeting i
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ORGANIZATION: THE ROCKBESTOS COMPANY NEW HAVEN, CONNECTICUT REPORT INSPECTION INSPECTION N0.: 99900277/86-01 DATE:
1/21-22/86 ON-SITE HOURS:
16 CORRESPONDENCE ADDRESS: The Rockbestos Company A Member of the Marmon Group ATTN: Mr. R. S. Thayer General Manager Post, Office Drawer 1102 New Haven, Connecticut 06504 ORGANIZATIONAL CONTACT: Mr. George Littlehales, QA Manager TELEPHONE NUMBER:
(203) 772-2250 NUCLEAR INDUSTRY ACTIVITY: Currently the testing laboratory at the New Haven plant conducts all thermal aging, loss-of-coolant accident (LOCA) simulations, post-LOCA sample evaluation testing, and flame testing for Class IE (safety-related) electrical equipment qualification (EQ) on Rockbestos cable. Nuclear related product manufacturing at the New Haven plant comprises approximately 5 percent of the plants total output.
[bj
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v
(
ASSIGNED INSPECTOR:
S. D. Alexander, Equipment Qualification Inspection ae Section (EQIS)
OTHER INSPECTOR (S):
M. Jacobus, Sandia National Laboratories (SNL)
APPROVED BY:
i)! A b M%zrP 2-2 4[$(,
U. Totapovs, Chief, EQIS, Vendor Program Branch Date INSPECTION BASES AND SCOPE:
A.
BASES:
B.
SCOPE:
This inspection consisted of review of the status of the requal-ification program (RP) on currently manufactured Class IE cable types.
Emphasis was on observing post-LOCA sample evaluation testing of RSS-6-100 series, adverse nuclear service coaxial, twinaxial and triaxial cable samples.
PLANT SITE APPLICABILITY:
San Onofre 1 (50-206); Haddam Neck (50-213); Nine Mile Point 1 (50-220); Dresden 2 (50-237); Millstone 1 (50-245); Dresden 3 r
(50-249); Turkey Point 3 and 4 (50-250/251); Palisades (50-255) (continued on next page) 103
ORGANIZATION: THE ROCKBESTOS COMPANY NEW HAVEN, CONNECTICUT REPORT INSPECTION NO.-
99900?77/86-01 RESULTS:
PAGE 2 of 6 PLANT SITE APPLICABILITY:
(continued)
Monticello (50-263); Quad Cities 2 (50-265); Point Beach 1 (50-266); Peach Botton 2 and 3 (50-277/278); Prarie Island 1 (50-282); Indian Point 3 (50-286); Pilgrim 1 (50-293); Zion 1 (50-295); Point Beach 2 (50-301);
Zion 2 (50-304); Kewaunee (50-305; Prarie Island 2 (50-306); Main Yankee (50-309); Arkansas 1 (50-313; Calvert Cliffs 1 and 2 (50-317/318);
Fitzpatrick (50-333); St. Lucie 1 (50-335); Millstone 2 (50-336); San Onofre 2 and 3 (50-361/362); Arkansas 2 (50-368); McGuire 1 (50-369); LaSalle 1 and 2 (50-373/374); Med. Coll. Hanover (50-377); St. Lucie 2 (50-389); Lacrosse (50-49); Nine Mile Point 2 (50-410); Catawba 1 and 2 (50-413/414); and WPSS-3 (50-508).
A.
VIOLATIONS:
None.
B.
NONCONFORMANCES:
None.
C.
UNRESOLVED ITEMS:
None.
D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
1.
(0 pen) Nonconformances (84-02, Items B.1 and R.2; 83-03, Items B.1 - B.5):
The NRC inspector took no action on these production nonconformances during this inspection.
2.
(0 pen) Nonconformances (83-04, Item B.1 and B.2; 83-02, Item B.1 example (2); Unresolved Items 83-04, Item C; 83-03, Item C; 83-02, Item C.2; and Followup Item 81-01:
These items were recapitulated in inspection report 99900277/85-01.
They are under continuing review.
They will remain open until evaluation of the requalification program is completed and validity of prior tests affected by the items can be determined.
104
ORGANIZATION: THE ROCKBESTOS COMPANY NEW HAVEN, CONNECTICUT REPORT INSPECTION NO.-
99900277/86-01 RESULTS:
PAGE 3 of 6 E.
OTHER FINDINGS OR COMMENTS:
1.
During this inspection, the NRC inspectors reviewed corrective actions from previously identified nonconformances, reviewed data from test plan (TP) No. 5802 on RSS-6-100 series samples which had been generated since the last inspection, observed post-LOCA sample evaluation testing on these samples, and reviewed related documentation. While the NRC is following this program closely to verify compliance with regulatory requirements, the responsibility remains with individual licensees who intend to utilize the results of these tests in support of qualification to evaluate them with respect to the requirements of specific plant applications.
The RSS-6-100 post-LOCA testing observed consisted of the a.
following:
(1) Removal of samples from LOCA chamber and initial visual examination.
(2) Separation of samples and detailed visual examination.
(3) Straightening of samples, marking lengths to be immersed for the voltage withstand test.
(4) Coiling of samples on mandrels of a diameter equal to 40 times overall sample diameter and visual examination.
(5) Immersion of samples in tap water, one hour soak, 80 VAC/ mil (of insulation wall thickness) voltage withstand test (dielectric. proof test).
(6) Insulation resistance (IR) tests.
(7) Close visual examination.
e t
105 l
ORGANIZATION: THE ROCKBESTOS COMPANY NEW HAVEN, CONNECTICUT REPORT NO.-
99900277/86-01 INSPECTION RESULTS:
. PAGE 4 of 6 b.
Test Results:
dielectric withstand (DEWS) tests are as follows:The sam Sample Description Thermal Aging DE_WS Al RSS-6-100A/LE Coaxial Aged Pass A3 RSS-6-100A/LE Coaxial Unaged Pass B2 RSS-6-104/LE Coaxial Aged Pass B4 RSS-6-104/LE Coaxial Unaged Pass C2 RSS-6-109/LE Triaxial Aged Fail C3 RSS-6-109/LE Triaxial Unaged Pass 01 RSS-6-112/LE Twinaxial Aged Conductor A Conductor B Fail D3 RSS-6-112/LE Twinaxial Unaged Pass Conductor A Conductor B Pass E2 RSS-6-113/LE Coaxial Aged Pass Pass E4 RSS-6-113/LE Coaxial RSS-6-110A/LE Coaxial Unaged Pass F2 Aged Pass F3 RSS-6-110A/LE Coaxial RSS-6-115/LE Triaxial Unaged Pass G2 Aged Pass G4 RSS-6-115/LE Triaxial Unaged Pass NOTE:
All samples received 200 Mrads gamma radiation and thermal aging on two of each type was 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> (dry oven) at 100 C.
With the samples in the DEWS tank, water was noted dripping from some of the cable ends.
voltage throughout the test and passed the DEWS ~testAll of the c Immediately prior to the DEWS on triaxial sample C2 and found to be about 30 kilohms. insulation resistance (IR
,its er When the DEWS test was begun, the leakage current increased rapidly and tripped the current set.
Subsequent attempts to slowly apply the DEWS test voltage resulted in voltage up to about 1000 VAC (hi h recorded) before tripping.
None of these attempts was g est successful.
The leakage current fluctuated radically at different voltages and different times, possibly indicating moisture in the cable.
The IR after the DEWS attempt had increased to about 500 kilohms.
for conductor A of twinaxial sample D1.Similar behavior was observed IR was about 500 kilohms.
Before the DEWS, the high as 1100 VAC, and the final IR was about 1.5 megohmsS 106
ORGANIZATION: THE ROCKBESTOS COMPANY NEW HAVEN, CONNECTICUT i
REPORT INSPECTION NO.-
99900277/86-01 RESULTS:
PAGE 5 of 6 c.
Other Observations: The condition of the RSS-6-100 samples following qualification test series, most notably after the LOCA simulation, was similar to that observed in previous tests of this series. Detailed examination of the samples indicated deposits of a dark, powdery substance, particularly at points of contact between the cables and the mandrel. Some shallow indentations in the insulation of the cables where they rested on the mandrel were identified.
No other visible damage to cable insulation was found.
No nonconformances were identified during observation of this set of tests.
2.
The NRC inspectors reviewed qualification test plan TP 5802, Rev. 3 for Rockbestos adverse service coaxial, twinaxial, and triaxial i
RSS-6-100 series cables. The following comments pertain to the l
documents examined:
0 a.
TP p. 6: TP 5802 specifies IR galues of 10 megohms /1000 ft.
for B and C sample types and 10 megohms /1000 ft for the rest.
Withlyft.sgnples,itwillbenecessarytomeasurevaluesin the 10 to 10 megohm range to demonstrate conformance to this specification. Rockbestos is reevaluating the specification.
b.
TP p. 7: Sandia National Laboratories Report, SAND 81-2027/
1 of 2, was cited to support the assertion that synergistic effects of simultaneous radiation and thermal aging have been considered and comparable results would be obtained in using simultaneous or sequential radiation and thermal aging with either sequence, c.
TP p. 11:
Soak time specifications omitted from the previous revision of the TP were included in the current revision.
d.
TP p. 24: The specification for LOCA chemical spray flowrate was given as 15 gpm instead of the correct value of 1.5 gpm.
This error was corrected in the current revision to TP 5802, e.
The documentation included aging calculations which used an activitation energy of 2.75 eV for the type KXL-100 radiation cross-linked polyethylene cable insulation. The inspector verified that the aging calculations were correct for the l
values used, and that the testing used as a basis for the calculations was applicable to the cable currently being tested.
107
ORGANIZATION: THE ROCKBESTOS COMPANY NEW HAVEN, CONNECTICUT REPORT INSPECTION NO.-
99900277/86-01 RESULTS:
PAGE 6 of 6 3.
Review of Test Data The test data generated for TP 5802 prior to the inspection was reviewed. -The two samples which failed the DEWS test had not held voltage throughout the post-LOCA exposure. Sample C2 held voltage for about 40 days and sample D1 held voltage for about 65 days.
irs prior to and during the test ranged as follows:
Temperature ( F)
IR Range (megohms) 6 Ambient prior to test 0.7 - 5.0 X 10 342 12 - 94 322 28 - 180 302 64 - 350 251 560 - 2600 3
227 1.7 - 14 X 10 The two sequential ramp times were similar at about twelve seconds to 280 F and forty-five seconds to 370 F.
During the first ramp, two voltage sets tripped and would not reset. The lead wires were found shorted to the autoclave due to insulatian damage resulting from a pressure extrusion. They were replaced prior to the second ramp.
4.
Failure Analysis Rockbestos is investigating the DEWS test anomalies, i.e., failure of this test by Sample C2 and conductor A of Sample D1.
Preliminary indications are that the samples contained moisture. The intrusion is presently believed to be attributed to insulation damage during test pressure extrusions and subsequent repairs.
4 108
ORGANIZATION: TELEMECANIQUE, INC.
WESTMINISTER, MARYLAND REPORT INSPECTION INSPECTION
)
N0.: 99901011/85-01 DATE: 12/17,31/85 ON-SITE HOURS:
20 CORRESPONDENCE ADDRESS:
Telemecanique, Inc.
ATTN: Mr. J. V. Erhardt, Vice President Operations Engineered Controls 2002 Bethel Road Westminister, MD 21157 l
ORGANIZATIONAL CONTACT:
M. Fenneteau TELEPHONE NUMBER:
(301) 876-2214 NUCLEAR INDUSTRY ACTIVITY: Telemecanique manufactures Motor Control Centers used for various applications in nuclear power plants.
ASSIGNED INSPECTOR:
[
Mw 3/2w/P(
Date J. B/Jacpop(Reactive Inspection Section (RIS)
OTHER INSPECTOR (S):
K. R. Naidu, RIS (12/17/85 only) 24!f4 APPROVED BY:
E. W. Merschoff, ief, RIS, Vendor Program Branch Date INSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR 50 Appendix B, 10 CFR Part 21 B.
SCOPE: This inspection was conducted to evaluate Telemecanique's corrective action for defects previously identified with Gould (now Telemecanique) Motor Control Centers. The items inspected included closing failures in 100 amp frame size circuit breakers in addition to trip specification changes, broken barrier plates, and undersized contact carriers in size 1 and 2 Motor Control Starters.
PLANT SITE APPLICABILITY:
Beaver Valley 1 and 2 (50-334, 412),
Pilgrim 1(50-293), Millstone 1, 2, and 3 (50-245, 336, 423),
River Bend 1 and 2 (50-458, 459), and Seabrook 1 and 2 (50-443, 444),
109
ORGANIZATION: TELEMECANIQUE, INC.
WESTMINISTER, MARYLAND REPORT INSPECTION NO.-
99901011/85-01 RESULTS:
PAGE 2 of 5 A.
Violations:
1.
Contrary to Section 21.21 of 10 CFR Part 21 and Gould Quality Assurance Procedure (QAP) No.15.3, Telemecanique (formerly Gould) did not properly evaluate or report a potential Part 21 defect concerning cracked plastic barrier plates in size 1 and 2 Combination Reversing Motor Starters.
(85-01-01)
B.
Nonconformances:
1.
Contrary to Criteria V and VIII of Appendix B to 10 CFR Part 50, terminal screws used in the wiring of size 1 and 2 Combination Reversing Motor Starters were omitted from material lists used in the fabrication of those units.
Consequently, incorrect (too long) screws were used which punctured the plastic barrier plate and subsequently jammed the starter.
(85-01-02) 2.
Contrary to Criteria VII of Appendix B to 10 CFR 50 Telemecanique (formerly Gould) did not assure that circuit breakers purchased from ITE Electrical Products conformed to the procurement documents.
Specifically, Telemecanique (formerly Gould) did not assure that ITE had a valid 10 CFR 50 Appendix B and 10 CFR Part 21 program in effect.
(85-01-03)
C.
Unresolved Items:
None.
D.
Status of Previous Inspection Findings:
1.
(0 pen) Nonconformance (84-01, Item B.1) Gould did not ensure that Siemens-Allis, a subvendor, developed and implemented a quality assurance program to meet the requirements of 10 CFR Appendix B as stated in Gould's Purchase Orders (P0s). This item is considered open pending further evaluation.
2.
(0 pen) Nonconformance (84-01, Item B.2) A test program was not established for the process of dedicating components received as commercial grade items from subvendors for use in safety related Class 1E components intended for installation in nuclear power plants.
This item is considered open pending further evaluation.
110
ORGANIZATION:
TELEMECANIQUE,INC.
WESTMINISTER, MARYLAND REPORT INSPECTION NO.-
99901011/85-01 A.._
RESULTS:
PAGE 3 of 5 3.
(Closed) Nonconformance (84-01, Item B.3) There was a lack of documentation to show that Gould, Finksburg, pursued the resolution of adverse audit findings identified in letter QAL-820223 dated February 23, 1982, to Siemens Allis, Belle-Fontaine, Ohio.
Telemecanique letter serial E095 dated July 23, 1985 to the NRC stated the corrective action response to the above audit findings was received from Siemens-Allis.
E.
Other Findings or Comments:
1.
On March 5, 1984, Public Service Company of New Hampshire reported to the NRC a problem with Gould supplied size 1 and 2 Combination Reversing Motor Starters. The problem consisted of the use of improper terminal screws for terminations of internal wiring. The screws were too long and when fully tightened cracked a plastic barrier in the contactor assembly. The resulting plastic chips then caused the contactor to jam. An inspection of 105 starters at Seabrook Units 1 and 2 revealed 12 with plastic scored and 11 with plastic broken by the screw.
Information supplied by Telemecanique revealed that the above starters were manufactured between April 1980 and July 1981.
Similar starters were supplied to River Bend within the same time period. When asked to review Telemecanique's (formerly Gould) Part 21 evaluation on this subject the inspector was presented with two pages of undated, unsigned notes which inadequately addressed the problem. Specifically, the evaluation failed to address the question of whether the same screws were used on contactors manufactured during the same time period and subsequently installed at River Bend.
It appears that due to a lack of evidence indicating otherwise, notification to River Bend should have been made as to the possible existence of the problem at their facility.
Violation 85-01-01 and Nonconformance 85-01-02 are identified in Section A and B of this report as a result of Gould's failure to follow Section 8.3 of QAP 15.3, of their Quality Assurance procedures which states that a detailed investigation and written report should have been made.
i The use of the improper screws was found to be the result of not including screw size information on material lists used during l
manufacture of the starters.
Nonconformance 85-01-01 is identified in Section B of this report as a result of Gould's failure to follow OAM 8 of their Quality Assurance Manual which states that controls should have been established to ensure only correct and accepted items were used in the manufacture of the starters.
111
ORGANIZATION: TELEMECANIQUE, INC.
WESTMINISTER, MARYLAND REPORT INSPECTION NO.*
44401011/85-01 RESULTS:
PAGE 4 of 5 2.
On November 8, 1983, Public Service of New Hampshire reported to the NRC a defect in molded case, 100 amp Frame E2, E4, E6, HE4, and HE6 multi-pole circuit breakers installed in 480 volt motor control centers supplied by Gould.
The defect concerned the failure of one of the three poles to close when the circuit breaker was switched from the "off" to "on" position.
Gould had reported this problem to the NRC on September 26, 1983 under 10 CFR Part 21.
The circuit breakers in question were manufactured by ITE Electrical Products and sold to Gould as commercial grade equipment.
In a letter dated September 20, 1983 from ITE to Gould, ITE stated the above problem had been discovered in January 1982 and corrected in February 1982, 22 months prior to Goulds 10 CFR Part 21 report.
ITE also stated that the subject breakers had not been qualified for nuclear use.
ITE/Gould's corrective action to the closing problem appears to have been adequate but a time lapse of 22 months before reporting the problem was unsatisfactory. Although Gould purchase orders to ITE specified 10 CFR 50 Appendix B and 10 CFR Part 21 requirements it appears ITE had no such program in effect.
In a f
letter dated August 20, 1984 from ITE's Engineering Manager to Gould, ITE stated its ability to comply with 10 CFR 50 Appendix B and 10 CFR Part 21 as " unknown." Nonconformance 86-01-03 is identified in Section B of this report as a result.
3.
On March 15, 1982, Gould reported to the NRC under 10 CFR Part 21 a problem with undersized contact carriers in size 1 and 2 motor control starters. The undersized carriers were causing the contactors to jam and subsequently burning out the contactor coils.
The Part 21 report stated that only starters with date codes between May 11, 1981 and March 11, 1982 were suspected and recommended that utilities inspect Motor Control Centers in which suspected faulty starters were known to have been installed.
Since only Motor Control Centers where suspect starters were known to have been installed were identified, all Motor Control Centers were not inspected at this time.
On November 18, 1985, Public Service of New Hampshire reported to the NRC that the same carrier problem originally reported in March 1982 was reoccuring.
Similar problems were recently discovered at Millstone Unit 3.
112 I
ORGANIZATION: TELEMECANIQUE, INC.
WESTMINISTER, MARYLAND REPORT INSPECTION NO.-
99901011/85-01 RESULTS:
PAGE 5 of 5 Upon inspection, the carrier problem seems to have resulted from the shrinking and bending of the plastic carrier material immediately after the carriers were molded. This problem still exists but Telemecanique is now inspecting 100% of the carriers for proper dimensions before and after assembly into the starter units.
The reason for the problem reoccurrence seems to result from the fact that utilities only inspected those motor control centers where suspected faulty starters were known to have been installed as reported by Gould.
Starters with suspect date codes may have then been subsequently installed in Motor Control Centers not thought to have originally contained the faulty units. The affected plants are Beaver Valley 1 and 2, Seabrook 1 and 2, Millstone 3, River Bend 1 and 2, and Pilgrim 1.
4.
In April 1981, Gould made a change to the published specifications covering the adjustable trip range values at which size 0-4 motor starters could be set. The Gould changes were relatively small and with the exception of one coil which was slightly modified, consisted of simply relabeling the coils and publishing new specification sheets. The change was a result of a change in the Underwriters Laboratory factory calibration specification for testing after manufacture from -30%/+10% to 20%.
Public Service Co. of New Hampshire expressed a problem with this change due to the fact that some Motor Control Centers manufactured for Seabrook Stations Units 1 and 2 had been fitted with starters manufactured before the specification change. Additional motor control centers were to be supplied with the newly labeled starters.
Gould has relabeled all starters manufactured for Seabrook before the specification change and has exchanged trip coils as required to bring the starters within the utility's calibration requirements.
e 113
ORGANIZATION: TEXAS INSTRUMENTS, INC.
JOHNSON CITY, TENNESSEE REPORT INSPECTION INSPECTION N0.: 99901028/85-01 DATE(S): 8/28, 29/1985 ON-SITE HOURS:
12 CORRESPONDENCE ADDRESS: Texas Instruments ATTN: Mr. M. McDonnell Vice President, Industrial Systems Division Erwin Highway Johnson City, Tennessee 37605-1255 ORGANIZATIONAL CONTACT:
R. Carper TELEPHONE NUMBER:
615-461-2056 PRINCIPAL PRODUCT: Recorders and Process Instruments.
NUCLEAR INDUSTRY ACTIVITY: Less than 1% of the orders are nuclear related.
i i
ASSIGNED INSPECTOR:
K. R. Naidu, Reactive Inspection Secion (RIS)
Date l
OTHERINSPECTOR(S):
i 2
d APPROVED BY:
E. W. Merschoff fChief, RIS, Vendor Program Branch Date y
INSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR Part 21 and 10 CFR 50 Appendix B.
B.
SCOPE: Review the implementation of Texas Instruments' quality assurance program and 10 CFR Part 21 reporting program.
PLANT SITE APPLICABILITY: Trojan Nuclear Power Plant (50-344).
115
ORGANIZATION: TEXAS INSTRUMENTS, INC.
JOHNSON CITY, TENNESSEE REPORT INSPECTION NO.-
99901028/85-01 RESULTS:
PAGE 2 of 4 P
A.
Inspection Issues Portland General Electric Company issued Purchase Order (P0) Nos.
N-29235, 29244 and 29235 for the supply of ten TIGRAPH 200 type recorders. The P0 required that the recorders meet the requirements of IEEE-323 and IEEE-344 but did not invoke compliance to 10 CFR Part 21 or 10 CFR 50 Appendix B.
Texas Instruments (TI) supplied ten TIGRAPH 200 type recorders to Portland General Electric Company (PGE), Portland, Oregon for installation at the Trojan Nuclear Power Plant (Trojan). TI issued Certificates of Conformance (CoC) stating that the recorders were qualified to IEEE-323 and IEEE-344 standards. The objective of the inspection was to determine the validity of the CoC.
B.
Background Information TI manufactures programmable controllers and recorders at this manufacturing facility for use at chemical processing plants. TI recently introduced the TIGRAPH 200 process recorder.
This equipment can record analog signals from one to six input channels on heat sensitive paper.
Printing is accomplished with a fixed solid state thermal print head having a horizontal print resolution of 100 dots per inch. All moving parts are confined to the chart drive system. TI stated that they received several inquiries from owners of nuclear power plants concerning these recorders.
Florida Power & Light Company and Tennessee Valley Authority performed pre-award surveys but did not qualify TI as an approved vendor due to inadequacies of TI's quality assurance program.
C.
Inspection Results 1.
The TI QA manual does not comply with criteria I, II, X, and XVIII of 10 CFR 50 Appendix B.
Therefore, the validity of the CoC is questionable to the extent that there is no assurance that recorders manufactured will meet the quality of the recorders which successfully withstood environmental and seismic tests.
2.
TI had not developed and implemented a procedure to evaluate deviations or inform the licensee or purchaser of the deviations in order that they may evaluate the deviations.
However, the requirements of 10 CFR 21 were not imposed on TI by PGE.
116
ORGANIZATION: TEXAS INSTRUMENTS, INC.
JOHNSON CITY, TENNESSEE REPORT INSPECTION NO.
99901028/85-01 RESULTS:
PAGE 3 of 4 D.
Inspection Findings and Other Comments 1.
Plant Tour The TI Product Manager accompanied the NRC inspector on a plant tour of the areas associated with receipt inspection, storage of components, assembly of components on printed circuit boards, wave soldering machines, inprocess inspections, assembly of recorders and test of the recorders. Detailed procedures to test the recorders were available at the work station.
2.
Review of the TI Quality Assurance (QA) Manual Review of the TI QA manual indicated that it does not meet the requirements of 10 CFR 50 Appendix B in the following areas:
a.
Criterion 1.
Section 2 of the QA manual describes the organization of the Industrial Systems Division of the TI facility located in Johnson City.
The organization does not depict the role of quality control activities in the irrplementation of this QA program. Job descriptions of the QC/QA personnel are not specified.
b.
Criterion II - QA Program.
Section 3 briefly discusses the TI program which is supplemented by information provided in sections 4, 5, 6, and 9.
The role of Quality Control is not adequately described in this section.
Furthermore, TI Management's periodic review on the adequacy of the QA program is not specified.
c.
Criterion X - Inspections.
Sections 5.5, 5.7, 6.2, 6.3, 6.9, and 6.13, address the inspections performed during in-process, final and testing. However, the role of quality control is not mentioned.
d.
Criterion XVIII - Audits.
Section 13 describes a system of audits to verify compliance to requirements. However, the frequency of the audits and the qualification of the j
auditors are not mentioned.
Furthermore, management audits l
or independent audits to determine the adequacy of the QA program are not addressed.
117
ORGANIZATION: TEXAS INSTRUMENTS, INC.
JOHNSON CITY, TENNESSEE REPORT INSPECTION NO - 44901028/85-01 RESULTS:
PAGE 4 of 4 3.
Review of Quality Assurance Records a.
The II TIGRAPH department maintains a log in which test data on TIGRAPH recorders are documented by the serial numbers. The CoC furnished to PGE listed the serial numbers of the recorders; the inspector reviewed the test data on these serial numbers and determined them acceptable.
b.
Wyle Laboratories, Huntsv9 1e, Alabama performed a mild environment aging analysis on the TIGRAPH recorder. The report dated July 1985 summarizes that the process recorder will withstand radiation effects of 5.071 E3 rads gamma (10-year total integrated dose at 33 C (91.4 F)) without deleterious effects.
c.
Wyle Laboratories, Huntsville, Alabama, subjected the recorder to a seismic simulation on May 3 and May 8, 1985 and determined that the specimen possessed sufficient integrity to withstand, without compromise of structure or electrical functions, the prescribed random multi-frequency tests.
The CoC is based on the tests performed on a specimen recorder. TI does not have an acceptable QA program which meets 10 CFR 50 Appendix B.
The validity of the CoC is questionable to the extent that recorders manufactured subsequent to the testing may not have the same quality if it is not manufactured under an acceptable QA program.
D.
Persons Contacted
- R. Carper, Product Marketing Manager, Texas Instruments M. Ferguson, Legal Advisor, Texas Instruments D. Schewk, Manager, Quality & Reliability, Texas Instruments R. Wetherall, Manager, Product Assurance, Texas Instruments D. Fricky, Quality & Reliability Engineer, Texas Instruments G. Bawgus, Product Assurance Inspectcr, Texas Instruments
- C. Presnell, Marketing Manager, Texas Instruments
- Denotes those persons who attended the exit interview at the conclusion of the inspection.
E.
Exit Interview The inspector met with individuals identified in Section D and discussed the scope and findings of the inspection.
118
I l
ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION NUCLEAR FUEL DIVISION PITTSBURGH, PENNSYLVANIA REPORT INSPECTION INSPECTION N0.: 99900005/85-01 DATE(S):
11/12-15/85 ON-SITE HOURS: 92 CORRESPONDENCE ADDRESS: Westinghouse Electric Corporation Nuclear Fuel Division ATTN: Dr. Richard Slember, General Manager Post Office Box 355 Pittsburgh, Pennsylvania 15230 ORGANIZATIONAL CONTACT: Mr. R. Cost, Manager of Quality Assurance TELEPHONE NUMBER:
(412)374-2359 NUCLEAR INDUSTRY ACTIVITY: Nuclear fuel assembly supplier for Westinghouse designed reactors.
A /) /7
/
ASSIGNED INSPECTOR:
)
$ + 4 r*/
2k/fd Date R.
L~. Cilimberg, Specpl Projects Inspection Section(SPIS)
OTHER INSPECTOR (S):
R. L. Pettis, SPIS J. C.,, Harper, Reactive Inspection Section APPROVED BY:
36 John W. Craig, Chief, SPJS, Vendor Program Branch Date INSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR 50, Appendix B and 10 CFR 21.
B.
_ SCOPE: This inspection was made to review procedures and practices in the metallurgical and chemical testing laboratories.
PLANT SITE APPLICABILITY:
PWR facilities with fuel supplied by Westinghouse.
119
ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION NUCLEAR FUEL DIVISION PITTSBURGH, PENNSYLVANIA REPORT INSPECTION NO.: 99900005/85-01 RESULTS:
FAGE 2 of 7 A.
VIOLATIONS:
None.
B.
NONCONFORMANCES:
1.
Contrary to Section 17.1.5, " Instructions, Precedures, and Drawings,"
of WCAP 8370, Revision 10A and WCAP 7800, Revision 6A, dated August 1984, and Quality Control Instruction (QCI) No. 108857, Revision 6, dated October 22, 1984, an autoclave was left uncovered when not in use.
(85-01-01) 2.
Contrary to Section 17.1.5 of WCAP 8370, Revision 10A and WCAP 7800, Revision 6A, and QCI No. 108823, Revision 0, rod pressurization ~was determined according to S0I R-0618 and not 501 R-0563 as specified by the QCI.
(85-01-02) 3.
Contrary to Section 17.1.5 of WCAP 8370, Revision 10A and hCAP 7800, Revision 6A, and Product Assurance Procedure (PAP) 2.2.2, Revision 4, dated December 2,1982, management approval sign-off for qualification of a metallurgical laboratory technician was missing from the evaluation sheet.
(85-01-03)
C.
UNRESOLVED ITEMS:
None.
D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
Not applicable.
/
E.
INSPECTION FINDINGS AND OTHER COMMENTS:
1.
Entrance and Exit Meetings Westinghouse management representatives were informed of the scope of the inspection during the entrance meeting. The inspection findings and observations were summarized during the exit meeting on November 15, 1985.
120 l
r
ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION NUCLEAR FUEL DIVISION PITTSBURGH, PENNSYLVANIA INSPECTION REPORT.
RESULTS:
PAGE 3 of 7 NO.: 99900005/85-01 2.
Chemical Laboratory a.
Measuring and Test Instruments The NRC inspectors reviewed procedures and practices for the control of measuring and test instruments used in the analytical laboratory. The chemical analysis procedures (C0CL) specify the steps for the calibration and standardization of test and measurement equipment for each analytical method. Standards are used to verify the accuracy of each analytical method. When standards do not meet the statistical requirements of the analytical method, calibration curves are re-established by analyzing additional standards prior to an analysis being performed by that method.
b.
Procedure Review The NRC inspectors reviewed 14 procedures which covered:
the determination of impurities in uranium, uranium spectrographic standards, isotopic composition, potentiometric titration, x-ray fluorescence, uranium in uranium hexafluoride, fluoride by specific ion electrode, forme + for analytical procedures, technician training program, sample log, uranium dibxide powder blend chemistry and isotopic sampling, excess uranium control, sampling and release of in-process uranyl nitrate, and use of instructions for quality control. These procedures are contained in three categories of documents:
the analytical laboratory manual, product and process procedures (C0CL), and Quality Control Instructions (QCI). The procedures were technically correct and complete. The procedures had also been properly reviewed and contained the required signature approval, c.
Observation of Analyses The NRC inspectors observed technicians performing sample analyses utilizing the procedures discussed above. Observations are itemized below:
l (1) A technician used one of three mass spectrometers to analyze uranium for isotope content in accordance with procedure number C0CL M-01, " Isotopic Uranium Analysis,"
Revision 3, dated November 11, 1985.,The steps of the 121
r l
ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION NUCLEAR FUEL DIVISION PITTSBURGH, PENNSYLVANIA l
~
REPORT INSPECTION l
?
NO.: 99900005/85-01 RESULTS:
PAGE 4 of 7 -
)
procedure were properly followed. A record of analyses p results and standards checked are %ept in a. log book. Any changes to results entered in the log are made with one ~
line crossing out the entry, and the correction is initialed and dated.
(2) A technician determined fluoride content in accordance with' procedure number C0CL P-03, " fluoride by Specific Ion Electrode," Revision 1, dated November 8, 1985.
The steps of the procedure were followed without deviation and
~
standardswereanalyzedtoverifytheaccuracyoftHeythod.
(3) A technician determined the, uranium content in a standard sample in accordance with procedure number C0CL U-01,
" Uranium by Potentiometric Titration," Revision 6, dated October 1, 1985. This procedure is commonly known as the Davis and Gray Method. The technician demonstrated competence during this analysis as he timed the steps of the procedure with a stopwatch and obtained the !.tandard value within the statistical accuracy of the method.
(4) A technician analyzed multiple standards and generated a calibration curve pricr to sample analysis. All analyses were performed acceptably in accordance with procedure number C0CL U-03, " Uranium by X-Ray Fluorescence,"
Revision 3, dated September 7, 1984. A New Brunswick Laboratory (NBL) standard was used for the sample analysis.
The technician was well trained for performing the analysis and he obtained the standard value within the statistical accuracy of the method; d.
Training The NRC inspectors reviewed the training requirements for technicians.
These requirements are specified in attachment 11.1-1 to the Analytical Laboratory. Manual, dated November 15, 1983.
On the job training is used as the primary training method. The technicians are required to analyze 6 series of
" blind" standards prior to qualifying on each method. Training records are maintained for each technician including the analysis the technician is qualified to perform. The' training files of four, full-time technicians and one part time technicien indicated that written tests are used to show proficiency for
.f several methods.
,G
\\
s s._
u2 y
'F r
f i
ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION NUCLEAR FUEL DIVISION PITTSBURGH, PENNSYLVANIA REPORT INSPECTION N0.: 99900005/85-01 RESULTS:
PAGE 5 of 7 e.
Conclusions Based on the above reviews and observations the NRC inspectors determined that the technicians are well trained and are using calibrated equipment in accordance with written procedures.
These procedures are used to perform analyses or, samples to verify the quality of final products and support the quality control of in-process material.
No items of nonconformance w
were identified in this area.
3.
Metallurgical Laboratory (Met. Lab.)
a.
Procedure Review and Observation of Analysis The NRC inspectors reviewed procedures which covered autoclave testing, operation of tensile tester, fuel rod pressurization, metallographic sample preparation, metallurgical records and documentation, corrosion sampling and evaluation, U02 pellet 1
chamfer measurement, and tool and gage inspection.
j (1) During review of QCI No. 108857, " Operating Procedure for Autoclaves," Revision 6, dated October 22, 1984, the NRC inspectors noted that section C.6 requires that autoclaves be covered when not in use. One autoclave not in use in the Met. Lab. was missing a cover. This procedural requirement is a practice to ensure the proper water quality in the autoclave. Control of water quality is also provided by a resistivity test. Met. Lab. personnel put the top on the autoclave when they were informed by the NRC inspectors that it was not in place.
Nonconformance Item 85-01-01 was identifed in this area.
(2) A review of QCI No. 108823, " Fuel Rod Pressurization Records j
Check and Disposition Practices," Revision 0, dated March 26,
?
1985, indicated that rod pressurization be checked using Standard Operating Instruction (S0I) number R-0563, Revision 20, dated January 18, 1983. The NRC inspectors determined I
that S0I number R-0618, Revision 15, dated November 5, 1985, was the document being used by manufactering personnel.
The inspectors reviewed both S01s and determined that rod pressurization checks were acceptable using either S01.
QCI 108823 was corrected to properly reference SOI R-0618 prion to the completion of inspection.
Nonconformance Item 85-01-02 was identified in this area.
[
123
ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION NUCLEAR FUEL DIVISION PITTSBURGH, PENNSYLVANIA r
REPORT INSPECTION NO.: 99900005/85-01 RESULTS:
PAGE 6 of 7 (3) Corrosion testing of fuel pins was performed by a Met. Lab.
technician in accordance with QCI No. 108857. Procedural steps such as the rinse cycle, pH, resistivity measurements, autoclave preparation and start-up, and coupon weighing were followed as required by the procedure.
(4) A review of QCI No. 108860, "U0 Pellet Sample Preparation 2
and Chamfer Measurement," Revision 1, dated July 22, 1985, and observations of the measurement and evaluation of U0, pellet chamfer by the technician indicated that the work was in accordance with the procedure.
(5) The NRC inspectors reviewed the calibration and calibration certificates for scales, measurement scope, tensile tester, microhardness tester, micrometers, and master blocks. Based upon this review the inspector determined that measuring instruments were properly calibrated and traceable to standards of the National Bureau of Standards, b.
Qualification and Training The NRC inspectors reviewed the requirements for qualification of a Met. Lab. Technician Level C.
Certificate of Qualification, Q.C. Form 423, Revision 1, dated February 1, 1985, requires that a technician be certified in accordance with PAP 2.2.2 which further requires that the technician make metallographic evaluations including micrographic evaluations. This review revealed that the Indoctrination and Evaluation Sheet was completed by the technician and signed off by the appropriate manager. However, the manager's signature was missing for visual metallographic examinations and, therefore, micrographic examinations. The inspectors determined that the technician was qualified to perform microscopic evaluations and knew how to use the microscopic quality standards.
Nonconformance Item 85-01-03 was identified in this area.
c.
Conclusions Based on the above reviews and observations the NRC inspectors determined that the technicians in the Met. Lab. are using calibrated equipment in accordance with written procedures other than the exceptions noted.
Further, corrective action was initiated for the items of nonconformance identified in the Met. Lab. prior to the conclusion of this inspection.
124
ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION NUCLEAR FUEL DIVISION PITTSBURGH, PENNSYLVANIA REPORT INSPECTION NO.: 99900005/85-01 RESULTS:
PAGE 7 of 7 F.
PERSONS CONTACTED:
K. Barrett, Level B Technician, Chemical Lab E. Black, Level A Technician, Chemical Lab
- D. Brown, Level C Technician, Metallurgical Lab
- J. R. Bush, Quality Control Manager H. Cory, Mechnical Inspection Manager
- R. R. Cost, Operations Product Assurance Manager
- J. E. Hart, Process Engineering Manager K. Hunter, Ceramic, Metallurgical Lab
- L. D. Kays, POPA Manager
- E. Keelen, Manufacturing Manager F. Kulas, QC Supervisor A. LeGrand, Chemist G. Lindler, Level B Technician, Metallurgical Lab
- C. Perkins, Quality Control Manager T. Pham, Chemist i
- R. K. Pollard, Quality Control Engineering Manager
- A. M. Schwartzman, OPTIM Director, Metallurgical Lab J. Snyder, Level C Technician, Metallurgical Lab i
B. Torrey, Wet Chemist E. Waters, Level B Technicial, Chemical Lab C. Wessinger, Level B Technician, Metallurgical Lab 1
M. Wessinger, Level B Technician, Chemical Lab l
H. Whitaker, Chemist
- G. R. Wilson, QC Supervisor, Chemical Lab
- G. D. Workman, Analytical Services Manager
- present during exit meeting J
l l
l 125
ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION PITTSBURGH, PEN?iSYLVARIA I
REPORT INSPECTION LINSPECTION fl0. - 99901043/89-01 DaTE-11/10-14: 01/11-j6 nN-SITE HOURS?
o's CORRESPONDENCE A.0 DRESS: Westinghouse Electric Corporation Nuclear Services Integration Division ATTN:
T. Christopher, General Manager Monrceville Pittsburgh, Pennsylvania 16230 ORGANIZATIONAL CONTACT:
M. A. Marlyn TELEPHONE N'JMBER:
412-374-7504 NUCLEAR INDUSTRY ACTIVITY:
Refurbishment of 480 vclt circuit breakers used i
as Reacter Trip Breakers including installation of Undervoltage Trip j
Attachments.
1 E
ASSIGNED INSPECTOR:
~
e K. R. Naidu, Reactive Inspection SectI6i(RIS)
~0 ate OTHER INSPECTOR (S):
J. B. Jacobsor.,,P vember10&l6,1986) dbV!R APPROVED BY:
E. W. Merschoff. Chi Tris,VencorProgramBrancTi
~
Date INSPECTION PASES AND SCOPE:
A.
BASES:
10 CFR Part 21, Appendix 8 to 10 CFR 50.
B.
SCOPE: Witness tests (11/10-14/85) of UnderYoltage Trip Attachments
]
(UVTAs) which failed at D.'C. Cook 2, review the manufacturing history and avality assurance records of the replacement UVTAs, obtain additional information on broken spring latches observed at Watts Bar and the mal-1 functioned UYTAs reported at Kewaunee on June 8, 1984 and November 7, 1985.
PLANT SITE APPLICA~BILITY:
D.C. Cook Units I and 2 (50-315; 50-316),
Kewaunee (50-305), Watts Bar (50-390; 50-391).
l 127 l
l
ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION PITTSBURGH, PENNSYLVANIA REPORT INSPECTION NO.:
99901043/85-01 RESULTS:
PAGE 2 of 10 A.
Violations No violations were identified during this inspection.
B.
Nonconformances No nonconformances were identified during this inspection.
C.
Unresolved Items No unresolved items were identified during this inspection.
D.
Status of Previous Inspection Findings The items of noncompliance discussed below were identified when the manufacturing operations were located at East Pittsburgh and are discussed
(
in Inspection Reports 99900330/83-01 and 83-02.
Since then, the activities E
for the manufacture of switchgear components and accessories for nuclear power plant applications have been transferred to the Seco Road facility (Docket No. 99901043). Documentation on actions taken by Westinghouse could not be retrieved readily. Westinghouse agreed to consolidate the information in an auditable form for the following items.
This information will be reviewed during a subsequent inspection.
1.
(0 pen) Violation (83-01):
The violation identified that a current copy of 10 CFR Part 21 was not posted. This violation remains open l
pending review of measures established to prevent recurrence.
2.
(0 pen) Nonconformance (83-01, Item A): The nonconformance identified that from the available inspection records it could not be established that 21 replacement UVTAs received 100 percent quality control checks. This item remains open pending verification that adequate measures have been established to prevent recurrence.
3.
(0 pen) Nonconformance (83-01, Item B):
The shop order information issued for replacement UVTAs did not identify the applicable drawing revisions to be used for the required 100 percent inspection of
" critical items." Changes were verified to be documented in Shop Order 02VN202 on August 8, 1983.
This nonconformance remains open pending verification that adequate measures have been established to prevent recurrence.
4.
(0 pen) Nonconformance (83-01, Item C): The nonconformance identified that Certificates of Conformance were on file for only five replace-ment UVTAs of the eight that have been shipped on Purchase Order No. 54X470254 to the McGuire site.
128
ORGANIZATION: WESTINGH0USE ELECTRIC CORPORATION PITTSBURGH, PENNSYLVANIA REPORT INSPECTION NO.-
99901043/85-01 RESULTS:
PAGE 3 of 10 This nonconformance remains open pending verification that Certificates of Compliance for the remaining three are available.
E.
Inspection Findings and Other Comments 1.
Review of UVTA Failures The inspector reviewed the status of the reported failures of UVTAs.
UVTAs actuate to trip 480 volt circuit breakers used as ReactorTripBreakers(RTBs).
a.
Background Information on UVTA Failures As a result of the February 22 and 25,1983, anticipated g
transients without scram (ATWS) events which took place at the Salem nuclear power plant, the NRC issued Bulletin 83-01 and s
formed a task force to assess the generic implications of these events. During these ATWS events, the UVTAs on Westinghouse type 08-50 RTBs malfunctioned and caused the breakers to fail to open. On March 11, 1983, Southern California Edison reported that three RTBs at San Onofre Unit ? and one of Unit 3 failed to open during testing of the UVTAs. These RTBs were manufac-tured by General Electric Company. As a result, the NRC issued Bulletin 83-04.
The actions of the task force resulted in the issuance of NUREG 1000 titled " Generic Implications of the ATWS Events at the Salem Nuclear Power Plant" and Generic Letter 83-28, delineating procedural and plant changes required.
Findings in NUREG-1000 were based, in part, on assurances that improved maintenance of the circuit breakers would improve the reliability of the UVTAs to an acceptable level for the short term, after which the longer term corrective actions would be implemented.
The longer term corrective actions which were directed by NRC Generic Letter 83-28, had two major aspects related to the RTB/UVTAs:
(1) the installation of a plant modification that provides for the automatic activation of the shunt trip coi1 of the RTB for any automatic reactor trip signal, and (2) relia-bility improvements in the RTB/UVTA based upon UVTA life testing by the vendor. The life test formed the basis for a UVTA replacement interval of 1250 cycles and a lubrication interval of 200 cycles. Westinghouse issued a purchase order to their East Pittsburgh facility to manufacture 187 replacement UVTAs intended to replace the existing ones.
129
ORGANIZATION: WESTINGHOUSE ELECTRfC CORPORMiON PITTSBURGH, PENNSYLVANIA e
REPORT INS 2ECTION MO.-
99901043/85-01 RESULTS:
PAGi 4 of 10 The recent failures at D.C. Cook Unit 2 involved PTBs that were refurbished by Westinghouse in July 1985 at the Assembly and Test (A&T) facility located at Seco Road, Monroeville The refurbishment included the installation of new UVTAs. The Unit 2 RTS "A" that failed on October 29, 1985 experienced approxfrrately 75 operations subsequent to the refurbishment.
The Unit 2 RTB "B" that failed the margin force test on November 3,1985 experienced approximately 30 operations subsequent to the refurbishment. In this short time, both UVTAs suffered a loss of their trip output force.
b.
UVTAs supplied to Kewaunee Westinghouse supplied to Kewaunee nuclear power plant (Kewaunee) four UVTAs with serial numbers 02Y-N213 - 8, 9,10 and 11 with l
operating coils rated for 48 volts DC along with Quality Release l
(QR) 70517 dated Fpril 16, 1984.
Kewaunee held these UVTAs in storage until June 1984 when they were assembled on RTBs and tested.
The UVTA with serial number 02Y-N213-9 nalfunctioned. The UVTA was returned to Westinghouse. Examination of this UVTA concluded that some of the parts en the UYTA had been replaced. Westinghouse notified the NRC in a letter NS-EPR-2928 dated June 8, 1984 that
{-
they performed an evaluation in response to the component malfunction, and determined that some of the components in the UVTA had excessive wear and therefore could not hnve been supplied by them.
Westinghouse further stated that thef examined their quality assurance records associated with the prodoction of the UVTA ar;d determined that the UVTA fully conformed to the inspecticn criteria when it was examined prior to placement in a sealed carton and shipped to Kewaunee. Westinghouse stated that the defective UVTA had been reworked utilizing used and cutd3ted p.1rts without their knowledge.
c.
UVTA Failure at D.C. Cock l
On October 29, 1985, the reactor trip breaker (RTB) associated with Train "A" reactor protection system failed to trip the O.C. Cook Unit 2 nuclear power plant due to the wilfunctioning of the VVTA.
The serial number of the UVTA is 02Y-N212-88. The RTB associated with Tr.ain "B~" functioned properly and tripped the plant. However, on November 3,1985, the same RTB failed to pass the UYTA force margin test.
(The RTBs are 480 volt DB-50 type circuit breakers which were manufactured by Westinghouse in the late 1950s and refurbishedbyWestinghouseinJuly,1986.) The serial number of 130
OPGANIZAT10N: WESTINGHOUSE ELECTRIC CORPORATION PITTSBURGH, PENNSYLVANIA F-REPORT INSPECTION NO.-
99901243/85 01 RESULTS:
PAGE 5 of 10 l
this UVTA is 02Y-N212 86. These UVTAs which are rated for 125 volts l
DC cperation were reroved from their respective RTBs and sent to the Assembly and Test Facility of Westinghouse located at Seco Road facility in Monroeville for further tests and evaluation. During November 11-15, 1985, these two UVTAs along with another UVTA taken from storage were subjected to extensive tests.
Representatives of American Electric Power Company (AEP) participated in the test program.
Representatives of the NRC observed the tests. The NRC inspectors observed a weld bead on the leading edge of the UVTA which failed (serial number 02VN 212-88) which restricted the free movement of the trip lever spring. Additional tests conducted by Westinghouse confirmed this as a probable cause. Westinghouse issued a Maintenance Advisory letter NS-NRC-85-3084 dated November 21, i
1985, to all users of DB-50 type circuit breakers as RTBs, describing the D.C. Cook Unit 2 event and advising users to visually inspect the UVTAs for anomalies such as weld beads or notches or grooves in.
the leading edges of the UVTA base that would prevent the successful i
tripping of the RTBs. Westinghouse representatives informed the NRC inspector during this inspection that the tests on the UVTAs were complete and that they sent a draft of the evaluation report to AEP.
d.
UVTA Failure at Kewaunee On November 7, 1985. Kewaunee performed tests on 08-50 type RTBs as required by NRC Bulletin 85-02, which was issued following the RTB failure experienced at D.C. Cook Unit 2, to establish the satisfactory functioning of the UVTAs for those plants which had not installed the automatic shunt trip. Two of the five RTBs (one spare) failed the 20 ounce margin test prescribed in the Bulletin. The UVTAs were lubricated after discussions with NRC personnel and the RTBs tested satisfactory. These UVTAs have coils rated for 48 volts DC.
2.
Review of 50.55(e) Report On July 26, 1985, Tennessee ValTey Authority reported a Construction Deficiency (50.55(e)) to the NRC. The report stated that mainte-nance personnel at the Watts Bar nuclear power plant observed broken spring release latches on 480 volt DS-416 and DS-206 type circuit breakers manufactured by Westinghouse. The device is a metal piece which releases the charging spring permitting the circuit breaker to close.
Upon receipt of an electrical signal to close the circuit breaker, a solenoid device is energized. A flapper device on the solenoid hits the spring release device, which in turn releases the charging spring.
Twenty five spring latches out of a 131 i
ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION PITTSBURGH, PENNSYLVANIA REPORT INSPECTION NO.-
99901043/85-01 RESULTS:
PAGE 6 of 10 total of 80,000 units manufactured were reported to have been broken between 1972 and 1984. Westinghouse traced the failures to the manufacturing periods of early 1975, April 1976 and early 1978, and attributed the failures to incorrect heat treatment of the device.
Westinghouse issued a Technical Bulletin NSID-TB-85-17 on August 14, 1985 and recommended the following corrective actions.
(a) Advise the responsible personnel that the breaker may be closed manually using the operator on the front panel of the breaker if the electrical closing does not occur on demand.
(b) For applications other than Westinghouse designed reactor trip function, the user should evaluate the situation to determine if this condition could affect safety.
(c) Inspect the latch levers during normal scheduled maintenance /
inspection of the DS type circuit breakers.
Corrective action taken to preclude repetition was to increase the radii on the curved surface of the device and lower the minimum acceptable Rockwell hardness limit.
3.
Review of the Purchase Order for Replacement UVTAs The Nuclear Systems Integration Division (NSID) of Westinghouse issued purchase order RPS 10177 dated December 29, 1983 to Westinghouse Electric Supply Company (WESCo) for the supply of 187 UVTAs. WESCO reissued the order as purchase order RPS 20054 dated January 9,1984 to Westinghouse East Pittsburgh, where the UVTAs were manufactured. The following information related to the items listed in the above purchase order, quantity for each item, Quality Release (QR) number, and the date of release:
a.
UVTAs with 48 volts DC coils drawing #23A9019G-75 Quality Item Release
-Date Number Quantity Number Released 1
20 N-69172 3/5/84 1A 20 N-70516 3/16/84 IB 20 N-71209 5/4/84 N-70537 1C 20 N-70520 5/4/84 10 10 N-70538 5/4/84 132
ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION PITTSBURGH, PENNSYLVANIA REPORT INSPECTION NO.-
99901043/85-01 RESULTS:
PAGE 7 of 10 b.
UVTAs with 125 volts DC coil drawing 23A9019G-61 Quality Item Release Date Number Quantity Number Released 2
20 N-69189 3/5/84 N-69174 3/28/84 2A 20 N-71210 4/16/84 2B 20 N-70539 5/23/84 2C 20 N-70540 5/23/84 2D 17 N-70541 5/23/84 j
c.
Document Submittal Forms (DSFs) were used as vehicles to review and approve procedures and drawings.
Change order 001 dated February 1,1984 to Purchase Order RPS 10177 conveyed the ap-proval of DSF E-9719 which in turn approved the design document review for UVTA, critical drawing list and Order Handling Proce-dure NQD-377-01 Revision 1.
d.
Change order 002 dated March 5, 1984 released seven UVTAs for DB-50 circuit breakers. Two were shipped to Virginia Electric Power Company (Surry 1), four to Carolina Power and Light Company (Robinson-2) and one was put in storage.
e.
Change Order 003 dated April 9, 1985 was issued to approve DSFs E-1027 and E-6230 which approved procedure NQD-377 Revision 2.
The following changes were incorporated in this revision:
(1) A typographical error was corrected.
Item 18 (latch) on the " Critical Parts" (CP) list was corrected from 309C010H01 to 309C010H02.
(2) Item la was added to the CP list to distinguish between the UVTAs with 48 volts and 125 volts DC coils.
(3) The drawing part numbers for the UVTAs with 48 volt and 125 volt DC coils were added to provide clarity.
f.
NSID QA personnel from Monroeville inspected the batches of UVTAs as they were being completed at the East Pittsburgh facility and witnessed the testing of the devices at random.
133
ORGANIZATf0N: WESTINGHOUSE ELECTRIC CORPORATION PITTSBURGH, PENNSYLVANIA 1
REPORT INSPECTION NO.-
99901043/85-01 RESULTS:
PAG 6 8 of 10 4.
Review of the UVTA Quality Assurance Records a.
One hundred and eighty-seven UVTAs were manufactured at the Westinghouse East Pittsburgh facility in various batches of 20, t
10 and 7.
The Quality Assurance records which consist of the quality control checklists including electrical tests are available at the Assembly and Test (A&T) facility located at Seco Road, Monroeville. The records indicate that whenever a batch of UVTAs were manufactured the appropriate quantity of the critical components were retrieved from storage and inspected. A single quality control instruction (QCI) 37 lists twenty components considered critical for the assembly of the UVTAs, the drawing number for the items and the quantity. A signoff on the checklist indicates that 100% of the items were inspected and determined acceptable. On completion of the assembly, functional tests were performed including a dielectric test.
Items not identified as critical components are listed on QCI-157.
Records indicate that NSID QA personnel went to the East Pittsburgh facility to inspect the final UVTA assemblies and selectively witness the dielectric tests. A quality release (QR) was issued to document the serial numbers of the UVTAs accepted. The UVTAs were then packed in individual boxes for shipment.
b.
The inspector reviewed QR D-70538 Revision 0 dated May 4, 1984.
The QR stated that for Purchase Order RPS 10177, ten UVTAs with serial numbers 02YN212-82 through and including 91 were released for storage at WESCO in Monroeville. The QR listed the attributes which were inspected and determined acceptable, such as, visual inspection, cleanliness, operating electrical test record, performance test record, packaging, Certificate of Conformance, Certificate of Quality and Certificate of Critical Components.
c.
QC D-70538 Revision 1 dated March 22, 1985 indicated that UVTAs serial number 02YN212-82, 83, 84 and 85 were released for installation to Tennessee Valley Authority circuit breakers.
The intent of the revision was to document that six UVTAs were left in storage at that specific time, j
d.
QR 0-70538 Revision 2 dated July 29, 1985 indicates that six UVTAs in stock were released to A&T facility.
e.
Trip Report NPQA-84-147 dated May 7, 1984 documents the final surveillance performed on a batch of thirty-one UVTAs.
100%
t inspection was performed on the following attributes:
13 4
ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION PITTSBURGH, PENNSYLVANIA REPORT INSPECTION NO.-
99901043/85-01 RESULTS:
PAGE 9 of 10 (1) All tru-arc retaining clips (2) Captive pin design ("c" ring)
(3) Machine finish of trip latch (4) Tack welding (5) Surface finish (6) Serial number (7) Witnessing test per Operating Procedure 920071 revision B on four UVTAs with serial numbers 0YN212-61, 75, 84 and 91.
The test results were acceptable.
The serial numbers of the thirty-one were 02YN212-61 through 91.
5.
Review of the D.C. Cook Refurbishment Records The RTBs from D.C. Cook Units 1 and 2 were refurbished at the A&T facility. This refurbishment consisted of inspecting the RTBs, documenting the initial findings, replacing unacceptable parts, replacing UVTAs, and documenting the results of final inspections.
A shop traveller was used to document the inspections.
Procedure CGG-02-1985-1 Revision 1 was used to perform the inspections.
Review of the records indicate that all RTBs were in acceptable operating condition before shipment.
F.
Persons Contacted Westinghouse Electric Corporation Nuclear Services Integration Divison (NSID)
Replacement Components Service (RCS)
J. Epstein, Manager Equipment Qualification Group J. Jelovich, Manager A. K. Deb, Engineer C. Geiss, Engineer 135
ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION PITTSBURGH, PENNSYLVANIA REPORT INSPECTION NO.
99901043/85-01 RESULTS:
PAGE 10 of 10 Assembly & Test Facility (Seco Road)
- J. Hill, Manager R. Folino, Quality Assurance
- G. Krauter, Manufacturing Engineer
- P. Kushner, Quality Control Inspector
- A. Macrae, Manufacturing Engineer Quality Assurance Department J. Evans, Manager, QA NSID RCS Group F. B. Hyland, Manager B. F. Barnett, Quality Engineer T. Whalen, Quality Engineer Standards and Audits S. Cunningham, Manager Nuclear Products Division P. McManus, Manager Audits & Systems M. A. Marylyn, Manager S. Long, Senior Quality Assurance Engineer Nuclear Safety Department E. Figenbaum, Engineer
- Denotes those persons who did not attend the exit interview on January 16, 1986 at the conclusion of the inspection.
136
INDEX FACILITY REPORT NUMBER PAGE Anchor Darling Valve Company Williamsport, Pennsylvania 99900053/86-01 1
Cooper Energy Services Mount Vernon, Ohio 99900373/85-01 15 Corporate Consulting & Development Company, LTD Research Triangle Park, North Carolina
,99900511/85-02 21 Corporate Consulting & Development Company, LTD Research Triangle Park, North Carolina 99900511/86-01 27 The Foxboro Company Foxboro, Massachusetts 99900225/85-01 33 General Electric Company Nuclear Field Services Dcpartment King of Prussia, Pennsylvania 99901001/86-01 43 Georgia Institute of Technology Atlanta, Georgia 99900903/85-01 49 Joseph Oat Corporation Camden, New Jersey 99900251/86-01 53 Limitorque Corporation Lynchburg, Virginia 99900100/85-01 65 Minnesota Mining and Manufacturing Company Saint Paul, Minnesota 99901038/85-01 87 Pacific Valves Long Beach, California 99900075/86-01 91 The Rockbestos Company New Haven, Connecticut 99900277/86-01 103 Telemecanique, Inc.
Westminister, Maryland 99901011/85-01 109 3
Texas Instruments, Inc.
Johnson City, Tennessee 99901028/85-01 115 137
INDEX (continued)
FACILITY REPORT NUMBER PAGE Westinghouse Electric Corporation Nuclear Fuel Division Pittsburgh, Pennsylvania 99900005/85-01 119 Westinghouse Electric Corporation Nuclear Services Integration Division g
Pittsburgh, Pennsylyanta 99901043/85-01 127 m
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This periodici21 covers the results of spections performed by the NRC's Vendor Program Branch that have been i s ributed to the inspected organi-zations during the period from Janua 1 6 through March 1986. Also included in this issue are the resul s of ertain inspections performed prior to January 1986 that were not include in previous issues of NUREG-0040.
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