ML20236K562
| ML20236K562 | |
| Person / Time | |
|---|---|
| Issue date: | 10/31/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0040, NUREG-0040-V11-N03, NUREG-40, NUREG-40-V11-N3, NUDOCS 8711090267 | |
| Download: ML20236K562 (166) | |
Text
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NUREG-0040 Vol.11, No. 3 LICENSEE CONTRACTOR AND VENDOR INSPECTION STATUS REPORT l
QUARTERLY REPORT JULY 1937 - SEPTEMBER 1987 UNITED STATES NUCLEAR REGULATORY COMMISSION
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f Available from Superintendent of Documents U.S. Government Printing Office Post Office Box 37082 Washington, D.C. 20013-7082 A year's subscription consists of 4 issues for this publication, j
i Single copies of this publication l
are available from National Technical l
Information Service, Springfield, VA 22161 l
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NUREG-0040 Vol.11, No. 3 LICENSEE CONTRACTOR AND. VENDOR INSPECTION STATUS REPORT QUARTERLY REPORT JULY 1987 SEPTEMBER 1987 l
i oIe"r*uS!ish dr 987 Division of Reactor inspection and Safeguards Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 p..,,
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CONTENTS Page'
- 11. Preface...............................................................-
111 2. Repor t ing. Forne t....................................................
'y
~3.
Inspectors Reports....................................................
1 4.~ Selected Information'Notic'es..........................................
127
'i 5'.Index.................................................................
163'
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6'. Table of Vendor Inspect! ion Reports
. Related to Reactor' Plants...........................................
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i PREFACE A fundamental premise of the Nuclear Regulatory Commission's (NRC) nuclear facility licensing and inspection pro "am is that licensees are responsible s
for the proper construction and safe opiration of their nuclear power plants.
The total government-industry system for the inspection of nuclear facilities has been designed to provide for multiple levels of inspection and verification.
Licensees, contractors, and vendors each participate in a quality verification process in accordance with requirements prescribed by, or consistent with, NRC rules and regulations. The NRC inspects to detennine 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 quality verification programs.
In implementing this multilayered approach, a licensee is responsible for developing a detailed quality assurance (QA) plan.
This plan includes the QA 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 case 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. Once accepted by NRC, 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 controls may be applied by the NRC to previously accepted QA programs.
When design and construction activities were high, finns 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 I
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 addition to the above, the Vendor Program Branch has begun inspections at licensee facilities covering the areas of procurement of replacement parts for use in safety-related systems and licensee / vendor interface programs as requested in Generic Letter 83-28. This edition of the White Book contains copies of the inspection reports of inspections completed to date.
Subsequent issues will contain those reports that are issued in the quarterly report period covered by that White Book.
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 pregrams. 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 nu 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 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 require special attention. The White Book also con-tains copies of I&E Information Notices, concerning vendor issues released during the calendar quarter.
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.
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ORGANIZATION:' COMPANY, DIVISION I
CITY, STATE-LREPORT.
INSPECTION.
INSPECTION NO.: Docket / Year / Sequence DATE:
ON-SITE HOURS:
CORRESPONDENCE ADDRESS:
Corporate Name Division
-ATTN:
Name/T1tle Address-City, State Zip Code 1
' ORGANIZATIONAL CONTACT: Name/ Title
' TELEPHONE NUMBER:-
Telephone Number NUCLEAR INDUSTRY _ ACTIVITY:
Description of type of components, equipment, or services supplied.
y ASSIGNED INSPECTOR:
harre/ Vendor Prograin Branch Section Date OTHERINSPECTOR(S):
Name/ Vendor Program Branch Section l:
1 APPROVED BY:
-Name/ Chief - Section/ Vendor Program Branch Date w
INSPECTION B*iSES 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 I
facilities for which equipment, services, or records were examined during the inspection.
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ORGANIZATION:
ORGANIZATION CITY, STATE REPORT INSPECTION N,0. :
RESULTS:
PAGE 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 applicable to the organization being inspected.
B.
NONCONFORMANCES:
Shown here are any inspection results determined to be in Tonconformance 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 wnich 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 nonconfomance 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 item.
If this section is omitted, all previous inspection findings have been closed.
E.
INSPECTION FINDINGS AND OTHER COMMENTS:
This sectic,n 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 spcial 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)
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.1 INSPEC'IORS REPORTS 1
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ORGANIZATION: ALLIED NUT & BOLT COMPANY, INCORPORATED KIllG OF PRUSSIA, PENNSYLVANIA REPORT INSPECTION INSPECTION NO.: 99901093/87-01 DATES: 06/23-26/87 OH-SITE Moline an CORRESPONDENCE ADDRESS: Allied Nut & Bolt Company, Incorporated 520 Hertzog Boulevard Post Office Box 670 King of Prussia, Pennsylvania 19406 i
ORGANIZATI0tiAL CONTACT: Gerald Korin, Quality Assurance llanager TELEPHONE NUMBEPs:
215-275-2200 NUCLEAR INDUSTRY ACTIVITY: Naterial supplier - fastener components.
i-f[t 7 ASSIGiiED INSPECTOR:
W--
- d. C. Harper, Program Development and Reactive D&te Inspection Section (PDRIS) 01HER IllSPECTOR(S):
A. J. Pjura, Brookhaven National Laboratory APPROVED BY:
h JM 7 /o g 7
_d. C.(/5 tone, Chief, PDR15, Vendor Inspection Branch a
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INSPECTI0h BASLS AND SCOPE:
A.
BASCS:
10 CFR 50, Appendix B and 10 CFR Part 21.
i B.
SCOPE:
The inspection at Allied Nut & Boit was conducted in order to review compliance with and the implementation of 10 CFR 50, Appendix B and 10 CFR Part 21.
PLAliT SITE APPLICABILITY:
50-354 (Hope Creek); 50-272 and 50-311 (Salem Units 1 and 2).
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ORGANIZATION: ALLIED NUT & BOLT COMPANY,. INCORPORATED KING OF PRUSSIA, PENNSYLVANIA 1
REPORT,
INSPECTION 1
. Mn. QQQn1nQ1/A7-01 RFSill TS :
PAGF 7 nf A A.
INSPECTION ISSUES:
1
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The inspection at Allied Nut & Bolt (AhB) was performed as a part of an j
ongoing VIB fastener evaluation program. The inspection was conducted to review ANB's compliance with and implementation of 10 CFR 50 Appendix B 1
and 10 CFR Part 21.
I B.
INSPECTION FINDINGS:
1.
. Violations a.
Contrary to Section 21.6 of 10 CFR Part 21,~ANB failed to post copies of Section 206 of the Energy Reorganization Act of 1974.
(87-01-01) b.
Contrary to Section 21.31 of 10 CFR Part 21, ANB failed to pass the requirements of 10 CFR Part 21 on to certain sub-tier vendors that supplied material for nuclear customer use.
(87-01-02) i i
2.
Nonconformances
)
a".
Contrary to Criterion V of Appendix B to 10 CFR Part 50, and to the ANB QA hanual, QAP 2.0, paragraph 3.2, there was no objective evidence that ANB had procedures for training and qualifying personnel responsible for quality control and/or assurance activities. -(87-01-03) i b.
Contrary to Criterion VII of Appendix B to 10 CFR Part 50 and to the ANB QA Manual, QAP 7.0, Paragraph 2.3, ANB failed to perform a source evaluation of Pottstown Plating.
(87-01-04) c.
Contrary to Criterion XV of Appendix B to 10 CFR Part 50 and to the ANB QA Manual,-QAP 15, Paragraph 3.4, ANB failed to note a final disposition on numerous nonconformance reports..
(87-01-05)
C.
UNRESOLVED ITEMS:
None.
D.
OTHER FINDINGS AND COMMENTS:
Allied Nut & Bolt (ANB) is a relatively small distributor of both ferrous and nonferrous fasteners. ANB does not manufacture fasteners or perform any special processes inhouse.
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ORGANIZATION: ~ ALLIED NUT &' BOLT COMPANY, INCORPORATED i
KlHG OF-PRUSSIA, PENNSYLVANIA REPORT.
INSPECTION hn a
' QQ Q ninQ'UR7-n 1 RFSin TS:
PAGF 1 of A The inspection team evaluated the areas of Part 21 compliance, audits performed by ANB (internal and external), ANB nonconforming material' dispositions, calibration of measuring instruments, and training and indoctrination of personnel involved with quality. A total of 13 nuclear occument packages.were evaluated by the inspection team for compliance with customer purchase order requirements and applicable regulations (Part 21; 10 CFR S0, Appendix B).
1.
Part'21 Issues i
The requirements of Part 21 were not. imposed on certain of ANB's
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sub-tier vendors. The original P0's issued to ANB by the nuclear i
licensee imposed the requirements of Part 21 on ANB.
(See violation l
87-01-02.)
l Original PO#
Originating Sub-tier ANB PO#
Date Company Vendor Date P2-17927 10/30/86 Public Service Bristol Heat Q-8308 l
Electric & Gas Treaters 01/12/87 i
P2-17927 10/30/86 Public Service Lab Testing Q-8202 J
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Electric & Gas 12/23/86 l
P2-17927 10/30/86 Public Service Lab lesting Q-8375 Electric & Gas 01/19/87 i
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P2-179E7 10/30/86 Public Service Pcttstown Plating Q-8389 l
Electric & Gas 01/21/87
(
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P2-17927 10/30/06 Public Service Pottstown Phting Q-8249 l
Electric & Gas 01/05/87 l
P1-147349 05/19/86 Public Service Carpenter 1798 I
Electric & Gas Technology Corp.
01/21/85 I
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q P0 P2-17927 from Public Service Electric & Gas (PSEG) was issued to
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ANB for approximately 1000 SA193B7 and SA194 Gr 2H fasteners.
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Bristol Heat Treaters (BHT) performed tSe heat treating for the PSEG j
order. Lab Testing, Inc. supplied mecnanical and chemical testing on Lab Report No. C-3538 for the PSEG order. At the time of the inspection both BHT and Lab Testing were on the ANB Approved Vendors i
List'(AVL). Pottstown Plating supplied clad plating for 109 pieces
-1/2 - 13 x 1 1/2 SA193 Gr 7 bolts, 105 pieces 3/8 - 16 x 1 SA193 l
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0 ORGANIZATION: ALLIED HUT & BOLT COMPANY, INCORPORATED KING 0F PRUSSIA, PENNSYLVANIA-
' REPORT' INSPECTION
___ MA.*
QQQO10Q UR7-01 RESul_TS:
.PAGF 4 M R
'Gr 7 Hex Bolts, 105 pieces 1/2 - 13 x 2 1/2 SA193 Gr 7 Hex Bolts, 550 pieces 1/2 - 13 SA194 Gr 2H Hvy Hex Nuts, 105 pieces 1/2 - 13 x 1 SA193 Gr 7 Hex Bolts, 105 pieces 1/2 - 13 x 1 1/2 SA193 Gr 8 Hex Bolts. Pottstown Plating was not on the ANB AVL at the time the services were rendered for PO Q-8249. (See nonconformance 87-01-04.)
P0 P1-147349 from PSEG was issued to ANB for'a total of 30 A193 Gr 8 Bolts (30 PCS), and ASTM A194 Gr 8 nuts (20 PCS). Carpenter Technology Corporation supplied ASTM A193 Gr 8 bar to ANB. At the time of the inspection Carpenter was on the ANB AVL.
The following list is a' summary of subcontractors that ANB employea i
to render services _for parts supplied for nuclear safety related j
service. ANB did not impose the requirements of.Part 21 on these subcontractors.
F0#/Date Company Product PO Q6584 05/29/86 Lab Testing 1 - 8 Hvy Hex Nuts SA194 Gr 8 PO Q6570 05/26/86 Lab Testing 1" Test PCS 304SE P0 Q6582 05/29/86 East Coast Metals 3 PCS 1 x 128 304SS
.P0 Q6582 05/29/86-East Coast Metals 50 PCS 1 - 8 x 4" SA193 Gr 8 P0 Q6582 05/29/86 East Coast Metals 10 PCS 1 - 8 x 6 1/4 SA193 Gr B8
.P0 Q7831 11/06/86 Lab Testing 5/8 - 11 Hvy Hex Nuts.SA194 Gr B7' P0 Q8021.12/02/86 Lab Testing 3/4 - 10 Hvy Hex Nuts SA194 Gr b P0 09157 04/22/87 Gage Lab Corp.
Thread plug gage calibration P0 Q8402 02/06/87 Gage Lab Corp.
Thread plug gage calibration Gage Lab Corporation is ANB's sole supplier of calibration services I
for micrometers, verniers, ring gages and plug thread gages.
l Calibration The calibration of micrometers, vernier gages plug and ring thread gages were checked. A detail evaluation was made on the technical basis for calibration of 72 plug and ring thread gages, three 0-1" l
micrometers, two 1-1" uierometers, one 2-3" micrometers and two I
vernier gages. At the time of the inspection, all measuring equipment were properly calibrated and up to date in calibrations.
The technical basis for calibration of the measuring equipment was determined adequate.
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?., ORGANIZATION: ALLIED NUT.& BOLT COMPANY,' INCORPORATED KING OF PRUSSIA, PENNSYLVANIA
-REPORT.
. INSPECTION j
~ un. Cocnino?/R7-n1 RESULTS:
PAGE 5 of 8 j
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3.
Nonconforming Material l
Nonconforming Material Report (NCR)
' Criterion XV of 10 CFR 50, Appendix B, requires review and disposi-tion of. nonconforming items in order to prevent their. inadvertent i
'use or installation..During a review of some 43 NCR's for the years 1984-86, numerous NCRs were found to have no disposition, but' had a l
written. explanation of action.under.the remarks section (see nonconformance 87-01-05).
NCRs 151, 160, 161, and 166 were written because'of erroneous or missing certifications. A review of the purchase order files for each indicated that the proper certifica-tions'were supplied to ANB prior to shipment.
NCR #157 dated March 26, 1984. described a problem with Cardinal'
' Industrial Products Company (CIPC) 1/2 - 13 x 2.SA193 Gr B7 Heavy Hex Bolts purchased by P0-Q1645.(125 PCS). The description.of the
. nonconformance by the ANB QC Supervisor was " Bolt marked with B7-L 7 Mfg / trace C/U reviewed spec SA-193 & SA-320 found.nothing about l
acceptability of grade markin,gs." A final disposition of accept as l
is, rework, downgrade, scrap'or return was never determined by a check mark to complete the form (see nonconformance 87-01-05).
L LIPC sent ANB a letter on March 1984 which describes the bolt head markings.
In summary, CIPC explained tnat the dual head markings were their attempts to develop a program of product simplification.
This program involves parts that are run for one material specifi-cation but they simultaneously meet all the requirements of a second material specification.
Therefore, in keeping with the marking requirements of the bolting specifications, CIPC included grade markings for each specification on the bolt head. CIPC claimed they were advised that there is nothing which prohibits multiple marking j
of bolts in caces where bolting. products are manufactured to fully I
comply with more than one specification.
CIPC concluded that their certification for the bolts purchased on ANB P0 BQ1645 were in full compliance with SA193 Grade B7 since they had a mark of C for Cardinal, U as a heat code, B7 for the material specification and L7 for SA320 Grade L7 which the product fully complies with. ANB accepted CIPC explanation and closed out the nonconformance with QC supervisor's signature approval, i
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ORGANIZATION: ALLIED NUT & BOLT COMPANY, INCORPORATED KING OF PRUSSIA, PENNSYLVANIA REPORT INSPECTION un ?
QQQn1001/n7-01 RESULTS:
PAGF 6 of B l
4.
Corrective Action Request (CAR)
As part of the ANB annual review of NCR's, they evaluate previous years' audit. findings for trends that may be adverse to quality.
If adverse trends are found, the QC' supervisor is required to issue a CAR. From review of several CARS, it was noticea that CIPC was the only company reported to have excessive nonconformances by ANB's QC supervisor (CAR dated September 1, 1983).
In 1983, the QC super-visor issued a CAR regarding adverse trends with CIPC material to the ANB President, specifically, P0-Q 1098 for 2500 PCS 1/2 x 2 1/2 SA193 Gr B7 studs. The CAR description of the discrepant condition reads as follows, "The' material was purchased from Sumitomo on Cardinal P0 13308. Sumitomo bought the steel bars from Diado and they were heat treated by Hayasi Sieko.
Hayasi Sieko's report indicates tempering performed at 1080*-ll50*F. Cardinal's
.i report indicates tempering temperature to be 1150 F.
ASTM requires 1
1100 minimum for Gr B7 material ordered by Allied." As corrective J
action ANB inquired about and received a telex confirming actual tempering temperatures from Sumitomo. The tempering temperatures were confirmed to be in the acceptable range. Sumitomo was instructed in the future to submit actual tempering temperatures versus a temperature range to ANB.
5.
Audits a.
Internal S
QAP 18, Revision 1, requires ANB to perform internal audits on a quarterly basis to verify conformance to the provisions of the QA manual.
These quarterly audits evaluate each QAP program procedure over the course of a twelve month period.
The 1983, 86, and 87 quarterly audit packages were reviewed for completeness and validity. Ten quarterly audit packages were reviewed.
In all cases the ANB audit checklist was complete and approved by the ANB President. Also, in all cases, the formal report was submitted to the President within 10 days.
b.
External The ANB approved vendor's list (AVL) was reviewed. From cross-referencing suppliers identitied on purchase orders with the AVL, i
it was apparent that Pottstown Plating was not on the ANB AVL.
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0 ORGANIZATION:
ALLIED NUT & BOLT COMPANY, INCORPORATED KING OF PRUSSIA, PENNSYLVANIA REPORT INSPECTION nn.?
444nin41/R7-n1 RESULTS:
PAGF 7 of 8 (See nonconformance 87-01-04.) Pottstown Plating provided plating services for ANB under P0 Q8389 dated January 21, 1987, l
and P0 Q8249 dated January 5, 1987.
l P0 Q8389 required clad plating of 109 PCS of 1/2 - 13 x 1 1/2, SA193 l
Gr 87 and P0 Q8249 required plating of 970 PCS of SA193 Gr B7, SA194 l
Gr 2ii bolts and nuts.
In both cases, the plated nuts and bolts were l
to be supplied to Public Service Electric & Gas per their P0 P2-179275 j
dated October 30, 1986. Further, QAP 9.0, Revision 1, paragraph 2.1 i
states:
"When standard products are modified to meet special customer l
requirements and the work is subcontracted, the subcontractor is l
required to have a quality program, approved by Aliied, which I
provides for control of manufacturing processes." AhB indicated that Pottstown Plating facility had a satisfactory historical performance, however, a formal audit had never been performed.
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6.
Training and Indoctrination j
No procedure or instructions were in place that specifically addressed how Allied's corporate training requirements were to be implemented j
with respect to the requirements found in ANSI N45.2.6.
- Further, l
ANb certification process only addresses the qualifications of j
personnel performing inspection activities, and not those of auditors, j
(See nonconformance 87-01-03.)
Personnel folders of the General Nanager, Quality Assurance Manager,
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Quality Control Supervisor, and individuals assigned as inspectors l
l were examined with respect to their individual training.
It was determined that each individual was, in fact, qualified to perform i
their assigned quality related activity. For example, individuals j
passed inspection courses sponsored by the American Society for l
Quality Control, and have industrial backgrounds sufficient to meet l
the Level requirements of ANSI M45.2.6. Further, the QA Manager and
{
QC Supervisor both have met the auditing qualifications of ANSI i
h45.2.23, yet lillied's QA program does not commit to this standard
{
or any alternative.
]
Personnel certification forms, as described in ANB QAP 2.0 were found I
in each individual's folder attesting to his familiarity with the l
l requirements of the QA manual, yet there was no objective written evidence that the individuals had been indoctrinated on the QA manual l
and its latest revisions.
Interviews with the QC supervisor indicated that he was aware of the requirements of the QA manual's latest revision.
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ORGANIZATION: ALLIED NOT &' BOLT' COMPANY, INCORPORATED KING OF PRUSSIA PENNSYLVANIA
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' REPORT INSPECTION
'Nn -
QQQninon /A7-01 RESULTS:
PAGE 8 of C 7.
Plant Tour
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ANB is a distributor of fasteners, therefore, manufacturing and performance of special processe:, do not take place at the warehouse.
ANB has a segregated,. fenced in, locked and identified area clearly designated for ASME material.
The material in'the. area was tagged with a white tags indicating AND QC approved material.
The tags indicate PO number, item number, inspector and date. Six tagged items were compared with their status on file. The following information from the status tags were cross-referenced with the applicable P0s. No discrepancies were found.
Cert #/Date P0d/Date Heat / Description Trace On File QA Approval Vendor
-Q-2159 OlP453 T70 184281 Yes.
Texas 03/28/85 1-1/8-7 HHN Bolt SA194 G2A Q-2159 D36361 ADG 02/22/8b Yes A&G 03/28/85 1-1/4-8 HHN Engineering Q-2544 D48923 ANC 05/17/85 Yes A&G 05/23/85 3/4 HHN Engineering Q-4802 B18369 DH4 0047018 Yes Cardinal 12/10/85 1/8 HHN SA194 Gr 7 Q-5113 C6X267 HDD 06/01/87 Yes A&G 05/13/87 1-1/4-8 HHH Engineering Gr 7 Q-9402 C49284 AMA 06/11/87 Yes A8G 06/18/87 1/2-13 HA Engineering E.
_ PERSONS CONTACTED *:
Jerome Rosenstock John Jan Francisco, Jr.
S. Deleo G. S. Korin Al Pjura
- Present at the exit meeting.
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i ORGANIZATION: COLT INDUSTRIES.
J BELOIT, WISCONSIN REPORT-INSPECTION INSPECTION NO.:
99900300/86-01 DATES: 07/27-31/87 ON-SITE HolmS t 11 CORRESPONDENCE ADDRESS:
Colt Industries Fairbanks Morse Engine _ Dhision ATTN: Mr. Brian M. Saylor, Vice President Quality Assurance 701 L..
an Avenue Beloit, Wisconsin 53511 ORGANIZATIONAL CONTACT: Micheal. S. Horinka, Manager, Quality Assurance TELEPHONE NUMBER:
(6081 364-8314 NUCLEAR INDUSTRY ACTIVITY: Manufacture diesel engines and accessories and parts for standby diesel generator sets.
ASSIGNED INSPECTOR:
M P
9/aC/p/
P. J. Pre' scott, Program Development and Reactive Date Inspection Section (PORIS)
OTHERINSPECTOR(S):
I-25-f[
)
APPROVED BY:
M t
Date I
y,y. Stone, Chief, UR;5, Vendor Inspection Branch L)
J INSPECTION BASES AND SCOPE:
)
A.
BASES:
10 CFR Part 50 Appendix B and 10 CFR Part 21.
B.
SCOPE: This inspection was performed as a result of a Part 21 report L
issued by Colt Industries (CI) concerning a potential overspeed problem on the Colt PC-2 series engine, and the main bearing failures at Fermi 2 in late 1985 and early 1986.
PLfNF SITE APPLICABILITY:
Seabrook Station (50-443), Fermi 2 (50-341), Beaver
' Valley (2) (50-334). Hopt Creek (50-354), Shoreham (50-322), Farley 1 & 2 (50-348),(50-364).
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ORGANIZATION: COLT INDUSTRIES BELOIT, WISCONSIN REPORT INSPECTION NO.-
44400300/R6-01 RESULTS: 7/27-31/87 PAnF 7 nf 7 A.
VIOLATIONS:
None.
B.
NONCONFORMANCES:
None.
C.
UNRESOLVED ITEMS:
l None.
D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
1.
(0 pen) Unresolved Item (84-01-01)
Nonconformance Disposition - Nonconforming items are identified on various forms and subjected to a review process.
If the item is to be accepted as is, the variation must be approved by the Material ReviewCommittee(MRC).
According to Section 15.4.3.2 of the Qw tity Assurance Manual (QAM),
a.
items submitted to the MRC for acceptance as variation material must be such that (among other things) CI has design authority for the item (s)beingconsidered. The inspector reviewed selected variations approved by the MRC and noted several for components where Societe d' Etudes de Machines Thermiques (SEMT), the main designer of the Pielstick engine, had design authority.
Discussions with the Vice President of Engineering of CI indicated that this practice had been reviewed and concurred in by SEMT for 1
minor design variations. Further, the inspector reviewed a draft i
revision of the QAM which would address the practice.
i Status During this inspection, the NRC inspector requested a copy of the QAM to determined if the draft revision had been adequately 1
incorporated into the present QAM.
Discussions with the Manager of 1
i Quality Assurance indicated that due to personnel changes within l-CI and changes of job functions, the draft revision addressing the design authority practice had not yet been incorporated into the present QAM. The Manager of Quality Assurance did state however, that the QA department would revise the QAM prior to the next NRC inspection.
10
ORGANIZATION:
C0LT INDUSTRIES BELOIT, WISCONSIN l
REPORT INSPECTION
_NO - 44400100/AFi 01 RFSlHTS-7 /'27-31/R7 PAnF 1 nf 7 2.
(Closed) Unresolved Item (84-01-02)
OJerating and Maintenance Manual - lhe inspector requested the operating and Maintenance Manual (Manual) for the Shoreham diesel generators for review. CI stated that the Manual was not yet finalized, but that the Hope Creek diesel generators were essentially identical. The inspector reviewed portions of the Hope Creek Manual and noted two discrepancies with the normal operating parameter limits in paragraph H on page 4-11.
Engineering drawing No. P12609255, " Pressure & Temperatures PC2 and PC2 3 Diesel Engines," Revision 6, gives the operating parameter lindts to which CI tests the engines. The limits in the manual did not in all cases agree with those on the engineering drawing; e.g.,
the fuel header pressure band was 22-40 psi versus 25-35 psi, the lube oil header pressure band was 75-90 psi versus90-110 and the jacket water temperature band from the engine was 169 -180 versus specify at what load (s)pector, also noted that the manual did not 173*-183. The NRC ins those parameter limit (s) applied.
The NRC inspector reviewed all associated documentation and inter-viewed CI personnel from the engineering department and utility department personnel responsible for the manual. The CI personnel j
interviewed indicated that all parameters listed were with the diesel engines operating at 100 percent load. CI personnel further stated that for conditions other than 100% load, the engine parameters should be between the 100% load value and the alarm set-points, also available in the manual.
3.
(Closed) Unresolved Item 84-01-03 Load Test - The inspector reviewed the Shoreham diesel generator procedure, " Load Tests," dated March 8, 1984, and noted that it tested the diesel generator at a design load of 100 percent or 4430 KW and a power factor of 1.0.
The Shoreham specification for the diesel generator imposes IEEE Standard 387-1977, "IEEE Standard Criteria for Diesel Generator Units Applied as Standby Power Supplies for Nuclear Power Generating Station." Paragraph 6.3.1. of IEEE-387 l
requires load tests to be performed at the engine manufacturer's factory at rated power factor. The Shoreham specification defines rated power factor as 0.8.
Testin larger load on both the generator 'g at a power factor of 0.8. puts a and the engine of the diesel 1
generator set. Ci stated that they do not have the capability to test at a power factor other than 1.0 and that they had verbally inforned Shoercham of this fact.
11
ORGANIZATION: COLT INDUSTRIES BELOIT,, WISCONSIN
+
I REPORT INSPECTION J0. - 99900100/R6-01 RFSill TO 7/97 11/R7 PW a nf 7 The NRC inspector reviewed Colt Industries Test Report No. 901, and associated test drawings No. 10-560-395, SHT. A, note 3 which stated in part that, this test shall be run on one unit only as a qualific-ation to demonstrate the capability of carrying the following loads without exceeding the system design limits.
[All loads to be at unity power factor.] CI personnel, also presented the NRC inspector with the cover sheet for the CI iest report 901 entitled " Stone &
Webster Engineering Corporation Review of Supplier's Technical Document" which was cpproved by a Stone & Webster representative April 30, 1985. Based on this, the test results have been reviewed and accepted by the licensee or his agent.
E.
_0THER FINDINGS AND COMMENTS:
1.
Colt PC-2 Series Emergency Diesel Generator The NRC inspecter reviewed Colt Industries deportability evaluation and corrective action related to a 10 CFR Part 21 report dated February 4, 1986, wnich identified a potential overspeed trip problem as::ociated with the Ccit PC-2 series emergency diesel generators. The Part 21 report was initiated by Colt after Public Service of New Hampstire (PSNH) (the licensee for Seabrook Station) reported that one of its diesel engines had tripped out on ovErspeed while it was being started for test purposes. The overspeed trip seemed to involve the rack boost cylinder on the control system.
To prevent further o(.currences, Colt Industries recommended that all Colt PC-2 series engines in nuclear services be modified to positively vent the air from the rack boost cylinder, and its associated shuttle valves, by making the source of that air the pilot air lines that control the main air start admission valves.
The pilot air supplies are controlled by scienoid valves that receive the start t gnal from the control system. These are three-way valves that positively vent the pilot air rignal to the atmosphere when the control system senses that the engint is started.
While reviewing CI's Part 21 Report File the hRC inspector noted correspondence that identified the same failure at the Farley Nuclear Plant (FNP) in September 1977.
The NRC inspector asked l
CI personnel if the failure had been evaluated at that time to deter-l mine if this could be a generic problem.
CI personnel presented the NRC inspectors with an inter office memo discussing the problem as dirty air (water & rust) causing the air start valve to occasionally stick open af ter a start and the air start solenoid hao de-energized.
L Ll' l
12
l ORGANIZATION: COLT INDUSTRIES l
BELOIT, WISCONSIN REPORT INSPECTION NO..
QQ400 ~400 /R A-Q 1 RFSt!! TS -
7/27-31/R7 PACF E nf 7 i
This held the fuel rack boost cylinder in the full open position l
causing the EDG to overspeed. CI recommended that the best solution l
would be to clean up the air, but the customer (FNP) chose to make a field modification to vent the air pressure rather that cleaning the air. due to cost considerations. The memo further stated that an Alabama Power representative from the Birmingham office would write i
a failure report on the event.
No further problems were noted
)
until PSNH reported the problem.
1 i
2.
Fermi 2 Emergency Diesel Generator No. 13 Thrust Bearing
)
Fermi-2 has experienced a series of emergency diesel generator (EDG) thrust bearing problems.
They occurred in January, November and December, 1985. Specifically, the EDG's exhibited an abnormal number of distressed or failed bearings that resulted in extensive engine damage.
In June 1986, Failure Analysis Associates (FAA) issued a report entitled " Investigation of Surface Scoring on Main Bearings: Fairbanks Morse 38 TD 8-1/8 Diesels at Fermi 2 Power l
Plant." The report discussed several problems with the No. 13 thrust bearing and some product improvement design changes were suggested. These included the elimination of the circumferential oil groove in the thrust bearing only, application of an electro-plated babbitt overlay to the bearing inner diameter, polishing of crankshaft journals to a 10 micro-inch RMS finish, and redesign of the thrust bearing to decouple the thrust faces from the radial section.
In a December 1986 meeting with Detroit Edison Co., FAA, a CI representative, and the NRC, the thrust bearing problems were discussed and it was felt that a possible design deficiency did exist with the No. 13 thrust bearing. As a result, the CI represen-tative stated that they would study possible design changes, which if successful would be incorporated in future thrust bearings as e ptoduct improvement. There were no discussions however, as to when the redesigned thrust bearing would become available.
As part of this inspection the NRC inspector interviewed CI personnel and reviewed the proposed design changes CI is making to the No. 13 thrust bearing as a product improvement. During the interview CI personnel presented the NRC inspector with three (3) engineering and experimental design change request forms; No. M202, dated November 12, 1986 No. M204, dated June 25, 1987 No. M209, dated January 26, 1987, and all associated design drawings.
The experienced design changes included several of the recommendations suggested in i
i 13 3
ORGANIZATION: COLT INDUSTRIES t
BELOIT, WISCONSIN REPORT INSPECTION Nn. 000nninn/RA 01 RF9 fit TS-7/97.11/n7 DarP A nf 7 the June 1986 FAA report such as; elimination of the oil groove, lead babbitt overlay and special machining processes to improve the oil film surfaces.
CI personnel stated however, that they had not yet approved any of the design changes and were not certain as to when the redesigned thrust bearing would become available.
CI personnel also stated that because the time frame was uncertain, that they would consider issuing a Service Information Letter (SIL) to their customers discussing the bearing problems, and specific preventive measures that can be taken with the present bearings until a satisfactory redesign can be approved.
3.
Lube Oil Strainers Some concerns has arised regarding the use of a single (simplex) lube oil (L.0.) strainer in the emergency diesel generator lube oil system.
The NRC inspector discussed this issue with CI personnel and requested the following information:
1.
What type of testing is performed on the L.0. strainers at the factory?
2.
Did C1 perform an analysis to determine how long an engine would run if the L.0. is contaminated?
3.
What did the customers specify as to the type of L.0. strainer in their purchase specification?
CI's response was the following:
1.
The engines are factory tested with only recommended L.0. end utilizing L.0. strainer configuration as specified by the customer.
2.
No analysis has been performed on engine run time as it applies to L.0 contamination.
3.
CI personnel presented the NRC inspector with (3) approved purchase specifications; Hope Creek, Shoreham and Beaver Valley.
Hope Creek's purchase specification required a full flow 40 micron simplex strainer.
The other two utilities required a multiple element, continuous flow strainer, which CI stated cannot be interpreted as a duplex design.
The NRC inspector L
14
ORGANIZATION:
COLT INDUSTRIES BELOIT, WISCONSIN I
REPORT INSPECTION.
NA =
QQQAn100/RA nl DFRnlTR.
7/97 11/R7 DanF 7 ef 7 also reviewed the L.0 piping isometrics at CI for the subject utilities and all piping diagrams reflected a single strainer arrangement. 'CI also stated that no specific problems have been brought to their attention regarding' the single strainer I
arrangement by their nuclear customers. Beaver Valley did,
]
however, report a high differential pressure across their L.0.
I strainers during testing'in May 1987 but CI attributed the i
problem to improper cleaning of the strainer bowl by the 1
utility.
,{
4.-
Design Control The Vendor's Quality Assurance Manual was reviewed with regard to-the specific requirements for control of design activities. The i
project engineering guidelines that further expand.on these require,
ments were also reviewed. These guidelines include design input j
and design review checklist, and procedures for the issuance of' orders and release.
To assess conformance to these requirements, CI's design control procedures were reviewed.
It was determined that in all instances, verification of the design documents had been performed as indicated by the proper sign-off.
(Verifications performed primarily by the " design review" method.) All nuclear I
design activities are conducted by the Supervisor of Mechanical Systems Engineering, with verification performed by his supervisor.
Drawings are prepared by draftsmen and reviewed by supervision.
5.
Plant Tour - The inspectors toured the CI manufacturing, testing and training facilities during the inspection in the company of CI officials.
Items witnessed included component machining, parts assembly, pipe bending, welding qualification.
l Wo nonconformances were identified.
F.
PERSONS CONTACTED:
P. J. Prescott, USNRC 1
P. McAlpine, Mgr. Parts Marketing, FM M. Horinka, Mgr. QA Nuclear & Military M. Armfield, QA Engineer P. Danylur, VP Engr.
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ORGANIZATION: COOPER ENERGY. SERVICES GROVE CITY, PENNSYLVANIA l
l
, REPORT.
' INSPECTION.
INSPECTION NO.: 99900317/87-01 DATE: 03/30-04/3/R7 ON STTF N01199-Al CORRESPONDENCE ADDRESS:
Cooper Industries I
1 Cooper Energy Services ATTN: Mr. F. Bruce Stolba, Vice President and General Manager
'150 Lincoln Avenue
- l Grove City, Pennsylvania 16127 ORGANIZATIONAL CONTACT:
W. H. Allen Lambert, Manager of QA TELEPHONE NUMBER:
(4121 458-8000 i
NUCLEAR INDUSTRY ACTIVITY: Original equipment manufacturer of standby diesel i
generators for nuclear service.
Current sales in parts, repair and service i
only. No current orders for standby diesel generators at commercial nuclear facilities.
I l
1 ASSIGNED INSPECTOR:
A.
,Y c/iy>/i7 P. J. Prescott, Program Development and Reactive bate Inspection Section (PDRis)
OTHER INSPECTOR (S):
E H. Trottier, NRR M. Schuster, Consultant, BNL l
APPROVED BY:
M W%Y 18' O es C. Stone, Chiet, IS, Vendor Inspection Branch Date y>
q I
INSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR 21 and 10 CFR 50, Appendix B.
SCOPE: This inspection was l
B.
significant 10 CFR 50.55(e) performed in response to two recent and reports involving Cooper standby diesel generators.
In' addition, the purpose of this inspection was to verify corrective and preventive actions taken in response to findings of the e
l previous hRC inspection (86-01, July 28-August 1, 1986).
[
PLANT SITE APPLICABILITY:
Byron 1 and 2 (50-454, 455); Braidwood 1 and 2 (50-l 456,457); Cooper (50-298); Nine Mile Point 2 (50-410); Palo Verde 1, ?, and 3 (50-528,529,530); South Texas 1 and 2 (50-498, 499).
i 17
1 1
ORGANIZATION: -COOPER ENERGY SERVICES GROVE CITY, PENNSYLVANIA REPORT INSPECTION Nn - 40nnn117/R7-01 RESULTS:
PAGE 2 of 12 A.
VIOLATIONS (1) Contrary to 10 CFR 21.31, Cooper Energy Services (CES) issued Purchase Orders (P0s) 532190275 dated January 1, 1983, 532190034, l
dated January 1, 1984, 3921G6372, dated June 23, 1986, and 3921G6941, dated October 17, 1986 to a subvendor (Vander Horst Corporation of Olean, New York) for the plating of critical parts i
without specifying that the provisions of 10 CFR 21 apply.
(Critical parts are defined as safety related.)
(87-01-01)
B.
HONCONFORMANCES (1) Contrary to CES Quality Control Procedures QCP-10-6, Section 4.1.2, and QCP-10-13, Section 3.2, a document that was of significant value 4
in maintaining, reworking, repairina, replacing, or modifying an item or that could have been of r ificant value in determining the cause of an accident or malfunctic of an item (Work Order Form #451) was not maintained until the item to which it pertained was decommissioned from service.
(87-01-02)
(2) Contrary to CES Quality Control Procedure QCP-10-15, Section 4.11, no calibration frequency requirements were provided for the hardness i
testing equipment in Inspection Procedure ISF-5-2.
(87-01-03)
(3) Contrary to CES Quality Control Procedure QCP-10-6, Section 4.5.2, and CES Material Specification SE-89-2N, no certificate of conformance was available for plating repairs performed by a subvendor, Vander Horst Corporation of Olean, hew York.
(87-01-04)
C.
UNRESOLVED ITEMS (1) CES stated that its design drawings would be revised to delete
" NOTE 7," which allows production personnel to waive stress relief requirements (without documentation) for connecting rods being used in safety-related applications.
(2) CES stated that it would provide an interim memorandum requiring that 10 CFR 21 be applied to all critical / safety-related purchase orders.
It also indicated that the appropriate procedures would be revised to include 10 CFR 21 requirements.
18
ORGANIZATION: COOPER ENERGY SERVICES GROVE. CITY, PENNSYLVANIA REPORT INSPECTION Nn - 000nn917/E7 01 RESULTSt PAGE 3 of 12 D.
. STATUS OF PREVIOUS INSPECTION FINDINGS (1)
Inspection No. 86-01 (July 18-August 1, 1986)
Nonconformance 86-01-01 (Closed):
Tne NRC inspectors found out-of-date procedures and an out-of-date procedures index in the heat treat area office. Specifically, Procedures HT-17N and HT-18AN were cited as being out of date according to the Standards Manual for Engineering Material Specifica-tions Index.
(These out-of-date heat treat procedures were the subject of Nonconformance B.1 identified during the previous inspection in March ~1984, Report No. 99900317/84-01.)
In closing this open item, the inspector reviewed the revision status of the Material Specifications Index (November 3, 1986) and compared the revision status of several procedures against the revision date tabulated in the index. The following presents a summary of this review:
Procedure No.
Rev. Date Index Rev. Date Status HT-3N 4/15/75 4/75 Current A-1AN 9/5/75 9/75 Current A-5BN 4/21/75 4/75 Current HT-17h 11/12/84 11/84 Current HT-18AN 2/21/85 2/85 Current
-As shown above, CES has taken effective corrective and preventive actions to control the revision status and use of metallurgical procedures.
(2) Inspection No. 99900317/84-01 (March 12-16, 1984)
Nonconformance 11 (Closed):
Technical evaluations to determine deportability as a 10 CFR 21 item (componcnt defect) were not being completed within the 30-day require-ment specified in CES Procedure QCP-10-14, " Quality Requirements for Reporting of Deficiencies."
The inspectors reviewed the latest revision to QCP-10-14 (Rev. 3 dated October 29,1986) and found that the section addressing evalua-tions of potential component deficiencies has been revised. The i
f Y.
19
ORGANIZATION: COOPER ENERGY SERVICES GROVE CITY, PENNSYLVANIA i
REPORT INSPECTION QQGnn117/A7 01 RESulTS PAGF 4 of 19 j
NO e
)
30-day requirement for Product Engineering "to evaluate the noncon-formance and determine if a deficiency existed" has been deleted from Section 4.2.2 of QCP-10.14. This is in recognition of the fact-j that, occasionally, it is not possible'to establish the root cause i
of a deficiency within 30 days..!t-should be noted, however, that CES has demonstrated strict adherence to the requirement to notify l
.ne NRC within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> that "a product deficiency exists such'that a major reduction in the degree of protection to public health and safety will occur."
As shown above, CES has taken adequate ' corrective ano preventive s
ar' ion to resolve this nonconformance.
9 1
o E.
OTHER FINDINGS AND COMMENTS-The purpose of the(in'spection conducted on March 30 - April A,1987 was gi to review the ev61uation an4 rcot ccuse analyses performed by CES for 3 N several recent and significant engine failures lat commercial nuclear) h j '
power plants. The cqrent status of -these analyses follows:
- 'ej i
(1) Con.tecting Rod Failure q
OnDe$ ember 23, 1986, engine 3B at Palo Verde Nuclear Station N
suffered a catastrophic failure of the number 9 master connecting rod. Both the r:dster (9R) and articulated (st) rods were ejected i
from the eng%e,'along with their pistoneana'the counterweights.
Ush{g 2he hot oil mist in the sump for fuel and drawing air in thraegh the hole in the engine ccsei the engine continued to run i aft @r the rincident for approximately 50 minutes at reduced speed.
o 0
l
, Plant personnel finally stopped it by spraying fire suppression foam fnto the oil sump through the hole in the case.
'(i l
, r l
i r
5 Examination of the pieces of the 9R (master) connecting rod showed riear' evidence of a fatigue failure, with the initiation point at 3
pe center oil hole between the articulated rod pin bore and. crank 1
pin, bore.
s
\\ t Uraetdllogrsphic examination revealed that both pin boressurfaces had been plated k a depth of approximately 1/16 inch (0.062 inch).
Research by CES into the manufacturingihistory of this rod showed,
that it had been overbored during machining and was iron plated.
(by a subcontractor) te bring it back into dimensional tolerance.
,\\
This, rod (andthreeo$ers)wereovermachinedinthearticulated aqu/ot crank pin Lare'at about the> same time in 1978-1979.s These cdonecLNgurods, and fQe more that underwent cosatic",itbn
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ORGANIZATION: COOPER ENERGY SERVICES l
GROVE CITY, PENNSYLVANIA j
REPORT INSPECTION l
QQQnn117/R7 01 RFStil TS -
PAGF R nf 17 NO
~
I plating, were probably the last such parts iron plated.
CES field I
service representatives had learned of engine distress in a two-cylinder gas compressor in 1977-1978.
Investigation pointed to iron plating as a contributing cause. Thus, CES discontinued iron plating in 1978-1979.
1 It is now clear that iron plating, because of its crystal structure (face-centered cubic), does not possess enough slip planes to be sufficiently ductile. A more ductile material would possess a body-centered cubic crystal structure). The low ductility of the iron plating, in the presence of a stress concentration (the oil hole between the crank pin and articulated pin bores), can cause it to crack even under relatively light loads.
In addition, micrographic
/
examination of the iron plating showed definite columnar grain E.,
orientation. This columnar grain orientation may be a result of a lack of proper plating process control.
N To explore further the circumstances surrounding this failure, the N
by the connecting rod at normal (600-rpm) port for the loading seen inspector reviewed the stress analysis re and overspeed (660-rpm) conditions. Using the combined rod assembly and master rod inertia loading, and a stress concentration factor of 2.0 for the oil hole geometry, the minimum cyclic fatigue factor (CFF) occurs at the minimum material thickness (cross-sectional area) between the articulated rod pin bore and the crank pin bore. The CFF value at this point is 2.98 at 600 rpm and 2.56 at 660 rpm. All other regions of interest typically have a more conservative value for CFF.
These CFF values are significantly above the CES target value of 3
2425, which is a judgment factor based on traterial and machining
')
hriations, unknown load f actors (operator errors), and calculational Varidtions. However, CFF values at CES can range from 1.8 to 2.5.
k' i
Ds Vesigning this engine with a minimum CFF of 2.25 appears to
)-
represent a significant margin above the " perfect knowledge" value
.1 of 1.0.
A CFF of 2.56 (at 660 rpm) in the area between articulated and crank pin bores that contains the oil hole where the break initiated is approximately 14% above the standard CES CFF of 2.25 i-for cuch an engine part.
The standard CFF of 2.25 is, in turn, 1251 above the perfect knowledge value of 1.00.
On the hasis of the NRC inspector's review of the failed engine parts
(\\
and the investigation conducted by the licensee (Arizona Public
$ervice' Con.pany), its materials consultant, and CES it appears that s
t 21 k
x_
ORGANIZATION: COOPER ENERGY SERVICES GROVE CITY, PENNSYLVAhlA REPORT INSPECTION NO - QQQnn117/R7-01 RESULTS:
PAGE 6 of 12 the minimum fatigue factor selected by CES was prudent and that the margins above its established minimum were significant.
It appears that the connecting rod failure experienced at Palo Verde Nuclear Station was caused by electroplated iron in the area of a stress concentration (radius of oil hole).
(2) Procurement The NRC inspectors examined several procurement records to determine whether CES procedures and their implementation met the requirements l
of 10 CFR 21, and if the applicable quality requirements were specified on the CES critical component purchase order forms. Those l
requirements are imposed on subvendors supplying CES with manufac-turing services or inaterials used in the manufacturing process. As part of the review, the inspectors examined four purchase orders placed with the Vander Horst Corporation for the plating of critical (safety-related) engine components. While reviewing these purchase orders, the inspector determined that CES had issued these purchase orders without specifying that the provisions of 10 CFR 21 were applicable. The purchase orders (P0s) were 532190275 dated January 1, 1983, 532190034 dated January 1, 1984, 3921G6372 cated June 23, 1986, and 3921G6941 dated October 17, 1986.
Before the conclusion of the inspection, CES personnel presented to the NRC inspector an inter-office memorandum dated March 19, 1987, which states:
All "N" Code, critical parts purchase orders must contain a reference to 10 CFR 21, Reporting of Nonconformances. Quality Control is requesting the following change to the standard verbiage for these purchase orders.
Material to be supplied on this purchase order is deemed to be a " Critical"/" Safety Related" part for Standby Diesel Engine Generatcr Sets in Nuclear Power Plants.
10 CFR 21 " Reporting of honconformances",
and the enclosed procurement drawings and specifications apply. All required documentation shall be shipped with and/or in advance of the product, and shall be legible and reproducible.
Failure to comply with the above is cause for rejection of your product.
CES personnel informed the NRC inspectors that a review of all PO files would be completed and that the above statement would be made an attachment to every applicable subvendor's P0.
Violation (87-01-01) was identified in this area.
.22
-)
'I ORGANIZATION: COOPER ENERGY SERVICES GROVE CITY, PENNSYLVANIA
.)
i REPORT-INSPECTION-wn'.
ooonn917/n7 n1 RESULTS:
PAGE 7 nf 12 (3) Documentation-The NRC inspectors reviewed Material Review Request (MRR) 2912 dated October 26,-1979, and all associated documentation. MRR 2912 was initiated to document a deviation that occurred at CES during the machining process of a crank shaft bore for the connecting rod that failed at the Palo Verde Nuclear Station in December 1986.
MRR 2912 was generated to iron plate the base material per Material Specification SE-89-2N and inspect the completed iron plating process in accordance with Quality Control Inspection Plan QC/IP-SVG-17.
During a review of the supporting accumentation for MRR 2912, the inspectors noted that Work Order 89998 issued December 5,1978 and referenced on QC/IP SVG-8 was not included as part of the completed work' package.
CES personnel stated that Work Order 89998 may have been discarded because the Record Retention Section in the Quality Assurance Manual was misinterpreted and that this section has been revised to include the work order form as a lifetime document.
Nonconfonnance item (87-01-02) was identified in this area.
(4) Calibration The NRC inspectors reviewed Sections QCP-10-15 and ISP-5-2 of the CES quality control procedures and compared their requirements with calibration records for the equipment.used during the inprocess inspection of various components being manufactured. Section 4.11.1 of QCP-10-15 requires that calibration frequency be in accordance with Inspection Procedure ISP-5-2.
The inspector reviewed ISP-5-2 but found no reference to' calibration frequency.
In discussions with CES personnel, the inspectors learned that Inspection Procedure ISP-5-2 was being' revised and that a reference to calibration frequency wccid be included in the new revision.
Nonconformance (87-01-03) was identified during this part of the inspection.
(5) Certificate of Compliance The NRC inspectors reviewed Material Review Request (MRR) 17053 and all associated documentation. MRR 17053 was initiated to document a deviation that occurred at CES during the machining process of the crankshaft bearing for critical class piston connecting rods utilized in safety-related application at the Palo Verde Nuclear 23 1
-_._._.-______m___
ORGANIZATION: COOPER ENERGY SERVICES GROVE CITY, PENNSYLVANIA REPORT INSPECTION An - ogonn117/A7-01 RESULTS:
PAGE 8 of 12 Station. MRR 17053 was generated to iron plate the base material per CES Material Specification Procedure SE-89-2N. The NRC inspectors reviewed Material Specification Procedure SE-89-2N, Paragraph 3.1, which requires, in part, that vendors performing plating in accordance with SE-89-2N provide CES with a certificate of compliance. On the basis of the documentation reviewed, the NRC inspector could not determine that a certificate of compliance had been issued by Vander horst Corporation to CES.
Nonconformance item (87-01-04) was identified in this area.
(6) Stress Relief Requirements The NRC inspectors reviewed MRR 10328 and all associated documenta-tion. MRR 10328 was initiated to document CES' upgrading of its commercial-grade piston connecting rods to a critical component that would permit use in safety-related applications.
In reviewing Detail Drawing MSV-4-2A, the inspectors found a drawing note (Note 7) that allowed production personnel to waive stress relief requirements for the connecting rod. However, the inspector was unable to determine whether or not production personnel had performed a stress relief process on the connecting rod.
In discussions with CES personnel, the inspector learned that Note 7 would be removed from the detail drawing and that the procedures would be revised to require documentation of the process performed.
Unresolved Item C.1 was identified in this area.
(7) Procurement The NRC inspectors examined CES procurement procedures to determine whether their implementation met the requirements of 10 CFR 21 and whether the applicable quality requirements were specified on the CES critical component purchase order forms. These requirements are to be imposed on subvendors supplying CES with manufacturing services or materials used in the manufacturing process.
The NRC inspectors noted that the procedures did not adequately specify that the requirements of 10 CFR 21 be applied to all critical / safety-related purchase orders.
In discussions with CES personnel, the inspectors was advised that the procedure would be revised to include 10 CFR 21 requirements.
Unresolved Item C.2 was identified in this area.
24
ORGANIZATION:
COOPER ENERGY SERVICES GR0VE CITY, PENNSYLVANIA REPORT IhSPECTION No - 444nn117 /97_ n1 RESULTS:
PAGE 9 of 19 (8) Starting Air Filter Bowl Failure On January 24, 1987, a standby diesel generator at Unit 1 of Houston Lighting & Power Company's South Texas Project failed to start. The engine was undergoing preoperational testing to establish that it could successfully perform a required series of 23 - consecutive starts when the failure to start occurred.
The licensee found that the cause of the failure to start was the 1oss of starting air, which was caused by both in-line air filter bowls falling off the filter.
In checking the starting air filters on the remaining two Unit I standby diesel generators, the licensee found that three of four valves were loose.
(Each unit at South l
Texas Project has three standby diesel generators; each diesel l
generator has two redundant air start lines, with a filter in each line.)
Initial information received from the licensee indicated that the air filter bowls were supplied without an optional C-clip to lock the buwl in place.
On checking with the subvendor who supplied the assembly (Norgren Filter Corporation), CES established that such a locking device is neither available, nor needed.
The filter is supplied with a deformable gasket (0-ring) made of either Buna-N or Viton. The 0-ring was found in place on the filter bowls. Thus, it is concluded that the air filter bowls were not properly tightened after installation and then engine vibration caused them to loosen further.
(9) Rocker Arm Failures Over the last 18 months, there have been 11 incidents of valve rocker arm failures of CES standby diesel generators at nuclear power plants. With one exception, Palo Verde, all rocker arms failed on engines at Commonwealth Edison plants.
The inspectors examined CES Laboratory Report #763, which details the analysis conducted by the CES staff metallurgist of the failed Palo Verde rocker arm.
In addition, the inspectors reviewed photographs of the failed part taken in the CES laboratory.
25
1 ORGANIZATION: COOPER ENERGY SERVICES GROVE CITY, PENNSYLVANIA REPORT INSPECTION An - Aqqnn117/A7-01 RESULTS:
PAGE 10 of 1?
The rocker arm failed in the summer of 1985.
It was cast by CES in accordance with Specification No. h-40N. Although the material composition did not meet the recommended chemical content, the mechanical properties are specified as dominant and were well exceeded. The tensile test bar machined from the broken casting had a tensile strength of 41,000 psi. The rocker arms require a tensile l
strength of approximately 3000 psi during normal service. Thus, the material strength of the rocker arm was fully satisfactory.
1 Visual examination of the rocker arm showed evidence of scuffing where the rounded top of the push rod seats into the underside of the rocker arm. Scuffing can occur if excessive clearance is set into the valve train during adjustment.
(The maximum clearance that can be absorbed by the hydraulic valve lifter is 0.125 inch; the value train is usually set with an initial mid-range gap of 0.000 to i
0.070 inch.) Because gray cast iron does not absorb significant I
impact loads, failures caused by impact are typical and have occurred in the past when valve clearance was excessive.
As stated above, the remaining 10 rocker arm failures occurred at nuclear stations operated by Commonwealth Edison Company. The failures are evenly split between Byron and Braidwood Stations; the most recent failure that preceeded the inspection occurred at Braidwood Station on March 29, 1987.
Although the root cause of all failures is yet to be estermined, the following is known as of this date:
a.
Both rocker arms (intake and exhaust) are found failed approximately 80% of the time.
b.
The crosshead (cam roller and follower assembly) also is found seized in approximately 50% of the rocker arm failures.
c.
The presence and guidance of CES field service engineers does not seem to affect rocker arm failures, d.
Ninety percent of the failures have occurred on engines that had 100-200 hours accumulated run time.
(One engine had 450 hours0.00521 days <br />0.125 hours <br />7.440476e-4 weeks <br />1.71225e-4 months <br /> accumulated.)
26
ORGANIZATION: COOPER ENERGY SERVICES
-I GROVE CITY, PENNSYLVANIA
' REPORT INSPECTION Nn. ooonn117/R7.01 RESULTS:
PAGE 11 of 12 i
Technical information disseminated by CES has not affected the
-failure rate.
Initially, it was thought that the Valve Tappet Adjustment (Hydraulic Lifters) procedure found in the CES technical i
I manual lacked sufficient detail. Upon review by the inspectors, it appears that the procedure is adequate for. use by suitably trained, qualified ano experienced maintenance personnel. A technical bulletin prepared and distributed by CES discussed required clearances to avoid crosshead seizures.
Past engine specifications required a 0.002 to 0.004-inch clearance between the crosshead body and the crosshead bushing. Because seized crossheads were discovered concurrent with broken rocker arms, it I
was thought that a larger crosshead to bushing clearance might i
alleviate the seizures. Thus, CES increased the recommended clearance to 0.002 - 0.006 inch. This had no noticeable affect on broken rocker arms - with or without seized crossheads.
CES is in the process of returning to the previous specification for crosshead bushing clearance.
i As was noted for the Palo Verde Nuclear Station rocker arm failure, valve clearance in excess of the ability of the hydraulic lifter to absorb it will lead to impact failures of rocker arms. This is thought to be the cause of two rocker arm failures, both of which occurred at Byron Station; November 1985 and May 1986.
l 1
Investigations by CES and licensees continue. As stated above, the root cause of these failures is yet to be determined.
j (10) Fuel Injection Line Fitting Failure On February 8, 1987, engine 2A at Palo Verde Nuclear Station sustained some superficial damage as a result of a fuel oil fire caused when fuel injection line SR separated from its fuel injector I
while the engine was running.
The fuel injection line that separated had been purchssed as a "made-up spare" from CES.
(Nonnally, the line and fittings between I
the fuel injection (" jerk") pump and fuel injector are made to fit l
the engine either during manufacturing process or in the field.)
l Investigation by CES revealed that these newly available made up spares were not correctly assembled by CES.
In particular, the nut intended to set and crimp the internal pressure sealine ferrule was insufficiently tightened so that the ferrule did not set itself into the fuel injection tube.
Since for field installation it can be 1
27
ORGANIZATION:
COOPER ENERGY SERVICES GR0VE CITY, PENNSYLVANIA REPORT INSPECTION Nn. qqqnn117/R7-n1 RESULTS-PAGE 12 of 1?
assumed that the ferrule is already set, the field torque specifica-tion is intended for making a secure mechanical connection only, not for setting the ferrule into the outer surface of the fuel injection tube. Thus, the fuel injection line pulled out during engine operation.
After all made-up spares that had been installed (20 on 6 engines) were removed at Palo Verde, they replaced them with fuel lines and fittings that were assembled using its own procedure and method to set the ferrule and ensures a satisfactory mechanical connection.
The key variable to be taken into account is fuel oil pressure (8,000 to 9,000 psig).
In addition, there are subtleties involving fuel line harmonic vibrations, clearances between the end of the fuel injection line and either the fuel injector or fuel injection pump body, and the construction detail of the nut that sets the ferrule and makes the mechanical connection.
In any case, the fitting should be capable of withstanding an internal fuel oil pressure of approximately 15,000 psig.
The inspector reviewed CES specification drawings of the nut (Drawing SF-312 issued February 1,1940) and the ferrule (Drawing SF-338-5 issued in 1946), as well as the SAE thread standards that are based on the Weatherhead (tubing and fitting manufacturer)
ERMETO ferrule and nut.
In addition, it should be noted that Bendix l
Corporation also supplies nuts that are slightly different from the CES Weatherhead-based nut.
l l
l After a careful review by CES and the successful installation of approximately 31,000 fuel injection lines of this design since about 1940, it has been determined that the failure at Palo Verde was caused by the incorrect assembly of replacement fuel injection lines at CES. The past (and current) method of installing fuel injection lines on engines under production is unchanged.
CES is satisfied that although field assembly of fuel injection lines can be problematic, there is no evidence of a fundamental design or procedural deficiency.
28
ORGANIZATION: FRANK ELECTRIC CCRFORATION YORK, PENNSYLVANIA REFORT INSPECTION INSPECTION h0.: 99901091/87-01 DATES: 04/20-23/67 Ok-SITE HOUPS-U CORRESPONDENCE ADDRESS:
Frank Electric Corporation.
ATTN: Mr. L. A. Shinneman President RD 5, P. O. Box CS Willow Springs Lane York, Pennsylvania 17405 ORGANIZATIONAL CONlACT:
P. Lynch TELEPHONE NUMBER:
(717'i 764-5959 KUCLEAR INDUSTRY ACTIVITY: Electrical panels for various applications.
f ASSIGNED IhSPECTOR:
f (x' A/L k24f;
__1
- k. 'R. haidu, Program Development and Reactive Date Inspection Section (PDRIS)
OTHER INSPECTOR (S):
1 P
APPROVED BY:
W e
~
ate ~/
J. V. Stcrc, Chief, PDhlS, Vencor 1nspection Branch INSFECTI0h BASES AND SCOPE:
A.
BASES:
10 CFR Part 21, Appendix B to 10 CFR 50.
B.
SCOPE: Review the implementation of the quality assuarance prograrn E selected areas during the manufacturer of par,els for cuane Arnold
)
and Limerick ar.d review records pertaining to panels furnishec.
PLANT SITE APPLICABILITY:
Duar,e Arnold (50-331),1.in,erick (50-353).
29
r 1
j ORGANIZATION:
FRANK ELECTRIC CORPORATION YORK, PENNSYLVANIA J
REPORT INSPECTION wn. ocon1noi/R7.nl RFSULTS-PAGE 2 of 9 A..
VIOLATIONS:
Section 21.21 of 10 CFR Part 21 states in part, "Each individual, corpora-tion, partnership or other entity subject to the regulations in th.is part shall adopt appropriate procedures to (1) provide for'(1) evaluating deviations or. (ii) informing the licensee or purchaser of the deviation to be evaluated unless the deviation has been corrected."
.Section-16 of the Frank Electric Corporation QA Manual states that nonconformances will be reported as required by 10 CFR Part 21.
Contrary to the,above, FEC did not develop a procedure to implement the above requirements.
(87-01-01)
B.
NONCONFORMANCES:
1.
Criterion IX of Appendix'B to 10 CFR 50 states, " Measures shall be established.to cssure that special processes, including welding, heat treating and nondestructive testing are controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria and.other special requirements."
Paragraph 5.10.1.3.(1) of the American Welding Society standard D1.1 (1982) requires that weld joints prepared to qualify a Weld Procedure Specification (WPS) should be examined by either radiographic or ultra d c testing.
Contrary to the above, welds prepared to qualify 'WPS CP202 Revision I were not subjected to either radiographic or ultrasonic examina-tions; they were only bend tested.
(87-01-02) 2.
Criterion XVIII of Appendix 8 to 10 CFR 50 states, in part, "A system of planned and periodic audits shall be carried out to verify compliance wfth all aspects of the quality assurance program and to determine the effectiveness of the program."
Paragraph 19.1 of the FEC QA manual states in part "To verify the effectiveness of, and compliance to the internal QA program, the QA manager is responsible to coordinate an internal audit. This audit will be perfonned annually, assuming that there is a representative sample available."
30
l l
ORGANIZATION: 1 RANK ELECTRIC CORPORATION j
'l LYORK, PENNSYLVANIA l
REPORT INSPECTION un - 40001091/n7 01 RESULTS:
PAGE 3 of 9 Contrary to the above FEC has not performed an internal audit since 1983 even though panels were being manufactured from 1984 to 1987 for installation in nuclear power plants.
(87-01-03)
C.
UNRESOLVED ITEMS:
No unresolved items were identified during this inspection.
]
D.
INSPECTION FINDINGS AND OTHER COMMENTS:
1.
Background Information FEC manufactures electrical panels to meet various technical specifications for installation in nuclear power plants.
FEC supplied safety-related electrical panels to the following nuclear power plants during 1984-1987.
i Duane Arnold Energy Center (DAEC) 5 Alternate Shutdown Panels (ASP) l 6 Alternate Shutdown Fuse Panels (ASFP) 2 Anticipated Transient Without Scram Panels (ATWS)
Limerick Unit 2 4 Plant Monitoring System Panels (PMS)
South Texas Project 2 Relief Valve Indicator Panels Arkansas Nuclear Power _ Station P. Relief Valve Indicator Panels On March 25, 1987, DAEC informed the NRC of a potential 10 CFR Part 21 defect pertaining to the ASPS manufactured by FEC, and installed at DAEC. Specifically, DAEC determined that welds in the ASPS installed in 1985 were not adequately qualified to the requirements of American Welding Society (AWS) standard D1.1.
This inspection was performed to review the adequacy of the FEC quality assurance program implementation curing the manufacture of the ASPS and other safety-related panels with specific emphasis on I
31
___ __ _ ___ _ _ _______ _ _________ __ ______ _ _ _ ________j
l ORGANIZATION:. FRANK ELECTRIC CORPORATION YORK, PENNSYLVANIA l
REPORT INSPECTION 400n1nQ1/E7.01 RFSlH TS:
PAGF 4 mf 9-NO h
welding. The inspection included a visit to Limerick Unit 2 to inspect the welds on the safety-related PMS panels.
2.
-Review of DAEC Panels Records a.
Technical Specification (TS) 7884-E-134-NI, prepared by Bechtel Power Corporation. Ann Arbor Division, Michigan, specifies the technical requirements for various panels manufactured by FEC.
Based on this TS, DAECLissued Purchase Order (P0) 009560 dated October 18, 1983 for 5 ASPS, P0 28511 dated July 16, 1986 for C ASFPs and P0 29110 dated August 22, 1986 for 2 ATWS panels.
The Technical Specification required welding to conform to AWS 01.1, 1982, specified the components to be used in t.he panels and their locations and required Certificates of Conformances certifying that the components met the IEEE-323 and IEEE-344 requirements. On May 14, 1984, FEC sent the following procedures to Bechtel Ann Arbor for approval:
- 5229 Cleaning and painting procedure
- CP-204 - Marking and cleaning procedure
- CP-H513 Storage requirements for control panels
- QC-1001 - Welding inspection and repair procedure The above procedures were approved and returned.
b.
AWS 01.1-1982 requires Weld Procedure Specifications (WPS) to be qualified by either radiographic or ultrasonic examination of the welds performed to the WPS. WPS CP202 was developed in 1977 to weld 2 sections of 4-inch diameter schedule 40 l
pipe with 30 bevels.
The Procedure Qualification Record (PQR)
)
ind? cates that radiographic examination of the welds on l
December 1, 1977 indicated the welds were acceptable. Welders I
were qualified to WPS CP202 after the welds performed by them l
Were subjected to guided bend tests-and determined acceptable.
In 1981, when WPS CP202 was revised, one of the essential variables, the gas flow rate was reduced from 35 CFH to 20 CFH. AWS D1.1 requires requalification of the WPS when an essential variable is changed ano includes a POR. The PQR for WPS CP202 Revision 1 was incomplete because qualifying welds were subjected to guided bend tests only and not radiographic or ultrasonic examination.
FEC stated that specimen weld coupons will be prepared and sent to an independent laboratory for radiographic and guided bend test examinations and a l
I 32
1 1
I 0 ORGANIZATION:. FRANK ELECTRif 09RPORAT10N.
' YORK, PENN!M' W A.
{
acn j
REPORT INSPECTION wn - coon 11101/n7_n1 RFSUITS:
PAGE 5 of 9 complete PQR will be prepared. One nonconformance was identified for performing welding with an inadequate PQR.
(87-01-02) c.
The following types of welds were utilized during the manufacture of the electrical panels for DAEC and are specified in Note 4 of FEC drawing 08-5883-3 Revision D.
(1) Fillet welds'on outside corners.
(2) 3/16" fillet welds,1" long, 6".or less on centers to weld sheet steel to channels.
-(3) 3/8" plug welds 6" or less to weld hinges to sheet steel.
(4) 3/16" fillet welds to weld channels to angle irons.
9
-(S) 1/4" fillet welds to weld angle iron to angle iron.
(6) 1/4" buttweld to weld channel base webb and flanges.
AWS D1.1 permits the use of WPS CP202 (with an adequate PQR) to perform all'the above welds according to paragraph 5.10.3.3.
This paragraph titled, " Piping and Tubing Qualification,"
states; "A joint welding procedure specification for groove welding of pipe or tubing qualified in accordance with 5-10-1 shall also constitute procedure qualification for fillet welding plate, or tubing in the same position qualified "
l I
d.
Two Supplier Deviation Disposition Requests (SDDR) were initiated relative to problems in welding. FEC initiated SDDR 87-004 dated February 27, 1987 stating that FEC did not
]
have a qualified WPS and filler material to perform structural
)
welding per AWS 01.1, specifically, buttwelds joining 2"x2"x1/4" angle irons at the bottom and top of two ASFPs.
It was proposed that DAEC supply a qualified welder, a qualified WPS and filler material to perform the reference welds. These welds are similar to the ones on the ASPS already shipped and installed at DAEC.
DAEC brought an approved WPS identified as P1-T and a qualified welder to FEC.
FEC rented a Gas Tungsten Inert Gas Welding (TIG) machine for this purpose. An open-butt technique was used to weld two 2"x2"x1/4" angle irons at the bottom and top.
The welds were inspected to FEC procedure QC-1004 which was approved by Bechtel on March 10, 1987.
l 33 i
l i
ORGANIZATION: FRANK ELECTRIC CORPORATION
. YORK, PENNSYLVANIA REPORT INSPECTIOP nn - coon 1no1/A7 n1 RFSHITS-PAGF f of 4 l
l SDDR 87-005 dated March 4, 1987 permitted FEC to use AWS D1.3 j
in lieu of AWS 01.1 for welding criteria. To satisfy the j
requirements of AWS DI.3, FEC prepared and qualified two WPS to l
}
weld 10 gage sheet steel in two weld configurations. Weld I
coupons were bend tested and determined satisfactory. FEC prepared an additional WPS to buttweld two 2"x2"x1/4" angle 1
frons. A backing plate was used. Weld coupons were sent to 1
DAEC's independent testing laboratory for radiographic examina-tion, bend and tensile strengths. During the examination, when I
the backing plate was removed, lack of fusion was observed I
along the root. The weld was rejected without any further
)
examinations. DAEC completed the weld by utilizing a qualified welder, a qualified WPS and equipment rented by FEC. The seismic qualification of the ASPS already shipped were reanalyzed and modifications were reportedly made to the ASPS to meet the seismic qualifications.
3.
Review of Limerick Unit 2 Panels I
FEC manufactured 15 Plant Monitoring System (PMS) multiplexer panels of which 4 were safety-related for Limerick Unit 2.
The welds on the panels were inspected at Limerick Unit 2 and the quality assurance records pertaining to the panels were reviewed at FEC.
a.
PO /18031-M-273-AC dated March 13, 1986, was issued by Bechtel San Francisco and referenced Material Requisition (MR) 8031-M-273.
The MR specified the technical requirements including the following:
(1) The panels were required to be welded to the requiren.ents of AWS 09-1.
(2) The types of cables for internal wiring and the types of terminal blocks to be used in the panels were specified.
(3) Four panels designated as 2CZ723 (Division 1) 2 DZ724 (Division 11), 2AZ725 (Division III) and 28Z726 (Division IV) were designated as 'Q' items. Two input harnesses were also designated 'Q'
- items, 10 CFR Part 21 was applicable to the
'Q'
- items, s
7 34 j
< +
i 4
j
' ORGANIZATION: FRANK ELECTRIC CORPORATION
~
YORK, PENNSYLVANIA I
.l
]
REFORT.
INSPECTION un. -oooninolin7 n1 RFSULTS!
PAGE 7 of 9
)
(4) _FEC was.to provide a seismic analysis / test report.for
.the.'Q' panels. SDDR #1 and Bechtel's letter to FEC
-dated.0ctober 2, 1986, Document 207407, stated that "FEC shall not provide additional analysis / test reports."
FEC stated that Bechtel engineering performed the seismic analysis on the panels, b.
-The welds on the 4 panels and general workmanship were inspected at Limerick Unit 2 for. the following attributes:
- 1) Porosity
- 2) : Size and location' of welds
- 3) Weld spatter (4) Undercut j
(5) Minimum effective throat The inspectors observed excessive weld spatter and reinforcement on some welds in two panels. The corresponding weld inspection 1
checklists' identified this deficiency and documented that corrective action was taken to remove the excessive weld spatter.
'4 c.
Review of the quality assurance documentation indicates the following:
(1) SDDR dated August 25, 1986 authorized FEC to ship the panels without the~ plug-in power cables.
l l-(2) WPSs100 and 101 were used to weld the panels. The WPSs l.
were qualified to groove weld 0.12 to 0.25" sheet metal in the overhead positions using butt and 55 double bevel joint preparations. Procedure qualification test welds l
were visually inspected and determined acceptable to meet 1
paragraph 2.4 of AWS D1.9.
Use of the WPSs to weld in all positions is acceptable per paragraph 2.3.1.9 of AWS 01.9 which states; " Qualification in the overhead position shall qualify the procedure in all positions."
4 Paragraph 2.2.2 of AWS D1.9 states; " Typical joint designs and details of joints are given in Appendix A7, Part A of this specification. Qualification of a groove weld joint may be used as a qualification of a fillet weld joint but not vice versa." FEC drawings 08-5794-8 and -9 specify fillet welds and groove welds. Therefore the WPS is considered acceptable for the welds performed.
35 a_____:___-_.
ORGANIZATION: FRANK ELECTRIC CORPORATION YORK, PENNSYLVANIA REPORT INSPECTION An - 04401041/A7-01 RESULTS-PAGE 8 of 9 i
(3) Welders who performed the welds were qualified to the I
WPS.
(4) The weld inspection checklist identified cosmetic weld spatter on two panels which was subsequently corrected.
(5) Certificates of Conformance were available for the electrical components installed in the cabinets.
4.
Review of the FEC Quality Assurance Programs l
Review of the FEC quality assurance (QA) program in selected areas indicated the following:
An up-to-date organization chart with names was not available a.
due to recent changes in personnel.
In April 1987, the chairman of the board replaced the president and the QA manager resigned.
An acting QA manager has been appointed in the interim.
FEC stated that an up-to-date organization chart reflecting the personnel changes would be completed and incorporated in the QA manual.
]
b.
Adequate procedural controls are effectively implemented in the
{
area of design control through bills of material, wiring i
diagrams, panels, and structural design drawings reflecting the l
technical specification requirements. The technical specifications reviewed specified the components to be used and their location in the panels.
c.
Adequate procedural controls exist for the control of purchased components, commensurate with the size of the FEC organization.
FEC either purchases components from a source listed on their approved vendors list (AVL) or a source listed on the purchaser's AVL. Receipt inspections are performed for incoming material which includes verification of receipt of the certificates of conformances for the materials supplied.
d.
Control of special processes was determined to be inadequate and one nonconfonnance was identified relative to an incomplete FQR. During the exit interview, FEC committed to scrutinizing their records and to organize them to be retrievable and auditaole.
1 1
~
36
ORGANIZATION: FRANK ELECTRIC' CORPORATION YORK, PENNSYLVANIA REPORT INSPECTION un. ooon1no1/n7.n1 DFStil TS :
PAGF 9 of 4 4
e.
In the area of audits, FEC did not implement the commitments in Section 19 of the QA manual. Specifically, paragraph 19.1 of the FEC QA manual states 'that to verify the effectiveness of and compliance to the internal QA program, the QA manager is responsible to' coordinate and perform an internal audit, The audit was to be performed assuming that a representative sample was available. The inspector informed FEC personnel that panels for DAEC, Limerick, South Texas Project and Arkansas nuclear power station were manufactured during 1984-1987, which constituted a representative sample of work in process and therefore audits should have been performed. The inspector infomed FEC personnel that failure to perform audits was an item of noncompliance contrary to Criterion XVIII of Appendix B to 10 CFR 50. Noncompliance 87-01-03 has been identified in this area.
E.
EXIT INTERVIEW:
i The inspector met with individuals mentioned in Section F and discussed the scope and findings of the inspection.
F.
PERSO 45 CONTACTED:
L. A. Shinneman, President C. L. Bollinger, Plant Operations Manager P. E. Lynch, Acting QA Manager D. A. Murphy, Sales Manager
- J. E. Campbell, Chief Engineer l
- Denotes that the individual was not present at the exit meeting on April 23, 1987.
37 i
l "0RGANIZATIONi GENERAL ELECTRIC COMPANY.
PHILADELPHIA, PENNSYLVANIA REPORT INSPECTION INSPECTION NO.: 99900219/87-01 DATES: 05/12-14,27-28/87 ON-SITE HOURS: 30 CORRE5FUNUtHUL AUUKLbb: b6nerai tiectric iompany Breaker Plant Operations ATTN:
P. J. Shaffer, General Manager 6901 Elmwood Avenue Philadelphia, Pennsyvlania 19162 ORGANIZATIONAL CONTACT:
J. Simpson
?
TELEPHONE NUMBER:
215-726-2796 hUCLEAR INDUSTRY ACTIVITY: Medium voltage circuit Dreaxers ano sw1Lcngear enclosures.
O
' k (U
? 9/57 ASSIGNED INSPECTOR:
K. R. Naidu, Program Development and Reactive Date Inspection Section (PDRIS) 7 APPROVED BY:
cuww Jam @ C. Stone, Chief, PDRIS, Vendor Inspection Branch Date INSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR Part 21, Appendix B to 10 CFR 50.
B.
SCOPE: Obtain additional information on the 10 CFR Part 21 item reported by Arizona Nuclear Power Project, review corrective actions taken to resohe a 10 CFR Part 21 report on 4.16r,v circuit breaker and review the implementation of the QA program in selected areas.
FLANT SITE APPLICABILITY: Shoreham (50-322); P610 Verde (50-530).
l 39 1
1 ORGANIZATION: GENERAL ELECTRIC C014PANY l
PHILADELPHIA, PENNSYLVANIA i
REPORT INSPECTION NO.*
99900219/87-01 RESUITS:
PAGF ? ef Q A.
VIOLATIONS:
No violations were identified during this inspection.
B.
NONCONFORMANCES:
)
Contrary to 10 CFR Part 50, Appendix B, Criterion XVIII and Section XVIII of the GE Breaker Plar.t Operations Quality Assurance Manual paragraph 3.2, quarterly process audits were not performed in areas other than Box Barrier Assembly and the manufacture of Rachet Wheels.
C.
UNRESOLVED ITEllS:
s No unresolved items were identified during this inspection.
D.
INSPECTION FINDINGS AND OlHER COPfiENTS:
1.
Background Information General Electric Company (GE), Breaker Plant Operations (BPO),
division manufactures medium voltage magna-blast circuit breakers
( C'B ). Until approximately 1984, BP0 supplied CBs to GE,.Burlington, Iowa for installation in metal clad switchgear cubicles. GE Burlington prepared Engineering Summary (ES) documents which listed all components assembled in the switchgear cubicles. These ESs were transferred to BP0 beginning of 1985. During the manufacture and supply of switchgear cubicles, when components were substituted due to inadequate environmental qualifications, ESs were revised; however, the revised ESs were inadequately controlled. Obsolete copies of ESs were available at BP0 and appeared to have been used by Arizona huclear Power Project (ANPP) to order spare parts.
2.
Keview of 10 CFR Part 21 Report On November C, 1965 Arizone Nuclear Power Project (ANPP), the owner of Palo Verde Nuclear Generating Station (PVNGS) notified the ARC of a potential defcct in the procurement of components frora the GE Switchgear Division in Philadelphia. Specifically, PVNGS identified that GE was processing their Class 1E purchase ordcrs (P0) for safety related Class 1E components as commercial grade components.
This potential defect was observed during an audit conducted by ANPP's agent, Bechtel Procurement Supplier Quality Department (PSQD) on October 16-17, 1985. ANPP took timely corrective action by withholding shipments of components 40
1 ORGANIZATION: GENERAL ELECTRIC COMPANY PHILADELPHIA, PENNSYLVANIA 4
i
' REPORT INSPECTION NO.-
99900219/87-01 RESULTS:
PAGF 3 of 4
.to PVHGS as' appropriate. During a followu) audit conducted during March, 1986, ANPP/PSQD determined t1at the' incorrect PO processing was attributed to GE neither meeting nor passing all the ANPP P0 requirements to participating subvendors.
The results of a reaudit, perforned in September 1986, by ANPP, indicated that GE neither implemented all the proposed corrective actions nor completed review of all ANPP spare parts P0s prior to June 1985.
The adverse findings identifled by ANPP/PSQD audits and GE's corrective actions to correct the adverse findings are as follows:
a.
BP0 personnel in Marketing and Order entry were not familiar in processing Class 1E component orders. Additionally, the GE district office in Phoenix did not forward all the quality requirements to BP0; instead they translated ANPP P0s to internal " Requisitions" and forwarded only the " Requisitions" to BP0. During the preparation of " Requisitions," the quality requirements were not included.
To preclude repetition of the above, the district offices-are now required to forward the customer P0s for nucicar applications to BPO. BP0 developed the following procedures:
QAl 108, Revision 1 dated September 16, 1986, titled, " Order Processing - BP0 Nuclear Switchgear Replacements." This procedure outlines the requirements to process safety-related spare parts. The BP0 shop orders for items for safety-related applications which are manufactured by BP0 are prefixed with 903. Additionally, the catalogue numbers for components procured from GE subsidiaries, are suffixed with 1E. The catalogue numbers for items procured from other than GE subsidiaries are prefixed with E.
Additional codes are used on purchase orders when Certificates of Compliance (CoC) are required to certify that components meet the various applicable IEEE standards.
]
QAI 109, Revision 2, dated May 20, 1987, titled, " Nuclear Related Material Quality Assurance Approval." This procedure establishes the requirements to ensure that material supplied to BP0 is suitable for Class 1E application, including receipt inspection of procured Class 1E material during which the components are inspected and verified to have CoCs.
BP0 procedure 110 titled, " Training of Personnel Involved 3
In Nuclear Orders" outlines the requirements to train those j
41
a ORGANIZATION: GENERAL ELECTRIC COMPANY PHILADELPHIA, PENNSYLVANIA REPORT INSPECTION NO.-
99900219/87-01 RFSulTS:
PAGF d nf 4 individuals involved in processing nuclear orders, manufacturing, inspecting and testing safety-related components. Documentation on training is required to be maintained for 5 years.
Training records indicate that various personnel received training on the above procedures, b.
ANPP determined that some of the components supplied by GE did not have the proper certification. The following are the contributing causes:
1)
The personnel GE assigned to process ANPP P0s did not have prior experience in processing Class lE P0s.
2)
Personnel knowledgeable in these activities were not in the review cycle.
l 3)
ANPP ordered some spares for circuit breakers which GE j
could not supply as Class lE items.
Instead, GE would furnish the spare parts as components for use in Class lE equipment. 'GE stated that individual comaonents in the 4.16kV circuit breaker (CB) or the switc'igear cubicle were not qualified to IEEE 323 & 344 since GE qualified the entire CB or switchgear cubicle with the CB. Therefore, GE considers it appropriate to certify that the ordered saare part was manufactured to the original drawings and t1erefore is suitable for use in Class lE equipment.
J Due to the above, either CoCs were not specified in the P0s or were shipped to GE with the component and the CoCs were raisplaced.
Therefore, GE was unable to substantiate qualification on some of the components.
GE provided ANPP with a list for which GE could not substantiate qualification due to lack of documentation.
4)
GE did not respond in a timely manner to develop written procedures to correct adverse audit findings.
GE marketing had committed to ANPP that all ANPP PCs, including those P0s initiated by Bechtel Power Corporation, agent for ANPP, would be reviewed to determine whether GE adequately conveyed the cuality requirements to their subvendors. GE stated that due to delays in the transfer of documents from their Burlington plant to BPO and manpower shortage, they were unable to meet the commitments in a timely manner.
42
I l
is ORGANIZATION: GENERAL ELECTRIC COMPANY PHILADELPHIA A, PENNSYl.VAHIA REPORT INSPECTION NO.-
404n0719/R7-01 R FSill TS :
PArJ G d o 5)
GE, while centralizing their computer files to improve I
their efficiency, inadvertently deleted a prefix code that triggers a note in preparing a P0 to a subvendor to require compliance with appropriate IEEE standards.
c.
Corrective action taken by GE BP0 appears to be adequate.
BP0 has not experienced problems since they started supplying renewal spare parts in 1985 for complete switchgear cubicles.
The spare parts intended for ANPP are in an access controlled area. GE stated that ANPP intends to cancel the previous P0s and issue new P0s.
Pending receipt of ANPPs new P0s, GE intends to hold the material at BPO.
3.
Review of Purchase Orders l
I The inspector reviewed two sets of P0s issued by ANPP that speci-fied safety related components. The first set, distinguished with a suffix F, was prepared by Bechtel Power Corporation (BPC) 1 as an agent for ANPP. The second set, without a suffix was
]
prepared by ANPP personnel. The inspector determined the j
following.
I a.
The P0s required GE BP0 to supply Class lE components such as medium voltage circuit breakers (CBs) and spare parts j
for CBs which are manufactured by GE BPO, and miscellaneous components such as relays, terminal blocks, etc. which are I
not manufactured by GE BPO. The terminology used in the l
P0sspecifying safety-related components appears to have i
caused some confusion. The GE BP0's distinction between l
" Class 1E components" and " components for Class 1E use" I
was not conveyed to AMPP when the P0s were issued. The broad classification is discussed later in this paragraph.
b.
As indicated in the ANPP's final 50,55(e) report dated January 7, 1987, Harathon type short circuiting terminal boards (SCTB) were erroneously ordered in one of the P0s.
SCTBs are exclusively used to terminate current transformers.
The error occurred because ESs were not kept current by GE Burlington after material was revised.
For instance, ESs 0269A7315 and 0269A7316 prepared in 1977 list items 971 and 973 as Narathon SCTDs. On April 15, 1980, these two ESs were revised deleting items 971 and 973 and adding 991 and 993 specifying GE type SCTDs. Other ESs 269A 7311, 7312, 7313 and 7314 were not corrected to show that GE 43
_---_-----___--------_-_-------J
a s;
w ORG N!ZAT10N: GENERAL ELECTRIC COMPANY PHILADELPHIA, PENNSYLVANIA J
REPORT INSPECTION NO t 49000210/R7-Ol RFSill TS t PARF r, nf 4 type-SCTBs'were furnished in the switchgear cubicles instead of Marathon SCTBs. Marathon type SCTBs are not qualified for safety-related use.
BPO supplies components that they mar.ufacture under a controlled i
quality assurance program as Class 1E components. BPO supplies
. procured commercial grade components as " qualified for Class 1E use," after verification as described below.
c.
In summary, the components ordered can be broadly classified as follows:
1)
Items manufactured at GE BP0 consisting of CBs, replacement spare components for CBs and stationary switchgear cubicles.
j GE supplies CBs as Class 1E components and spare parts as
" qualified for Class 1E use."
The GE certification methodology is discussed in paragraph 4.
)
2)
Items which BP0 purchases from other GE plants. For example, GE Meter & Control Business Department supplies 3
relays, terminal boards and switches. GE Sommersworth, located in New Hampshire, supplies current transformers.
GE Nuclear Energy Business Operations (NEBO), San Jose, California, supplies items such as low voltage molded case CBs, push buttons and transducers.
Devices which are received with a serialized number are supplied as Class 1E devices. Other components are supplied as "oualified for Class 1E use."
3)
GE purchases relays and other equipment manufactured by Agastat, Brown Boveri and Westinghouse. GE requires CoCs for these items. Some of the manufacturers serialize their components and reference the serial number in the CoC. GE supalies these comaonents as Class 1E with CoCs to certify tlat they meet tie relevant IEEE standards if applicable.
4.
GE Certification Methodology for Spare Parts GE BP0 manufactures spare parts in accordance with drawings originally used to manufacture the part. The spare part can consist of an individual part or an assembly of a spare part and a Class 1E component.
The following are typical examples.
44
I ORGANIZATIbH: GENERAL ELECTRIC COMPANY PHILADELPHIA, PENNSYLVANIA 1
REPORT INSPECTION NO.-
94900919/P7-01 RESHITS!
I PAnF 7 nf 4 a.
Auxiliary switch and plunger assembly identified as part number 6353570G012. GE certifies this part as " Qualified for Class lE use" component. This component consists of an auxiliary switch assembled on a plunger assembly in such a manner that when the CB operates, the plunger changes the state of the auxiliary switch contacts. The plunger assembly is manufactured by BP0 in accordance with original design drawing and is considered a commercial grade component. The auxiliary switch is procured as a Class 1E component. After assembling the plunger mechanism and the auxiliary switch, GE designates the entire assembly as
" Qualified for Class 1E use."
b.
Rear Busing part number 084500124G01.
This component is one of the three rear primary disconnect bushings. BP0 manufactures this component as a commercial grade component and certifies it " Qualified for Class 1E Use," based on the premisis that it was manufactured to the same requirements as the component which was installed in the switchgear which successfully withstood the IEEE 323 & 344 qualifications.
c.
Position Switch Operator (PS0) part number 0168C3344 group 002.
This component is a welded assembly which actuates a switch.
The function of the switch is to indicate whether the CB is connected or disconnected to the busbar. BP0 considers this couponent as a commercial grade item and certifies that it is qualified for Class lE use because it is manufactured in accordance with a drawing that was originally used to make a similar component. The similar component was used in the i
switchgear assembly which successfully withstood the seismic qualification.
5.
Review of 10 CFR Part 21 Report On February 19, 1987, GE BP0 notified the NRC that a AM-4.16-250-9HB type circuit breaker (CB) failed to trip electrically at the Shoreham Nuclear Power Plant, owned by Long Island Lighting Company (LILCO).
The plant was in mode 5 (cold shut down). The CB powered the "D" RHR pump motor and did not trip during surveillance testing.
The Cb had to be manually tripped to stop the pump motor. LILCO reported this matter as an abnormal occurrence on February 10, 1967.
J A weld failure in the trip crank was the cause of the failure of j
the CB to trip. The trip crank, identified as 105C9316 group one is a linkage component between the armature of the shunt trip solenoid coil and the CB trip latch. The weld that secures a pin to the trip crank failed. With this weld broken, the CB cannot be 45
ORGANIZATION: GENERAL-ELECTRIC COMPANY PHILADELPHIA, PENNSYLVANIA REPORT INSPECTION
'NO t'
QQ4nn?14/A7-nl RF9ntTRt Pinr o nf 0 li electrically tripped and has to be manually tripped. LILC0 removed the CB from service, disassembled the trip crank and sent it to GE l.
Philadelphia.
On March 3 and 11, 1987, GE sent 4 and 10 spare trip cranks respectively to Shoreham. On February 11, 1987, a GE service representative visited Shoreham and confirmed that the CB could not be electrically tripped.
The failed'CB was identified to be serial number 256A4664-012 and was manufactured-in-1975 at BPO. GE devised L
a go-no-go test procedure (TP) to inspect the integrity of the trip l
crank without disassembling the. breaker.
On March 9, 1987, the TP was implemented at Shoreham to inspect the CBs.
Eight of the remaining 85 CBs were judged to fail the go-no-go test criteria of the TP were returned to GE BP0.
It should be noted that failure of the test criteria does not necessarily indicate that the weld failed.
j Three of the eight trip cranks returned from Shoreham were subjected to 10,000 cycles of operation without failure.
The TP was implemented at Turkey Point nuclear power station and no trip cranks were rejected.
BP0 consulted a metallurgist to analyze the part failure.
Pre-liminary results indicated'that the failed trip crank was observed i
to be bent in a direction other than the direction of the load.
GE does not consider the single tailure at Shoreham a generic issue and does not plan to issue a Service Advisory Letter at this time.
GE plans to inspect an additional 100 CBs. GE proposed to inspect the CBs at Vermont Yankee nuclear power plant in the near future.
GE BP0 stated that in 1986, Tennessee Valley Authority (TVA) contacted them on similar problems in their 4kV switchgear installed at Watts Bar nuc1 car power plant. BP0 sent TVA 17 trip cranks. BP0 did not receive the replaced trip cranks from TVA and therefore could not determine whether there was a probicm at Watts Bar.
6.
Review of QA Program Implementation The implementation of the QA prcgram at BP0 was verified in the j
following areas:
a.
A plant tour of the manufacture of CBs and switchgear cubicle
)
assembly indicated that:
1)
Parts were controlled and stored in bins which were adequately identified.
2)
The calibration of test equipment was current.
46 n
a ORGANIZATION: GENERAL ELECTRIC COMPANY PHILADELPHIA, PENNSYLVANIA i
REPORT INSPECTION i
NO.-
99900219/87-01 RESUITS:
f9AGF nf 4 3) flaterial for Class lE use were stored in an access controlled room.
i 4)
Material intended for ANPP were stored in an access a
controlled enclosure.
5)
In-process inspections were being conducted as required.
i b.
Discussions with marketing and buyers indicated d. hat they were trained and knowledgeable in the requireraer0:s for parts i
intended for nuclear power plants i
c.
In the area of audits, QA had not performed quarterly audits as stated in Section XVIII of their QA manual.
Paragraph 3.0 states, in part, " Audits shall be performed in the,following areas: Products, processes, and systems." Paragraph 3.3 states process audits shall be conducted quahlr}y to measure the effectiveness of a manufacturing process M./<.; welding, stamping, painting, test, etc.) in meeting quality requirements.
The inspector determined that to date.in 1987, only two audits were performed:
Process P88-hB14 to manufacture "Rachet Wheels" was audited on April 21, 1981 and Process P12-HC47 i
to manufacture " Box Barrier" assembly was audited on April 14 1987. The inspector informed the BP0 QA nianager that failure i
to perform quarterly audits was a nonconformance contrary to Criterion XVIll of Appendix B to 10 CFR 50, (99900219/87-01-01)
E.
EXIT INTERVIEW:
)
l The inspector met with individuals nientioned in Section F and discussed the scope and findings of the inspection.
j
]
F.
PERSONS CONTAClED:
f P. J. Shaffer, General flanager R. H. Miller, Engineering Manager J. Simpson, Quality Assurance Manager C. Lambardo, Marketing Manager E. Luggan, Mechanical Engineer F. Broztek, Quality Assurance Engineer C. Buccieri, Marketing i
47 r
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1
'a I
DRGAN!ZATION: LIMITORQUE CORPORATION j
'LYNCHBURG, VIRGINIA r
REPORT INSPECTION INSPECTION
'NO.: 99900100/67-01 DATES: 6/24-7/1/S7 ON-SITE HOUR $
20'
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CdRRESPONDENCEADDRESS: Limitorque Corporation
/
ATTN: Mr. F. M Denham Execut pe Vice President 5114 Woodall h ad J
-P. O. Box 11318 s
Mr. P. McQuillan, QA Admink);trato'r Lynchburg YA 24505-1313 ORGANIZATIONAL CONTACT:
TELEPHOWE NUMBER:
(DC4) 528-4400 NUCLEAR INDUSTRY ACTIVITY: ~ Manufacturer of motor operated valve actuators.
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f ASSIGNED INSPECTOR:
Jeffr F.
, on, Speciel Projects Inspection Date Se on
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OTHER INSPECTOR {S)!
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APPROVED BY:
M.
f-12-g 7 VTE 2 Potapovs Chief, 5tl5, Vendor Intpection Branch ITif e '
1
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i INSPECTION BASES AND SCOPE:
,h 10CFRPart50,AppendixBandh'.0CFRPart21.
A.
BASES:
B.
SCOPE: This insp4:iten was made to review operational, quality assurance,undenvirpnmentalqualificationissuesjrepcrtedtotheNRC s
concerning t hitorque motor operated valve actuators.
?
a A review of,Limpur,que's quality assurance program was not conducted
.i dur/,ng this' inhoect ton.
- A-
/
R, PLANT SIE APPLICABILITY: ; limerick 50-352/353 Br3n,,s, wick 50-325/314, Peach Bottom 50+171/277/278, Trojan 50 344,. Browns Ferry 50-159/260/296, Rancho Seco 50-312. Arkan:,ar, Nuclear One 50-313. Een Orofre 50-206/361/362, Watts Bar 1
50 390/39P.
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LIMIT 0kQUE CCRPCDi7MN 5'
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' INSPECTION h
NO.: '99900100/67-01 RESULTS:
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VIOyTIONS:
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' 0, N1NCONFORMANCES:
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None.
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STt.TUS Ok PREVIOU59NSPECTION FINDINGS:
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'ALrevieA f previous' inspection findings was not performed. Previous j
inspectik,itemspilbereviewedduringafutureinspection.
j D.1 OTHER FINDINGS $YD C0K'iENTS: \\
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X 1. ' 7/etuator'Sizind and Performars1 N
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lt was determined that motors usdd in Limitorque act'uators 1sre '20%
s rated dutyJ.otors. DC motors are rated to run at 20% of the' motor's
~
maijaustrated torque output for five midutes 4hile AC motors have a
. ]
'5 W minv.e rating. For example, a 10 ft-b DC rated motor is' designed 1',
i
< to run for five minutes at 2 ft-lbs without excec.dingstemperature j
rira specifications.
It y,as also determined that a typical AC motor J
wiYi put out approximately 10% more and a DC motor 50% more than
, rated maximum torque before reaching a stalled condition.
iRecently a numbw f phnts hdy'e reported theOhave found actuators j
s
'thatcontainedundersizedmotors.)oThis.probleefppenstohave 1
t been caused by the ese of wrong d(fferential prcssu er, the fai?ure q' V,
to account for degraded voltage conditions, or the v[se' of incorrect stem f actor coefficients during original sizing calculations. These calculations were usually perforncddy the valve manuf%cturer or the plant architect. engineer.
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Hydraulic Loch T[ sting Progra s-t
[
Areviewwascond3ctedofLtNitorqnCs,,orogramtotestactuatorsfor the potential of spring pack hyd%v1F; lock. Spring pack hydraulic lock was shown to octar on an act.dator: removed from the Limerick Nuclear Power Raat during a spr.citNy ' conducted test.! Hyofaulic.,
forced % nuickO com) press during valve seating.pring pack is filled with g lock occurs when the s
Theitrappr d. gr em el r
A cannot h.< lieved quickly, the torque switch does not over;'and the l"
actuator. essentially becomes locked. This condtion results in '
I rapid % tor overheating and subsequent motor burneut.
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ORGAhlIAT ON: LIlilTORQUE CORPORATION i
LYNCHBURG, VIRGINIA REPORT INSPECTION
{
'NO.: 99900100/87-01 RESULTS:
? AGE 3 of 8 I
Limitorque has set up a similar actuator to the one that failed at Limerick and has tried to obtain the hydraulic lock condition.
)
Testing had not been completed at the time of the review and no firm i
conclusions had been drawn.
j l
3.
Spring Pack Stress Relaxation
]
Problems with spring pack relaxation on normally closed valves have j
recently.been reported. At Peach Bottom, Limitorque traced the j
problem to tne rebuilding of the spring packs with parts of slightly different dimension. Specifically, the bearing cartridge stems had be3n redesigned and had a difference of 1/16" in length. This 1
t difference was said to have caused the relaxation observed. A p\\
spring pack was recently returned to Limitorque from the Salem i
Nuclear Plant where it is to be analyzed. No reason for the I
y ksrelaxation of this spring pack had yet been determined. This h
problem was also reported to Limitorque by Trojan and Browns Ferry but Limitorque has not received spring packs from these plants for
]
analysis.
{
4.
Improper Spring Packs Installed in Brunswick Actuators d
( On June 22, 1987 the Brunswick Steam Electric Plant issued LER I
B7.-018 which described the discovery of two undersized spring packs i
that were thought to have been furnished with two actuators purchased from Limitorque some time ago. The actuators were part of n rorchase order in which four actuators were procured for use in i
tfi RHR system of Units 1 and 2.
The two actuators which were installed in Unit 1 were found to contain the incorrect spring packs whereas the two actuators in Unit 2 have not yet been inspected.
, Brunswick had no record of ever changing the spring packs in these
' actuators so their installation was attributed to vendor error. The incorrect spring packs that had been installed had part numbers that vs. 60-600-0011-1) git from the correct spring packs (60-600-0010-1 differed by one di Discussion with Limitorque revealed that no documented checks for s5 (proper spring pack installation were performed at the time these
' actuators were manufactured. Limitorque stated that large actuators j
such as those in question are normally manufactured in quantities of 4-8 and that they would investigate to see if similar actuators made during the same time frame may also have had incorrect spring packs installed.
v 51 s
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)
ORGANIZATION: LIMITORQUE CORPORATION LYHCHBURG, VIRGINIA REPORT INSPECTION NO.: 99900100/87-01 RESULTS:
' AGE 4 of 8 6.
Warped Limit Switch Rotors Supplied to Rancho Seco On February 10, 1987 the NRC received a 10 CFR Part 2'1 report from the Sacramento Municipal Utility District (SMUD) concerning warped limit switch rotors that Limitorque had supplied to the Rancho Seco Nuclear Generating Station. The report described a shipment of 50 rotors of which 7 were found to be warped sufficiently to prevent proper limit switch adjustment. Review of this issue with Limitorque determined that previously only samples of rotors were checked for straightness. Limitorque is now checking 100% of safety related rotors for straightness and is in the process of notifying all nuclear customers of the potential for warped rotors for those rotors previously supplied. Limitorque has not determined the cause for the warped rotors however, they did acknowledge the fact that u rped rotors had been discovered during the sample inspections that had been previously performed.
7.
Criteria fur Exceeding Published Actuator Thrust Limits A discussion was conducted with Limitorque concerning the criteria given to Arkansas Power and Light Company for exceeding published actuator thrust' limits on Limitorque's installed in the Arkansas Nuclear One plant. Limitorque stated that published thrust limits could be exceeded by 107. as measured at torque switch actuation by MAC or MOVATS type testing.
Actuators can then be operated for a maximum of 100 cycles at this condition. For actuators that have been found to have had thrust limits exceeded, Limitorque recomends performing NDE on the actuator housing, housing cover, and drive sleeve. Additionally they recomend performing a visual examination on the upper bearing for Brinell marks. The actuator can then be operated safely at or below the published thrust ratings, 8.
Cracked Grease Seals on Motors Shipped to San Onofre During a recent NRC inspection at the San Onofre Nuclear Generating Station the inspector was shown a number of newly purchased Limitorque (Reliance) motors that had been shipped to San Onofre with cracked or improperly installed grease seals. Limitorque is currently investigating the cause of cracked seals with their supplier Reliance Electric and has indicated appropriate actions will be taken as required.
52
ORGANIZATION: LIMITORQUE CORPORATION LYNCHBURG, VIRGINIA REPORT INSPECTION NO.: 99900100/87-01 RESULTS:
l ' AGE 5 of 8 9.
Shaft-Keyway Failures in Actuators at Watts Bar On November 12, 1986 the NRC received a report from the_ Tennessee Valley Authority concerning a failed motor pinion key and shaft in a limitorque actuator installed in the Watts Bar Nuclear Plant. The failure was reported to be similar to the condition reported in IE Information Notice 81-06 which addressed failure of motor pinion keys in Limitorque SMB-4 actuators with motor torques in excess of 100 ft-lbs. A subsequent letter from Limitorque recommended replacing the original mild steel keys with high strength keys on actuator sizes three and four with motors of 150 ft-lb or greater.
The failure at Watts Bar occurred on a size three actuator with a 100 ft-1b motor not requiring key replacement.
Limitorque analysis has shown that the safety factor of the motor pinion key is 3.9 for sheer stress and 3.26 for contact stress as calculated for the original mild steel key installed in a size three actuator equipped with a 100 ft-lb motor.
In a letter dated January 20, 1987 to TVA, Limitorque stated they believe the reported failure was due to the fact that the pinion key was found in a position where it only partially engaged the rotor pinion and shaft.
- 10. environmental Qualification Issues A.
Loading of actuators during testing It was determined that actuator loading during most of the Limitorque Environmental Qualification Tests was by means of a thrust tube. This method of testing provides a load only i
J at the end of the actuator closing cycle. The load is achieved l
by driving a stem into a stationary thrust tube. The tube stops the stem travel and the actuator is loaded until motor current is interrupted by means of the torque switch. The load achieved during testing is therefore directly related to the torque switch setting of the actuator. Review of Limitorque test report B0212 shows that the motor installed in the tested actuator was rated for 15 foot-lbs. Using average torque values obtained before the test, a motor torque output of 14 foot-lbs was calculated to have been achieved during the EQ test of this actuator.
In Limitorque test B0009, a 25 foot-lb.
rated motor was loaded to 13.4 foot-lbs. The ability of the Limitorque motors to put out full rated torque was therefore not proven during these EQ tests. Rated thrust outputs of the actuators were however achieved during botn B0212 and B0009 tests.
53
ORGANIZATION: LIMITORQUE CORPORATION LYNCHBURG, VIRGINIA REPORT INSPECTION N0.: 99900100/87-01 RESULTS:
PAGE 6 of 8 B.
Degraded voltage testing of actuator It was determined that Limitorque Report B0212 is the only report that describes testing of an actuator during degraded voltage conditions. The applied voltage in all other tests was the nominal rated motor voltage. Calculations show that at a 90% nominal voltage condition motor torque output will decrease by some'19%. Limitorque Report B0212 is for an AC Reliance motor. DC motor performance under degraded voltage conditions would be considerably different.
The statement'concerning degraded voltage in Limitorque Report B0058 was discussed. It was determined that the motors installed in the tested actuators were sized based upon calcula-tions done using nominal motor voltage. Had degraded voltage been taken into account larger motors would have then been required on some of the tested actuators. To take credit for the Limitorque tests licensees therefore need to show their actuator motors were properly sized for the applicable voltage conditions.
C.
Similarity analysis between Class B AC Peerless, DC Peerless, DC Reliance, and AC Reliance Motors A similarity analysis contained in a Wyle letter dated August 10, 1982 to Stone and Webster was reviewed. This letter contained a material analysis of the materials used in Class B AC Peerless, AC Reliance DC Reliance and DC Peerless motors.
The analysis showed that the materials used in the manufacture of the Class B DC Peerless, DC Reliance and AC Peerless motors were equal to or better than those used in the manufacture of the Class B AC Reliance motor tested in Limitorque report B0003. No deficiencies in the Wyle material analysis were noted.
D.
Qualification of Nylon Insulation Wire Joints Actuators equipped with dual voltage motors have been found to contain nylon insulation wire joints. These wire joints were used at Limitorque to make connections on dual voltage motors. Actuators tested in Limitorque reports B0003, 600376A, and 600198 contained dual voltage motors that likely contained some type of insulated wire joint. No documentation 54
ORGANIZATION: LIllITORQUE CORPORATION LYNCHBURG, VIRGINIA REPORT INSPECTION NO.: 99900100/87-01 RESULTS:
) AGE 7 of 8 as to the exact type or as to the joint manufacturer exists at Limitorque. Additionally, no configuration control exists that would ensure that wire joints are kept away'from conducting materials. There is therefore no documentation that exists at Limitorque that would supp et environmental qualification of these nylon insulation wire joints.
E.
Limitorque Test B0119 Limitorque test bO119 describes at test in which a Harathon 300 type type terminal boarc was used to power a motor to a Limiterque actuator. The terminal board was subjected to an inside containment type environment and was shown to adequately transmit power to the subject motor at selected periods through-out the LOCA simulated portion of the testing. Resistance readings of the terminal board (terminal to terminal and terminal to ground) were then taken by disconnecting the motor imediately after energization.
The lowest reading obtained durirg the testing for the Marathon 300 terminal board was 900 ohms. Since the motor performed adequately with the 900 ohms measured insulation resistance this value was set as the acceptance criteria for insulation resistance.
Upon review by the NRC inspector, it was determined that the measured 900 ohms insulation resistance was taken using a Biddle 21159 megger. Inspection of this megger revealed that readings in the 900 ohms range could only have been read on the meggers e
ohms scale. This megger has.four scales with four associated open circuit output voltages. With the ohms scale used in obtaining the 900 ohm reading, the output voltage applied to the terminal board would have been very low. From discussions with the manufacturer of the megger, Biddle Instruments, it has been determined that the meggers output voltage under the above conditions would have been approximately 2.2 volts.
Although the other tested boards exhibited insulation resistance readings better than 900 ohms, some of these readings were also obtained using the meggers ohms scale. Consequently, the ability of the other terminal boards tested in 80119 to adequately supply a 480 volt or 120 volt circuit is indeterminate. Under the test conditions, all terminal boards would have indicated 0 ohms i
on the Biddle megger using the 500 volt or 250 volt scale.
l 55 i
i 1
ORGANIZATION: LIMITORQUE CORPORATION J'
LYNCHBURG, VIRGINIA REPORT-INSPECTION NO.: '99900100/87-01 RESULTS:-
PAGE 8 of 8 Additionally, a data transcription error was discovered in the review. The values obtained at event 5.for the Curtis "L" type 1
board Nere really 800 ohrs, 700 ohms, and 1K, not 800K, 700K,
'i and 1K as indicated in test report B0119
.These values are j
therefore' below the 900 ohm Limitorque acceptance criteria.,
l 1
In conclusion, the ability of the Curtis "L " Harathon 1600,
{
Buchanan 0222, Buchanan 0524 and GE-EB-5 type terminal boards 1
to adequately supply'480 or 120 volt power to notor or control q
circuits can not be determined from Limitorque Report 80119.
I i
i 1
i 1
l
}
56
ORGANIZATION: NANCO CONTROLS, INCORPORATED MENTOR, OHIO REPORT INSPECTION IhSPECTION NO.: 99900378/87-01 DATES: 08/10-14/87 ON-SITE HOURS:
CORRESPONDENCE ADDRESS: Namco Controls, Incorporated ATTN: Mr. N. E. Swanson President-7567 Tyler Boulevard i
Mentor, Ohio 44060 i
ORGANIZATIONAL CONTACT: Mr. Douglas A. Coe, Sales Application Engineer TELEPHONE NUMBER:
(216) 946-9900 NUCLEAR INDUSTRY ACTIVITY: Supplies safety-related limit switches and conducts environmental qualification (EQ) testing of its designated limit switches for the commercial nuclear power industry and the military. All of the EQ testing is for commercial' nuclear power. Approximately 10 percent of its manufacture products are for commercial nuclear.
ASSIGNED INSPECTOR:
k(2 fM[c[t/ [ )l/Md 9/1 r/
R)(N. MoYst, Sp'ecial' Projects Inspection Section Date SPIS)
OTHER INSPECTOR (S):
D. Brosseau, Sandia National Laboratories bed k[a mp CJ.g APPROVED BY:
U. Potapovs. Chief, SPI}, Ve6 dor Inspection Branch
) ate INSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR Part S0, Appendix B and 10 CFR Part 21.
B.
SCOPE: This inspection consisted of:
(1) Technical evaluation of j
equipment qualification (EQ) test activities for safety-related equipment.
PLANT SITE APPLICABILITY: Various.
57 i
'0 ORGANIZATION:
NAMC0 CONTROLS, INCORPORATED MENTOR, OHI0 REPORT-INSPECTION NO.-
04Gnn17R/A7-01 RFSill TS ?
ParJ 9 nf n i
A.
VIOLATIONS:'
]
None.
B.
NONCONFORMANCES:
1 1.
Contrary to' Criterion V of Appendix B to 10 CFR Part 50 and Section 1
11 of Namco Standard' Practice (NSP) manual 20-0001 dated July 1985, 1
Revision A, the' director of engineering did not conduct a' seminar i
for the review of applicable portions of Namco Controls QA manual and Namco standard practices on "no less than a annual basis."
1 I
Namco's' last seminar was conducted on January 3,1986.
2.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Section 12.6.1 of NSP 20-0008 dated July 1985, Revisior. A, the calibration i
label for thermocouple serial number TC-11 did.not identify the date -
last calibrated and had the wrong date idelitified for when the calibration was due. Also, the c.alibration due dates for the same thermocouple were recorded on the gage record card as one date and as a different date on the instrument equipment sheet of Test Report T.R. 308-1.
C.
UNRESOLVED ITEMS:
None.
D.
OTHER FINDINGS OR COMMENTS:
1.
Background - Namco Controls had dedicated four limit switch 1
part numbers series for nuclear power plant class 1E applications; EA180 and EA740 series for insice containment applications; EA170-302 and EA705 for outside containment.
The EA180 series switch was developed in 1976 and first qualified to IEEE-323, 344 and 382 on March 1978. The series was requalified in August of 1981 to provide for product improvements and material changes.
Currently, the series is being requalified together with a receptacle and cable connector assembly and to also provide for recent material. changes.
The qualification report for this testing is currently being developed.
The EA740 and EA750 series limit switches evolved from the SL3
" snap lock" switch series designed in 1959. The EA740 series switch was seismically qualified to IEEE 344-1971 in 1974.
In 1976 the EA740 series switch was subjected to extensive materials development 58
l
'~ ORGANIZATION: NAMC0 CONTROLS, INCORPORATED
. MENTOR, OHIO REPORT-INSPECTION WA'.
QQQAA17R/R7.01 RFRf H T9 Danr 1 nf Q
\\
l 3
testing which resulted in qualification to IEEE-323, 344, and 382 in l
1978. There were further product improvements and material changes j
in 1979! Currently,'the series is being requalified together with a 1
receptacle and cable connector assembly and to also provide for l
material changes.
]
'2.
Technical Evaluation - The NRC inspector and Sandia Consultant (NRC inspection team) conducted a technical evaluation of three ongoing programs for qualification testing of safety-related electrical equipment. The following table summarizes the programs examined, i
equipment type, and types of documents examined.
Program Equipment Type Documents Examined j
1.
EQ180 Limit Switch with Test Plans (TP),
l material-changes & with Test Report (TR),
receptacle and connector /
Purchase Orders l
cable assembly.
(P0's),TestData (TD) and supplements to TR's.
i l
l 2.
EA740 Limit Switch with TP, TR, P0 and TD.
l material changes & with receptacle and connector /
cable assembly.
l l
3.
Material analysis of impreg-TD.
nation materials used for qualified limit switches.
The NRC inspection team reviewed the EQ process prescribed in each test report / test plan and reviewed test results for the above programs, including the bases for accelerated thermal aging and radiation for two programs.
1 Three test programs were examined to ensure that Namco had controlled
)
the following as applicable.
l l
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.
I 1
c.
Test acceptance criteria were established as described in the f'
test specification or in the design engineering documents, such i
59
\\
ORGANIZATION: NAMC0 CONTROLS, INCORPORATED MENTOR, OHI0 I
REPORT INSPECTION QQfinn17A /A7 01 DFR!ff TR-Dant A nf n MO
+
as calculations and engineering letters to meet the requirements cf IEEE-STD-323/1974.
l d.
Same test assemblies were used for all phases of testing and represented standard production items, e.
Environmental conditions were established and described (e.g..,
pressure and temperature profiles, and thermal aging factors were consistent with those outlined in the test specification or. test procedure).
f.
Test results were adequately reduced and evaluated against established acceptance criteria described in customer test specifications or purchase orders, g.
All prerequisites for the given tests as outlined in the test l
plan had been met.
h.
Test equipment included a description of all materials, parts, and subcomponents.
i,.
Notice of Anomaly reports were properly documented.
j.
Appropriate margins were applied, k.
Functional performance requirements were met.
No nonconformances were noted during this review. Tho fo; lowing j
describes each program:
1.
Program 1 and Program 2 The inspection team reviewed a current qualification test plan (QTP) 215 titled " Qualification test plan of EA180 and EQ740" which governs requalification of EA180 and EQ740 switches with new replacement materials. Because of deletion of the original asbestos filled materials (Rogers Rx490), the switch contact block and carrier now consist of a glass filled phenolic thermoset plastic by the same manufacturer (Rogers Rx865).
The EA740 switch also includes Amoco Torlon for the contact carrier assembly.
A review was made of the next generation QTP 207-1 titled,
" Test Plan for generic qualification of series EA180/EA740 switches...requalification, with electrical receptacles and 60
+
i 0 ORGANIZATION: NAMCO CONTROLS, INCORPORATED MENTOR, OHIO t
REPORT INSPECTION Nn - 4QQnn17A/A7 n1
- RFSill TS t P ArJ 6 ' nf A
{
connector / cable assembly with new contact block / carrier material
which governs ongoing tests 'by Namco of the housings with the new materials and with receptacles and cable / connectors. The testing consisted of various combinations of each family of l
switch housing _ type, along with EC210 receptacles and cable /
connector essemblies. The EC210 receptacles on both one inch and one half inch sizes are made up of BIW Bostrad 7E wire, 600v, 90 degree C_ rated, with stranded copper wire and CPSE jacket over EPR. The connector / cable assembly has EPR o-rings, Okonite VFR, 600v, 90 degrees C cable, with CPSE over EPR. The test plan called for a' minimum.of twelve inches of cable to be thermal and radiation aged. Pictures in the test reports showed that cable well in excsss of the minimum length was actually aged.
Both EA180/EA740 switches have a qualified life of 20 years at 54.1 degrees C with the Rx865.
Seal life was extended from earlier testing due to longer accelerated aging to 23.19 years at 50 degrees C.
Cables were all qualified in excess j
of 40 years.
Lead wire qualification for BIW was established by similarity to BIW report No. B915, November 1980, "Bostrad 7E cables, Flame and Radiation resistant cables for nuclear power plants." Cable assembly qualification was by similarity to Okonite report, " Qualification of Okonite Ethylene-Propylene Rubber Insulation for nuclear plant service," Form N-1, dated December 20, 1977.
The inspection team reviewed TR's 308-1 and 308-3 which are interim reports associated with QTP-215 for EA180 housings with the new materials but without receptacle end cable / connectors.
One test item was tested with a backup reperted in TR-308-3.
These reports will eventually be bouno in a new QTR.
The QTR will be reviewed during a future NRC inspection.
A number of supplements have been added to past QTR's and will be incorporated in upcoming QTR's. Supplement 1 addresses the effect of storage time and temperature on qualified life.
It was included in QTR-105 (EA-180) and QTR-111 (EA740) packages.
Supplement ? includes a thermal service capability analysis and radiation tolerance capability evaluation of the Rx865 material, in addition to the similarity presentation of material properties with the original Rx490. Supplements to_EA740 test reports include an evaluation of the replacement Amoco Torlon 4203L glass filled phenolic polyamite-imite, as well as the Rx865. The supplement analysis was performed by National Technical Systems (NTS).
The conclusion reached was that the replacement materials meet, and most times exceed the original 61
9 ORGANIZATION:
NAMCG CONTROLS, INCORPORATED MENTOR, OHIO REPOR1 INSPECTION Nn a
n00nn17A/n7 n1 R F9fil T9 -
PanF A nf n qualification of the Rx490 materials.
The analyses were brief but reasonable and sound in the approach and methods used.
Supplement 3 addresses the engineering tests which were conducted by hamco to support the conclusions of Supplement 2.
Engineering lab report LP1114-4 titled, " Engineering Tests of EA180 series limit switch with new materials," documents the results.
In this supplement to QTR-105 (for example), wear life for' extended cycles of operation was monitored. Normal wear was noted at 593,022 cycles (100,000 cycles qualified life) and testing was terminated at 1,356,356 cycles with no abnormal wear or synergistic affects noted. The samplec responded well to thermal stress, heat shock, cold shock testing, dielectric breakdown test, UL rating tests, and NEhA overload and rated endurance tests.
2.
Program 3
.The inspection team reviewed data sheets for tests completed, both in 1986 and recently, with impregnations of each of the two substances. Testing of housings with impregnations of mixtures of materials (Imprex and Impco) is currently in progress. Little new testing has occurred since the last NRC inspection in July 1986. Namco has developed a comprehensive impregnation specification that is now out for comment. This program will be reviewed during a future NRC inspection.
3.
Folicwup on 10 CFR Part 21 Report This report addressed cracked contact lever arms discnvered at Niagara Mohawk's Nine Mile point in qualified EQ 180 limit switches.
The contact levers were returned oecause of cracks through the plastic lever mostly in the direct area of the retaining clip rivet.
Namco performed life cycle testing on one of the contact lever arms returned by Niagara Mohawk and on a new production lever arm (EA 170-21-1302) which exhibited similar cracking.
Both arms exceeded the 100,000 cycle qualified life by a large margin.
Namco had also detennined that no other field problems attributable to this condition had been reported from a total of 54,500 switches shipped since 1979.
Based on the above, Namco concluded that the cracking condition was limited and would not impair the switch cperation.
However, the riveting
-process has been modified to reduce the probability of over-stressing and additional Quality inspections have been 62
ORGANIZATION: N14tCO CONTROLS, INCORPORATED MENTOR, CHIO 1
REPORT INSPECTION NO = QQQnn17R/A7-01 R FSill TS:
PArJ 7 nf A instituted to assure that the riveting process.is properly j
controlled.
i 4.
Configuration Control of Nuclear vs. Standard Grade The inspection team reviewed Namco Standard Practice (NSP) 20-0001, NSP 60-0010, NSP 20-0003, NSP 20-0005, NSP 20-0004 and NSP 40-0003. Topics covered by the above NSP's included drawings, instructions, specifications, procedures essential to the procurement, fabrication, assembly and testing of a product, identification and control of material which b
requires traceability, new release documentation, engineering j
change requests and engineering change orders.
It was determined after the above review that nuclear limit switches and component parts have unique part numbers assigned to differentiate them from standard products.
Further, these items.(huclear) are identified in stock by means of a unique flourescent label which contains part number, production order or purchase order number, date'of acceptance, inspection stamp and stock locations. Also the material handler filling a production order designated " process per NSP 690-0010" records on that order the lower level production order or purchase order numbers from the label.
Finished items are marked or labeled with the production order or purchase order number to assure traceability.
Implementation of the ai
,e process would have to be verified at Namco's manufactur4 g plant at i
Newton, North Carolina.
5.
QA Program Implementation Review The NRC inspection team reviewed three program files containing
)
documentation which supported testing efforts and selected j
sections from Namco's implementing procedures pertaining to j
the test lab.
Documentation from the three program files included purchase orders, EQ test plans, test reports and test data. The inspection review of the above documentation and 4
hSP's was to verify continued implementation of Namco's QA program. Two nonconformances were noted during this review which are described in Section 8 of this report. Nonconformance i
B.2 was corrected prior to NRC inspection team leaving Namco.
j 63
)
l l
'1 ORGANIZATION: NAMCO CONTROLS, INCORPORATED NENTOR, OHIO l
REPORT INSPECTION
)
'Nn - QQQnn17R/R7 n1 RESULTS:
PAGE 8 of 8 I
6.
-QA Manual and NSP Review The NRC inspection team reviewed the changes that were I
incorporated into the different sections of both.the QA manual l
and Namco Standard Practice Manual (implementing procedures) as-
{
'they pertained to the test lab. The-QA manual.and NSP's are also used at Namco's manufacturing plant at Newton, North Carolina. The NRC inspection team determined that the changes were minor and did not change Namco's QA program with respect to' the requirements of Appendix B of 10 CFR Part 50.
{
64
- _ =
i
,m o
o
'0 ORGANIZATION:-' PATEL ENTERPRISES, INCORPORATED i
HUNTSVILLE, ALABAllA
.i iREPORT INSPECTION INSPECTION i
j NO.: 99900931/87-01 DATES: 07/13-17/A7 nuaTTr unitet.
n
'l i
-CORRESPONDENCE ADDRESS:
Patel Enterprises, Incorporated ATTH: Dr. M. C. Patel, President 3400 Blue Spring Road, Northwest Suite D-3 Huntsville, Alabama 35810' I
ORGANIZATIONAL CONTACT: Mrm T. Utech TELEPHONE NUMBER:
(2n5) REC-Ennn l
NUCLEAR INDUSTRY ACTIVITY:
Patel Enterprises, Inc. (PEI) provide nuclear
-1 services related to the following: equipment qualification program develop-ment; material aging analysis; qualification documentation assessment; qualification maintenance program; and seismic qualification programs to include design'and analysis. Approximately 50 percent of the services.are nuclear power related. Approximately 20 employees are devoted to these efforts.
{
ASSIGNED INSPECTOR:
Mamde#pA }). 2720014 8/M/87 R. N. Moist, Special Projects Inspection Section Date (SPIS) l OTHER INSPECTOR (S):
i APPROVED BY:
2-<A p R-h 4'7 U. Potapovs, SPIS, Vendor Inspe: tion Branch Date INSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR Part 50, Appendix B and 10 CFR Part 21.
4 B.
SCOPE: This inspection consisted of:
(1) a technical evaluation of equipment qualification (EQ) test activities for safety-related equipment and (2) verification of implementation of the quality 1
assurance (QA) program.
PLANT SITE APPLICABILITY: Robinson Unit 2 (50-261), Brunswick Units 1 & 2~
(50-325,50-324), Cooper (50-298).
i I
65 n.
W i
ORGANIZATION: PATEL ENTERPRISES, INCORPORATED i
HUNTSVILLE, ALABAMA l
y REPORT INSPECTION kn!- '4440n431/R7-Ol-RESULTS:
PAGE 2 of 9 A.-
VIOLATIONS:
i None.
8.
NONC0hFORMANCES:
None.
C.
UNRESOLVED ITEMS:
None.
D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
1.
(Closed) Nonconformance (86-01-01)
Contrary to Criterion V of Appendix B to 10 CFR PART 50 and paragraph 6.5.2 of the Quality Assurance' Manual, test procedures PEl-TR-860500-01 for Job Humber (JH) 8605 and PEI-Tk-860100-01 for JN 8601 were revised, however, the original controlled copies-of both test proce-dures were not marked V0ID or SUPERSEDED.
The NRC. inspector verified that both test procedures have been properly marked SUPERSEDED. The NRC inspector reviewed a Patel letter that identified training had been conducted for key personnel at Patel by the Director of Product Assurance on June 23, 1986. The scope of the training reflected on selected sections of the QA manual and Quality Engineering procedures such as document control, control of engineering and design calculations, procurement document control and control of technical documents. The NRC inspector also reviewed j
snd verified that original controlled copies of test reports and test procedures for JN's 8617, 8701 and 8702 were marked SUPERSEDED.
{
2.
(Closed) Nonconformance (86-01-02)
Contrary to Criterion V of Appendix B'to 10 CFR Part 50 and Section 3.4 of Quality Engineering Procedure 4.1, Patel crally authorized their vendor to perform testing to different requirements than specified in Purchase Order 1024 for JN 8603 without documenting this authorization.
The NRC inspector verified that Patel had revised test procedure PEI-TR-860300-01 to revision A and that the changes requested by New York Power Authority were included in the revision. The NRC 66
ORGANIZATION: PATEL ENTERPRISES, INCORPORATED HUNTSVILLE, ALABAMA 1
REPORT INSPECTION wn - oconna'11/g7 nl RFSulTS:
PAGF 3 of 9 inspector reviewed a Patel letter that identified training had been
. conducted for key personnel et Patel by the Director of Product Assurance.on Jurie 23, 1986. The scope of the training reflected on Section 4.0 of Patel's QA Manual relating to procurement document control. The four test programs reviewed by the NRC inspector on this visit demonstrated that the requirements of purchase orders to Patel's vendors were being followed.
3.
(Closed) Nonconformance (86-01-03) i Contrary to Criterion V of Appendix B to 10 CFR PART 50 and Section 3.1 of Quality Engineering Procedure 3.4, it could not be determined i
that the final conclusions in assessment reports PEI-1R-83-4-33, JN 8305-12 and PEI-TR-82-4-50, JN 8201 were sound and accurate.
The NRC inspector verified that both assessment reports for JN 0305-12 and JN 8201 has been revised to include a section relating to functional requirements and functional analysis. With the additional information provided, it was determined that the final conclusions in both assessment reports were sound and accurate. The qualifica-1 tion documentation analysis checklist format in the Patel EQ Assessment Report now contains a reference column. This reference column requires a definitive answer to each question and provides a cross check back to the qualification documentation. Patel Engineering Verifier Log Form No. 912 now contains mandantory check
)
points relating to performance requirements and testing results.
i E.
OTHER FINDINGS OR COMMENTS:
J 1.
Technical Evaluation - The NRC inspector performed a technical evalu-4 ation of four test programs for qualification testing of safety-related electrical equipment. The following table summarizes the job number (JN) files examined, including equipment type, plant &
types of documents examined.
3 Equipment Type Plant Documents Examined 8702 Limitorque wiring Cooper Test Report (TR). Test
& Okonite splices Procedure (TP), Purchase Orders (P0's) & Test data (TD), Certification test report (CTR) 8701 Sta-Kon connectors Cooper TR, TP, P0, TD, CTR 67
_ _ _ _ _ = _ _ -
.0 ORGANIZATION: PATEL ENTERPRISES, INCORPORATED HUhTSVILLE, ALABAtlA
. REPORT INSPECTION Nn t 44cnno11/67-/)1 RESULTS:-
PAGE 4 of 9 g
Equipment Type Plant Documents Examined 8628 Preamp enclosure Robinson TR, TP, P0, TD, CTR 8617.Patel conduit Seal Brunswick TR,-TP, P0, TD, CTR' The NRC inspector reviewed the EQ process prescribed in'each test-report / test procedure and reviewed test results for the above programs, including the bases for accelerated thermal acing and radiation for one program.
Four test programs were examined-to ensure that Patel had controlled the following as applicable:
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.
i c.
lest acceptance' criteria were established as described in the test specificatiori or in the design engineering documents, such as calculations and engineering letters to meet the requirements of IEEE-STD-323/1974.
d.
.Same, test assembly was used for all phases of testing and represented a standard production item.
c.
Environmental conditions were established and described (e.g.,
pressure arid temperature profiles, and thermal aging factors) and were consistent with those outlined in the test specification or test procedure, f.
Test results were adequately reduced and evaluated against established acceptance criteria described in customer test specifications or purchase orders.
g.
/,11 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 reports were properly documented.
t 68
i
! f ORGANIZATION: PATEL ENTERPRISES, INCORP0 FATED HUNTSVILLE, ALABAMA REPORT INSPECTION Nn - 400nnG11/A7-01 RESULTS:
PAGE 5 of 9 j.
Appropriate margirm were applied.
k.
Functional performance requirements were met.
No nonconformances were noted during this review. The following describes each program:
i a.
JN-8702
Purpose:
The purpose of this test program was to qualify natural aged limitorque torque switch cablir.g, limitorque natural aged motor lead wiring and okonite splices for Cooper Nuclear Station (CNS).
Equipment
Description:
The test assembly consisted of two cable /
I splice circuit. loops inside a NEMA 4 enclosure. The enclosure 1
did not have a drilled weep hole. Circuit Loop 1 consisted of two segments of torque switch cabling (black braid) and three
. bolted Okonite T35/695 tape splices. Both segments had been naturally aged in service at CNS for twelve years and had been installed in valve operators located in the steam tunnel. The environmental conditions in the steam tunnel consisted of a high radiation and elevated ambient temperature.
Circuit Loop 2 consisted of two segments of motor lead wire (yellow braid) and three bolted Okonite T35/#95 tape splices.
Both segments were naturally aged in service at CNS for twelve years and installed in a valve operator located in the steam tunnel. Environmental conditions were the same as circuit loop 1 above. The lead wires for both loops were Rockbestos l
Firewall SIS 16 AWG 600V nuclear cable.
Test specimen mounting / setup: Review of the Patel test report showed that the test assembly was installed in the test chamber with the conduit opening oriented downward. The test assembly was electrically energized and subjected to the steam accident simulation test.
Results: Approximately 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> into the steam accident test the insulation resistance ricasurements were below one megohn, and remained below one megohm for the remainder of the test for circuit loop 2.
At fifty three hours 50 minutes into the steam accident simulation test, leakage current for circuit loop 2 was measured at 671 ma.
This measurement did not meet the acceptance criteria for leakage current which states that the l
69
ORGANIZATION: PATEL ENTERPRISES, INCORPORATED HUNTSVILLE, ALABAMA REPORT INSPECTION nn - oq4nn491/R7-01 RESULTS:
PAGE 6 of 9 leakage current shall remain below 300 ma. However, subsequent nieasurements showed that the leakage currents returned to below 200 ma. Patel directed their vendor to deenergize circuit loop 2 at sixty hours 30 minutes into the accident simulation test.
)
l I
Visual inspection after the accident simulation test showed that I
both segments of the motor lead wire (yellow braid) were frayed.
However, one segment was damaged more extensively than the other.
The conclusion in Patel test report stated that the failure of the naturally aged limitorque power wiring.(circuit loop 2) to meet the defined functional requirements was the result of an interruption in the cable insulation due to unknown causes, b.
JN-8701:
Purpose:
The purpose of this test progran,was to demonstrate the operability of T&B Sta-Kon connectors when subjected to steam accident and functional simulations for CNS.
Equipment
Description:
Patel test report showed that the test assembly consisted of four Sta-Kon connector pairs (RA18-250F and 18RA-250T) inside a NEMA-4 enclosure, assembled and provided by CNS. The enclosure did not have a weep hole. The connecting wiring used in the test assembly was Rockbestos Firewsil SIS AWG 600 V Nuclear cable. The enclosure simulated the gasketed s
coridulets at CNS. Two specimens were insulated with 3/B " PVC tubing. The other two specimens were coated with Patel conformal coating and then covered with 3/8" PVC tubing.
Test Specimen flounting/ Setup:
The entire test assembly was ins Giled in the test chamber with the conduit opening oriented downward.
The specimens were than electrically energized and subjected to the steam accident simulation.
Results: The Sta-Kon connectors were successfully qualified under steam accident and functional simulations. All visual inspections implemented during the course of testing revealed no damage as discussed in the test report.
It was determined by the NRC inspector that all acceptance criteria had been met.
c.
JN-8628,:
Purpose:
The purpose of this test program was to demonstrate that the Patel preamp enclosure will protect a remote charge amplitier from environmental conditions inside containment at 70
j ORGANI2ATION: PATEL ENTERPRISES, INCORPORATED HUNTSVILLE, ALABAMA REPORT INSPECTI0h 44Q00Q31/87-01 RESULTS:
PAGE 7 of 9 NO all times of required operability of the preamplifier including j
t during and following a postulated design bases event.
Equipment
Description:
The Patsi preamp enclosure is'a sealed enclosure outfitted with a Patel conduit seal to provide the necessary sealing for the conduit opening into the enclosure.
The enclosure itself is sealed using a GRAF0ll gasket which is compressed by the enclosure cover via 1/4" bolts. The only non-metallic materials in the enclosure assembly are the i
GRAF0Il gasket and fiberglass ir.sulation.
Test Specimen Mounting / Setup: A hardline coax was covered with Raychem WCSF-050-N heat shrinkable tubing along its entire length inside the test chcaber and to a point immediately outside a second Patel conduit secl which was used to seal the chamber pressure. The crimp splices inside the enclo:;ure were covered with (from 3" to 5") of Raychem WCSF-070-N heat shrinkable tubing to provide suitable electrical insulation.
l Results: Visual inspection after steam enviroment and chemical spray revealed extensive damage to the paint on the outer surface of the enclosure, but there was no indicaticn of noisture inside the enclosure or on the fiberglass insulation as indicated by Patel test report. Although the enclosure internal temperature did exceed the normal operating temperature during the testing the enclosure still permitted the preamp to remain operable by j
protecting it from the pressure transients and the borated l
water spray.
d.
JN-8617:
l
Purpose:
The purpose of this test program was to extend the qualified life of the Patel Conduit Seal for use inside contain-ment at the Brunswick Steam Electric Plant and to demonstrate reusability of the conduit seal grommets for use in the recctor building.
Equipment
Description:
The Patel conduit seal is a multi-cable mechanical interface that may be used to seal existing conduits l
and lead wires. The conduit seal will accept up to six (6) #12 AWG or #14 AWG cables, or up to eight (8) #16 AWG or #18 AWG cables by use of installed grommets.
71 i
i 1
ORGANIZATION:
PATEL ENTERPRISES INCORPORATED HUNTSVILLE, ALABAMA 1
REPORT IhSPECTION Mn - or 2no11/R7-01 RESULTS:
PAGE 8 of 9 Thermal Aging:
Every twenty four hours which simulates five years of service the seal coupling nut was loosened, the groraraet removed from housing, re-greased and retorqued, pressure boundry test performed and the assembly returned to the thermal aging oven until the aging goal was reached. This was performed on the four specimens to demonstrate reusability of the grommet.
Test Speciraen Mounting / Setup: The test specimens were mounted verticaily to penetrate the test chamber wall through 3/4" NPT threads. The test speciraens was than subjected to steam accident simulation test.
Results:
Post Accident visual inspecton revealed that the housings was discolored and cirty from the boiler chemical residue and pipe corrosion products frora the LOCA exposure. No damage was observed which could affect the operation of the conduit seals. The conduit seals met all functional requirements prior to and throughout the accident simulation.
It was further demonstrated that the qualified life of the conduit seals raay be extended from previous qualification levels and that grommets for use in the reactor building may be removed and replaced as many as seven times during their qualified life.
2.
Development of Qualification and Assessment Reports:
The NRC inspector reviewed Patels' Quality Assurance Manual and Quality Assurance Engineering procedures to determine what controls were established to assure proper development _of qualification packages and assessment reports. Six QA engineering procedures were reviewed that addressed the above topic and appeared to be adequate to raaintain control of engineering documents.
Thc NRC inspector revieweo a document titled "EQ program for Robinson" dated January 25, 1965. The purpose of this document was to present the philosopy and raethodolgy used to perf orm and document the EQ of the electrical equipment classified as important to safety at HBR-2.
Procedure PEI-TR-8310ll-01 titled " Procedure for review of EQ of each indivioual piece of safety-related electrical equipment utilized at the H.B. Robinson Nuclear Plant" dated September 11, 1984 was a part of the above docment which prescribed the raethodolgy of preparing EQ documentation. Topics included in this procedure were: 1) equipment /
nameplate data, 2) environmental Category, 3) safety related function,
- 4) EQ system component evalubtion worksheet, 5) qualification 72
ORGANIZATION: PATEL ENTERPRISES, INCORP0 RATED HUNTSVILLE, ALABNIA REPORT INSPECTION Nn - C40nn091/Rb n1 RFSULTS:
PAGF Q nf 4 limitations, 6) operations, 7) calibrations and maintenance procedures, 8) conclusions, 9) seismic qualification evaluation sheet and 10) quality assurance in accordance with Patel engineering QA program. Also an external interface document No. 5 Rev 3 dated October 19, 1984 between Patel Enterprises, Inc. & h. B. Robinson steam electric plant was reviewed. The purpose of the external interface document was to identify the interface, scope of responsi-bilities, lines of communication and document control methods between outside engineering, consulting and A/E's and the~ Nuclear Engineering & Licensing Department (MELD).
It was determined that Patels line of communication involving EQ was through NELD for design information and that interface between Patel's QA organiza-tion and CP&L's QA/QC organization need not involve NELD.
3.
QA Program Implementation Review:
The NRC inspector reviewed four JN files containing documentation which supported testing efforts. Documentation included purchase orders, EQ test procedures / reports, certification test reports and test cata.
It was determined that Patel was implementing their QA program in accordance with their QA Manual.
4.
QA Manual Review:
The NRC inspector reviewed four changes that were incorporated into revision G of the QA Manual. It was determined that the changes were minor and did not impact Patel's QA program with respect to the requirements of Appendix B to 10 CFR Part 50.
F.
PERSONS CONTACTED
- Dr. Patel, President, Patel
- T.Utech, Director Product Assurance, Patel G. Elau, Program Manager, Patel F. Roy, Project Engineer, Patel
- J. Jenkins, Vice President Nuclear Engineering, Patel
- Attended Exit Meeting.
- Separate Exit Meeting.
73
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ORGANIZATION: POSI-SEAL INTERNATIONAL
)
NORTH STONINGTON, CONNECTICUT REPORT INSPECT 10ll LNSPECTION NO.: 99900886/87-01 UATE: 07/06-10/87
)N-SITE HOURS!
64 CORRESP0fiDEhCE ADDRESS:
Posi-Scal International, Inc.
ATTH: Mr. Peter E. Arnold General Manager Route 49 and U.S. 95 North Stonington, Connecticut 06359 ORGAtlIZATIONAL CONTACT: Robert D. Barry, Manager of Quality
(' O3) 599-1140 TELEPHONE NUMBER:
d NUCLEAR INDUSTRY ACTIVITY: Manufacture butterfly valves.
4 ASSIGliED INSPECTOR:
IWA 6MM6 b-U -S'l l
J.F. Conway, Program Development and Reactive Date I
f,nspection Section (PDRVS) i OTHER IkSPECTOR(S):
C. Abbate, PDRIS 1
APPROVED BY:
.i &
_M Il/
7 J.StonefChief,PDRIS,~VendorInspectionBranch a
i 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 an allegation pertaining l
to the removal of hold tags from nuclear valves by a production manager.
l 1
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PLANT SITE APPLICABILITY:
Palo Verde Units 1, 2, 3 (50-528, 529 and 530).
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ORGANIZATION: POSI-SEAL INTERNATIONAL c?'
]
NORTH STONINGTON, CONNECTICUT L
I' REPORT-INSPECTION d ',
NO.?
494nnRA6/87-01
'RESULTS:
PArJ 7 M in A.
VIOLATIONS:
+i 1
Contrary to Section 21.31 of 10'CFR Part.21,'a review of documentation L
packages (DP) and purchase orders (PO) for butterfly valves fabricated under Section III of the ASME Code, revealed that while 10 CFR Part 21 was imposed on Posi-Seal International (PSI), PSI did not specify that y'
10 CFR Part 21 requirements.would apply on P0s H9657 (September 27,,1983),
l
'J0446 (November 4, 1983), and J8593 (January 29, 1985) to ABC0 Welding l
and Industrial Supply Comp (any; 12261 (April 1,' 1987) to Adironadack Steel Castings; G7646 and 7647 January 15, 1962), K0988 (June 18, 1965) and
'11448 (February 19,1987) to Congdon and Carpenter Company; 13035 (May 26, 1987) and 13076 (May 28, i987) to Empire Steel Castings; K2060 (July 29, 1985) to Hy-Temp; 12215 (March 31, 1987) to Jessop Steel; G7459 (January 6, 1982) to Lebanon Steel Foundry; 13097 (May 29,' 1987) to NQS; 12605 (April 24,.1987) to Pressure Vessel-Nuclear Steels; 11449 (February 19, 1987) to Quaker Alloy; and 12208 (March 30, 1987) to Texas Bolt.
l (87-01-01) l B.
NONCONFORMANCES:
1.
Contrary to Criterion IX of Appendix b to 10 CFR Part 50 and e
Sections 12.30 and 12.40 of the Quality Assurance Program Manual
{
4d (QAPM),itwasnotedthatthreeNDEprocedures-Nos.13S-HAR-UT-2f4.'
23.A.200-1986, and 23.A 031-1980 - used by Magnaflux/MQS Inspection
)
in February 1984 and June 1987 were not reviewed by PSI.
i (87-01-02) 2.
Contrary to Criterion IX of Appendix B to 10 CFR Part 60, Section
~
12.40C of the QAPM, and Section 9.6.1 of SNT-TC-1A, it was noted t
that PSI did not have the qualification records for four individuals from Magnaflux/MQS Inspection who had performed NDE in February 1984, February 1986, and June 1987.
(87-01-03) 3.
Contrary to subsection NCA-4134.4 of Section III-of the ASME Code j
and Criterion IV of Appendix B to 10 CFR Part 50, a review of P0s revealed that quality requirements were not; passed on to vendors for the following P0s:
11656 andzl1659 (March 2, 1987) to Airmotive Calibration Corporation; 3689 (April 24, 1979), 3735 (June 4, 1979) and 4857 (July 30, 1979) to Cambridge Plating; and 3692 (April 27, 1979) to Fountain Plating.
(87-01-04).
4.
ContrarytosubsectionhCA-4134.5of'.SectionIIIoftheASMECode and Criterion V of Appendix B to 10 QFR Part 50, written instructions or procedures did notrekist for the use of " hold tags";
on nucleer components.
(87-0)-00)I j
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'0 ORGANIZATION:
NORTHSTONINKi0ft, CONNECTICUT IT ?
. REPORT..
INS'PECTION h
HO.-
99900AM /87-01 '
RESULTS:
PAGE 3 of 10 C.
UNRESQLVED JTEMS:
~
I" y <, :,
r During the.ipspection, it was noted that 4-24" trotor operated b,utterf4y
'vab es were returned to PSI from the,Palo Verde nuclear site. -lhe vahes were returned in 1985 because of deterioration of the electroless nickel plating en the valve discs. A meeting was held on June 25,'1985
~
h.
withCombustionEngpneering(CE)personnelpertain$gtothisproblem..
PSI was unable to furnish documentation of the raeeting and the cause
,m 7
for the corrosive attack.cf the nickel plating. Accordingly, this item willremainunresolpeduntilitis'discussedwithCEandBechtel during a futuredGC inspection of their facilities.
s 1
D.
STATUS OF PREVIOUS IhSPECTION FINDINGS:
s}/
None.'Thisd[in'efirstinspeltionatthisfacility.
E.
OTHER FIliDINGS AND COW 1ENTSb
[
1.
Allegation 1
f' O
7p April 1987, the Region I office of the NRC was notified of an allegation pertaining to the rerncval of hold tags from nuclear xvalverg b order to get the valves shipped.
I.
The use of/h61d tags and other manufacturing. controls were q
reviewed 'piiryp the inspection. Section 9.30 uf PSI's QAPM y,
.,jf states thdt Wavelers are the official status focicators for g
,j nuclear gnmponehts and all other tags, etc. 6re used for visual l
aid on1 H The traveler is the control mechanism for fabrica-4 l
< tion in that the next operation on the. traveler cannot be perforr)ed until the previous operation has been signed off by tife QC inspector. All operations must be signed by QC.
.s L
Hole tags are used on nuclear components as a visuql aid to f',t y
W$n personnel that work is not to be performed orl that compel:ent until the hold is released.'. Hold tags are placed on cepponents which do not meet acceptance criteria.. The tags
,, d have two parts. One part is put on the component as a visual i
aid while the other part is kept by(Tt until the problem is i
5
'l dispositioned. After dispositioning both pris of the tag are destroyed.
Inpiditjon to the travelers, QC sign-offs, and' hold tags, PSI f the traveler er.d QA (;ersonnel perform a finatr,eviev.)hcid on the 6cponent is inspection reports before a compute /
released.
This is done in crder to assure that all cperations
{
on the t7aveler have been pe,rforraed. Until QA completes its 1
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Ng^ANIZATION: POSL-$EAL TEMPqIONAL-y (NORTiQ STON!srRON CONNECTICUT T
a 1
REPORW,',
' INSPECTION
_ h g ggj6/87-01 RESUMS:
' PAGE 4 of 10 N
1W reviev eM; releases the cenuter hold, a bill of lading or
, !c l shippir3 h pers cannot be printed; and the valve cannot be-i "N ::
shippedd' c
1
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}
The inspectors wa,e unable to determine if hold tags were removed from nucicar Salves or components. However, due to the additional controls ?hUhss implemented, a valve could not'be p
shipped saless all helds tave been officially released by QA or
)
QC. Subsequently, the allegation could not be substantiated. '
2.
Nondestructive Examination (NDE)
The NRC inspector reviewed Section 12 of the QAPM, six KDE procedures, and the' qualification records for nine NDE personnel.
Procedure QCS-001 " Qual.ification and Certification of Nonde-i structive Examination Personnel" documents PSI's qualification /-
certificatiorJprogram. A June 1985 letter from PSI's Manager y
ofTQualitytdFisherControls(FC)appointedtwoindividuels i
,,from FC as Lel vel III examiners for PSI. FC's written practice, s
procedure No, FMP 2J1 "American Society for Nondestructive j
Testing Perstonel Qualification and Certification" was stamped QA approved and r{gned by PSI's Manager of Ouhiity. '"<
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It we's tfited that hagnaflux/MQS Inspection performed ultrasonic 3
'(UT) radiographic (RT),'and liquid penetrant (PT) testing for j
PSI. HQS personnel E. Toper performed PT and R. Boswell i
b'
\\
performed RT in June 1987 on components for Palo Ve?de (reference i,ipb%6.39620h. UT was performed in February 1984 by D. Trott on stems fcDaver Valley (reference job No. 20011). For Clinton a
(referenceNod No. 36190), R. Thompson performd RT on discs in February 1986. -PSI did not have copies'of the qualification records fdr the four examiners identified above (sco Noncon-f
, s\\
formance 87-01-03). A review cf three procedures from MQS and used for the NDE testing indicated that PSI aid not review and approve the byoced.phs (the Noncontormance bliCl-02). The procedures vere ho.13S-HAR-UT-09 " Ultrasonic to ASME Section III" <
dated February 22,\\l979; No. 23,A.200-1986 " Penetrant Examination Fluorescent Dye' General Requirements" dated October 27, 1986; and No. 23.A.031-19GG " Penetrant Eximinacion" dated January 21, 1987.
In general, the type of inforndtMn found in the record files for I:
the seven HDE personnel (five - Level II PT and two-Level 1 PT) s from PSI included t.d.ucational background, resumes, cw t.lficates of training completed, record of qualification, certification statercents,
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ORGANIZATION: POSI-SEAL INTERNATIONAL l
, NORTH STONINGTON, CONNECTICUT i
p REPORT INSPECTION l
NO.-
99900886/87-01 RESULTS:
PAGE 5 of 10 examinations (general, specific, practical) and eye exams.
The qualification records appeared to satisfy the requirements of 0W SNT-1C-1A. A review of the NDE procedures at the PT work station indicated that the current procedures were in the manual.
3.
Calibration of Measuring and Test Equipment (H&TE) e The NRC inspector reviewed Section 16 of the QAPM, standard instructions,. calibration schedule cards, and certifications for
/
reference standards used by service vendors to calibrate M&TE.
An, observation of M&TE at various work stations was also per-formed to assure that M&TE are properly identified, controlled and calibrated at specified intervals. The MATE selected for review include the following: a thermometer, two ampmeters, two voltmeters, five verniers, 10 micrometers, 15 pressure gages, two gage block sets, and a deadweight tester.
All the M&TE was-in current calibration and a sticker showing the calibration'due date was attached to each item. Airmotive Calibration Corporation was the vendor that calibrated the two-gage block sets and the weights for the deadweight tester. The l
pressure gages are calibrated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to testing and after each test or. series of tests not to exceed two weeks.
1,
'It was verified that the pressure gages used on testing of nuclear valves had been calibrated prior to and after each test l
l-(four in 1986 and three in 1987). The ampmeter and voltmeter l
on the two " Miller syncrowave 500" welding machines were f
checked, and all four instruments were in current calibration.
4.
Training i
Training records for four years were reviewed. The records l
i consisted of a training folder which included a general training matrix, specific training matrices for each department
.i i
of PSI, and training course attendance sheets.
The records and l
training were in accordance with Section 18 of the QAPM. The general matrix identified positions in PSI and the sections of i
the QAPM to which employees were to be trained. The specific l
departmental matrices identified individuals and the sections of the QAPM to which the individual was trained. The j'
attendance records identified the section of the QAPM that was 4
being covered by the training, the date, the instructor and the employees who attended the training. Employees were trained to their specific section(s) of the QAPM each time a new revision i
of the manual was issued.
L l
79 l
[
j) 0 ORGANIZATION: POSI-SEAL INTERNATIONAL NORTH STONINGTON, C0hhECTICUT REPORT INSPECTION RESULTS:
PAGE 6 of 10
~
NO.-
99900826/87-01 5.
Manufacturing Controls During the inspection, manufacturing controls were reviewed by observing the work being performed in the shop and by a documentation / procedure review. Areas examined-included shop travelers, material identification and traceability, in-process and final inspection, and inspection personnel qualifications.
In-process and completed travelers for nuclear work were reviewed. The travelers identify the operations to be executed, QA/QC inspections, and maintain material identity and trace-1 ability. The travelers which were reviewed were complete and in accordance with QAPM Section 8, " Material Control and Traceability" and Section 9, " Shop Travelers for Manufacturing Assembly.and Testing." The inspections had been performed when required by qualified QC personnel. As outlined in QCS-008,
" Certification of Inspection and Test Personnel," dated November 3, 1986, the inspectors were identified on Certifi-cation Record, individual certification records were present and'it was verified that the QC stamps used on the travelers corresponded to qualified inspectors.
During the review it was noted that hold tags are used on both nuclear and non-nuclear work. A description of how the hold tags were used was given by PSI personne b but in examining the hold tags it was noted that a hard copy of a hola tag was in the possession of QC not the soft copy as stated. Although 3
section 14.5.0 of the standard Quality Prograni describes the use of hold tags on non-nuclear work, written instruction or procedure did not exist for the use of hold tags on nuclear work (see Nonconformance 87-01-05).
6.
Welding The NRC inspector reviewed Section 10.0 of the QAPM pertaining to welding procedure specifications (WPS), procedure qualifi-cation records (PQR), welding material control and welder qualifications.
The Welding Engineer is responsible for pre-paring WPSs and certifying the results of tests / examinations on PQRs. A distribution list for each WPS is maintained by the Welding Engineer who currently is also the Supervisor of Engineering.
In the weld area which ccnsists of two welding booths, book No. 6 containing WPSs was reviewed, and eight WPSs were compared with master copies retained by the Welding 80
ORGANIZATION: POSI-SEAL INTERNATIONAL N0RTH STONINGTON, CONNECTICUT i
REPORT INSPECTION
'l NO ?
99900RR6/67-01 RESULTS:
PAGE 7 of'10
-l Engineer to assure that current procedures were at the work I
station. WPSs in book No. 2 (QA Supervisor) and book No.-1 (Manager of Quality) were also reviewed. The reviewed WPSs in all three books were in agreement with the master copy. The eight WPSs along with supporting PQRs included the following:
'No. 1-01-1 " Manual Shielded Metal Arc Welding," Revision 0; No.
1-01-1-SR, " Manual Shielded Metal Arc Welding of P-No. 1 Carbon Steel," Revision 0; No. 1-04-NUC " Manual Gas Tungsten Arc Welding of P-No. 1 Carbon Steel," Revision 4; No. 1CV-04 " Manual Gas Tungsten Arc Welding of Impact Tested P-No. 1 Grade 1 and 2 Carbon Steel," Revision 2; No. 5F71-04-6 " Manual Gas Tungsten Arc Hard-Facing Weld Metal Overlay of P-No. 5 Base Metal," Revision 0; No. 8-01/04-01 " Manual Shielded Metal Arc / Gas Tungsten Arc Welding of P-No. 8 Stainless Steel,"
Revision 2; No. 8-04-1-SA " Manual Gas Tungsten Arc Welding of P-Ho, 8 Stainless Steel," Revision 0; and No. BCV-03-01
" Automatic Submerged Arc Welding of P-No. 8 Group No. 1 Stainless Steel." Welding on nuclear valves was not being performed during the inspection.
The welder qualification test (WQT) records for four welders-were reviewed.
It was noted that the four welders had been qualified to four WPSs (i.e., 1-04-NUC, E01R10, 8-04-1 and E08R10) which were used on past nuclear orders. The certifi-cation of the WQTs indicated that the welds had satisfactorily passed a guided-bend, tensile, and/or charpy impact tests in accordance with Section IX of the ASME Code. Testing of j
coupons was conducted by Bridgeport Testing or Dirats Testing-Laboratory.
(
l In the weld material storage area, low hydrogen covered elect-rodes were stored in a padlocked oven which was operating at l
250 F.
A metal cabinet which was also secured with a lock i
contained the stainless steel weld material. The QC Technician J
is responsible for storing and issuing welding materials.
]
Control of material is maintained by heat / lot numbers. The j
welding gases were stored in an area specifically designated for such storage.
The weld history log, which is maintained by the Welding
)
Engineer, was reviewed for the time period 1983 to the present.
]
The log is updated on a monthly basis and shows the dates that one of three welders performed welding to four processes (i.e,
)
GTAU, SMAW, SAU and GMAW). A review of the nuclear weld j
material list (NWML) in the welding material storage area j
l 81 1
ORGANIZATION: POSI-SEAL INTERNATIONAL NORTH STONINGTON, CONNECTICUT j
REPORT INSPECTION NO.: 99900886/87-01 RESULTS:
PAGE 8 of 10 identified 11 reference Nos. (i.e., nuclear jobs currently in j
house) in which the weld material was purchased on three P0s.
1 A review of the NWML in the possession of the Welding Engineer I
agreed with the reference Nos. and heat Nos. on the listing j
posted in the weld material storage area.
7.
Procurement The inspectors examined the procurement process at PSI. This included a review of Section 7 of the QAPM Quality Control i
Standard (QCS) 015, " Quality Assurance Approved Vendor List; Vendor Survey, Audit, Evaluation," specifications and POs from PSI customers, documentation packages, the PSI Approved Vendor i
List (AVL) and P0s to vendors on the AVL.
1 i
P0s to material suppliers / manufacturers, calibration vendors, NDE vendor, plating vendors and heat treating vendors were reviewed to assure that applicable technical and QA program requirements were included or referenced in the'P0s. The requirements of 10 CFR Part 21 were imposed on vendors in two cases. These two P0s were for limit switches. The P0s which did not invoke the requirements of 10 CFR Part 21 include three P0s for weld wire - US593, J0446 and H9657; twelve P0s te material suppliers / manufacturers - K0988, G7459, G7646, G7647, 11448, 11449, 12208, 12215, IEE61, 12605, 13035, and 13076; one PO for heat treating - K2060; and one P0 for NDE Services -
13097 (see Vialation 87-01-01).
Additionally, the requirement that a QA Program be specified in procurement documents was not included in six P0s to two plating vendors and one calibration vendor. These P0s include 3669, 3692, 3735 and 4851 to plating vendors, and 11658 and l
11659 for calibration services (see Nonconformance 87-01-04).
8.
Documentation Packages (DP)
Sixteen DPs for Section III, Class 2 and 3 butterfly valves were reviewed. Depending on the customer, the DP consisted of the following documents: QA Records Validation signed by the Nanager of Quality, Document Index, /sssembly Drawing, Material Traceability List, certified Material Test Reports (CMTR),
Certificate of Compliance (C of C), weld history (i.e., casting repair), wall thickness measurement reports, PT and RT reports, cleaning report, final test reports, and an ASME Code Data p
Report. The Assembly Drawing details individual parts by 82
J J
ORGANIZATION: POSI-SEAL INTERNATIONAL NORTH STONINGTON, CONNECTICUT 1
REPORT INSPECTION HO.-
99900RR6/87-01 RESULTS:
PAGF 4 nf 10 number and contains general notes (e.g., stamping and heat i
treat requirements), QA notes (NDE and welding procedures), and required nuclear procedures (body, disc and seat tests; cicaning; painting, and packaging). The CMTRs were for the bcdy, disc, gasket retainer plate, gasket retainer bolts, weld material, stems and disc pins. PSI's CMTR certifies that all tests and operations performed by PSI or their subcontractors for the body, disc, gasket retaining and retainer bolts, welding filler metal, and leak-off pipe are in compliance with the requirements of Section II, III and IX of the ASME Code.
In addition, the valve stems and disc pins were procured and manuf actured in accordance with a QA program which meets the requirements of Appendix B to 10 CFR Part 50. The CMTR is signed by the Manager of Quality.
t The 16 nuclear customers designated by a PSI reference No. and year include tqe following: 27813~('81) Toledo Edison / Davis Besse, 28988 ('82) UE&C/Seabrooke,_30611 ('83) Westinghouse /
Beaver Vallej, 32436 ('83) Cleveland Electric Illuminating /
Perry, 34836 i'84) Gibbs & Hill /Susquehanna, 35685 ('85)
Carolina Power & Light /H. B. Robinson, 35910 ('85) TVA/
Sequoyah, 36190 ('85) Illinois Power /Clinton, 38216 ('86) Duke Power / Catawba, 38605 ('86) TVA/ Watts Bar, 39838 ('86) Illinois Power /Clinton, 39620 ('86) Bechtel/Palo Verde, 39156 ('86)
UE&C/Seabrook, 40599 ('87) Duke Power /McGuire, 40846 ('87) Duke Power / Catawba, and 40933 ('67) Illinois Power /Clinton.
9.
Audits The NRC inspectors reviewed the internal audits for 1986 and part of 1987. The 19 criteria outlined in PSI's QAPM were f
covered in the audits. Each audit included an audit plan, j
cudit checklist, and audit report. Audit findings were documented in Corrective Action Reports (CAR), and deficiencies were corrected in a timely manner. The audits which were reviewed were performed in accordance with the criteria outlined in Section 17 of the PSI QAPH.
J Ten external audits were reviewed during the inspection. The j
auditing o# vendors is performed using both Section 17 of the QAPM "
p lity Control Standard (QCS) 015, " Quality Assurance l
Apr"
. Vendor List; Vendor Survey, Audit, Evaluation," dated 1
C.._er 16, 1984. The records which were reviewed included audit checklists and reports and were complete and demonstrated L
that PSI is performing vendor audits in accordance with their l
procedures.
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83 f.
ORGANIZATION: POSI-SEAL INTERNATIONAL NORTH STONINGTON, CONNECTICUT l-REPORT INSPECTION fJn - 444nnRRA/H7-01 RESul TS:
PAGE 10 of 10 10.
Plant Tour The inspector toured the PSI manufacturing facility at various times during the inspection in the company of PSI officials.
l Items witnessed included receipt inspection, component storage I
area, weld material storage area, machining, assembly, welding I
area, testing, final inspection, PT area, cicaning, painting, l
and packaging.
l l
F.
PERSONNEL CONTACTED:
l 1
- M. Shannon, North American Counsel l
- K. Gadaree, Accounting Manager I
- R. t. Walter, Director of Operations i
- R. J. Stitel, Manager Nuclear Sales l
- S. J. Baker, Materials Manager l
- D. S. Quirk, Applications Engineer I
- E. W. Good, Manufacturing Manager
- D. Mashi, PIP Manager
- B. S. Nichols, Quality Assurance Engineer l
- F. D. Riccioli, Marketing Manager
- F. Canastar, Manager Human Resources l
- J. Corey, Manager Engineering
- P. E. Arnold, General Manager
- J. Azzinaro, Director, Marketing and Sales
- E. W. Whitford, Facilities Engineering Supervisor
- R. D. Barry, Manager of Quality J. Jones, Lead Inspector E. Banville, Shipping Clerk N. Place, QC Technician J. Rodgers, Welding Engineer
- Attended exit meeting.
84
3 ORGANIZATION: POWERPLANT SPECIALISTS, INCORPORATED COSTA MESA, CALIFORNIA REPORT INSPECTION INSPECTION NO.:
99901086/87-01 DATES: 5/19-20/P7 nN 9TTF HnHPR-A
[
CORRESPONDENCE ADDRESS: Powerplant Specialists, Inc.
ATTh: Mr. Richard G. Engel President 666 W. Baker Street Suite 111 Costa Mesa, California 92626 ORGANIZATIONAL CONTACT: Richard G. Engel TELEPHONE l1 UMBER:
( 7141 557-26M NUCLEAR INDUSTRY ACTIVITY: MSR internal modifications at North Anna, Surry, and Sequoyah in 1984.
ASSIGNED INSPECTOR:
h6 6W W 7-//-f[
JAnbs T. Conway, Progrq Development and Reactive Date
[1spection Section (PDJIS)
APPROVED BY:
c*'a
/
D Jamds C. Stone, Chief, PDRIS, Vendor Inspection Branch Date IhSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR Part 50, Appendix B and 10 CFR Part 21.
B.
SCOPE: The purpose of this inspection was to review the qualification of welders utilized in the modifications performed Ly Powerplant Specialists (PS) of MSR internals at Virginia Power Company's (VPC) North Anna and Surry nuclear facilities.
PLAllT SITE APPLICABILITY:
ficrth Anna 1 and 2 (50-338/339) and Surry 1 and 2 (50-280/ES1).
85
m ORGANIZATION:
POWERPLANT SPECIALISTS, INCORPORATED COSTA MESA, CALIFORNIA REPORT INSPECTION J n-99001nR6/A7-01 RFStfl TS i PAnF 9 nf'a A.
VIOLATIONS.
None.
B.
NONCONFORMANCES:
j None.
C.
UNRESOLVED ITEMS:
None.
j D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
None.
This was the first inspection at this facility.
E.-
OTHER FINDINGS AND COMMENTS:
1.
Quality Control (QC) Manual Various sections of PS's QC Manual, Revision 2, dated August 6, 1986 were reviewed.. The QC Manual contains a description of the QA program which is used for field assembly and field repair of power boilers and field fabrication and installation of power piping in accordance with Section I of the ASME Code and ANSI 831.1. The manual is also intended to meet the requirements of the National Board Inspection Code, and the Jurisdictional Requirements for the repair of power boilers and piping and pressure vessels, lt was noted that welders are qualified to Section IX of the ASME Code.
2.
Moisture Separator Reheater (MSR) Modifications - VPC In discussions with the President of PS, it was learned that PS was a silent partner with Southwestern Engineering Company (SEC) who manufactured and tested the MSRs at their Joplin, Missouri facility. The hSRs for the North Anna and Surry nuclear facilities were fabricated to the requirements of Section VIII of the ASME Code.
During a review of records at SEC corporate offices in Los Angeles, California, it was noted on purchase order No. 56076 dated March 21,1985 ano No. 54929 dated May 1,1984 that SEC requested PS to furnish all labor and equipinent to install MSR internals at Surry and North Anna, respectively. All the welding 86
1 ORGANIZATION: POWERPLANT SPECIALISTS, INCORPORATED COSTA MESA, CALIFORNIA J
l REPORT INSPECTION NO.-
QQQn1nA6/87-n1 RFStHTS.
PArJ 9 nf IJ was to be performed to qualified welding procedures furnished by SEC.
PS was responsible for qualifying the welders to the require-
. ments of Section IX of the ASME Code.
The. NRC inspector reviewed welder qualification records for approximately 40 individuals who performed work at North Anna and/or Surry..The Welder Qualification Tests (WQT) indicated that the individuals were qualified to from one to four welding procedurespecifications(WPS). Three of the WPSs (1-1-SMA-G1, 1-8-SMA-G1, and 8-8-SMA-G1) along with Procedure Qualification Records, which were all on SCE letterhead, were reviewed. PS could not furnish a copy of the fourth.WPS, which was No. 1F01-G1.
All the WQTs were on SEC letterhead and signed by G. A. Johanson or G. E. Harmon, who were employees of PS. Approximately 15 of the WQTs did not contain an identification number, letter, or. symbol as required by paragraph QW-301.3 of Section IX of the ASME Code.-
PS utilized Triple-A Testing Laboratories (TTL) located in Paramount, California to qualify approximately 17 welders via the guided-bend tests, per paragraph QW-462.2 of Section IX of the ASME Code. A review of all WQTs (Form QW-484 from Section IX) at the TTL facility confirmed that the 17 welders were certified by TTL in May, July, August, September, and October 1984.
Documentation was not available at PS to indicate that PS had surveyed or audited TTL to verify that the laboratory was qualified to test welders in accordance with'
.I l..
Section IX of the ASME Code.
J The inspector had earlier received a sample of welder qualifications i
for 12 individuals from VPC. Where copies of two WQTs on SEC
]
letterhead were reviewed at TTL, it was noted that PS did not have i
copies for five of the individuals (B. Cedarholm, H. Poe, R. Marshall l
Hall, B. Phillippe, and P. McGraw) nor for a G Dorame.
On 21 WQTs the certification indicated that the welder had satis-f factorily passed a guided-bend test (QW-462.2) conducted by SEC, l
but there was no laboratory test number on the WQT, nor supporting i
documentation to verify that the testing had been performed.
The l
affected welders included D. Ackley, C. Barrings, C. Beverly, D. Burt, D. Dooley, A. Engle, D. Gaudin, M. Gomez, C. Green, C. Hood, S. Iwanski, C. Logan, J. Lumbattis, J. McDaniel, J. McLacklin, E. McWhorter, R. Owens, R. Rice, T. Segundo, P. Toomes, and
-G. Washburn.
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ORGANIZATION:
POWERPLANT_ SPECIALISTS, INCORPORATED COSTA MESA, CALIFORNIA
. j
' REPORT INSPECTION i
NO.*
999010R6/A7-01' kFSill TS -
PACF a nf e i
In addition,10 WQTs' certified that the welder was qualified by the alternate method of radiographic testing (RT), in accordance i
with'QW-304 and QW-305. However, PS did not have supporting documentation to verify that RT had been performed either on a test coupon or on an-initial proauction weld.. The affected j
welders in this case included E. Andrade,'b. Gilbert, M.~Gomez, D. Houston, K.'Johansen, E. Jones, J. Lumbattis, K. Mason, R. Mundale, and G. Washburn.
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, ORGANIZATION: sPRECISION HEAT TREATING COMPANY HONTEBELLO, CALIFORNIA ~
REPORT INSPECTION
. INSPECTION h0.: 99901096/87-01 DATES:
05/24-29/82_
OM-SITF Wrm.
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- CORRESPONDENCE ADDRESS: Precision Heat Treating Corpany ATTN:. Brian Launder- ~
Vice President 1532 South Greenwood Avenue i
i Mor.tebello, California 90640 ORGANIZATIONAL.CONTACl:. James Lastra, Quality Control 1 tanager TELEPHONE-NUhbER:
213-724-4880 NUCLEAR IN0051hY_ ACTIVITY: Heat treating.
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.k d-7!2 Qn ASSIGNED.lNSPECTOR:
J. C.\\ harper, Prograb Development & Reactive Uat/e
-l l
Inspection Section (PDRIS)
OTHER IhSPECTOR(S):
ll APPROVED BY:
Ap
[,.
9 jf7 Y
J. C//Stonc,. ChMf', PDRIS, Vendor Inspection Bronch Date JNSPECTION BASES AND SCOPE:
A.
BASES:
B.
. SCOPE:
The inspection was performed to review and evaluate the adequacy 1
l of the heet treating practices of Precision Heat Treating ~ Company (PHT)..
.PHT was a contructor to Cardinal Industrial Products Corporation (CIPC -
j now defunct). CIPC supplied fasteners hcVing mechanical properties lower
.than allowed by material specification.
PLAhT S'iTE APPLICABILITY:
50-456; 50-457 (Braidwood Station Units 1 and 2).
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ORGANIZATION: PhECIS10ti HEAT TREATIhG CCisPAllY I
K0hTEbELLO, CALIFORflIA REPORT IllSPEC110h NO.: 99901096/87-01:
RESULTS:
PAGE 2 of 7 I
A.
VIOLATIONS:
1.
Contrary to Section 21.11 of 10 CFR Part 21, Phl 1ailec to adopt l
c Part il procedure for evaluating and reportir.g deviations.
i (87-01-01)
B.
h0hC0hFORMANCES:
1.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and paragraph 5.3 of the PHT QC Manual, PHT failed to irdicate tiraes fcr charging and removing heat treatec material from the furnace on the furnace charts.
(67-01-02)
C.
OTHER FINDINGS OR COMMENTS:
Background and Conclusions
.Or. May 18-22, 1987, the NRC inspector reviewed and evaluated records at Cardinal Industrial Prcducts Incorporated (CIPI) pertaining tc the heet treatnent of fasteners supplied by the now defunct Cardinal Industrial Product Corporation'(CIPC). Previous Vendor Branch inspections had identified r.urrerous prcble. ras with the CIPC quality systerr.
CIPC supplied Ccrmonwealth Edison Con.pany (CECO) with fasteners having mech 6nical properties icwcr than those allowed by the American Society of Mechanical Enginccrs Boiler and Pressure Vessel Ccde (ASl4E Code).
'Upon review of the CEC 0 report dated July 1, 1986, " System Materials Analysis Department Report en Testing of Solting Materish for Braidwood
)
Station," the report concluded that heat treating probleros existed with j
the fasteners havirig low mechanical prcperties. CECO's testing prcgram censisted cf examining a total cf 23 SA193 Grade-B7 or B16 san.ples from five different heets of material. All of the samples tcsted t..et their i
requireo chernical content with the t.xception of sample numbers 17 and 19 which were.05 and.02 per cent, respectinly, above the spcified range for manganese.
Sample 21 frcm heat E-2702 was substantial) below the required Ultimate Tersile Strength (UTS) end Yicld Strer.sth (YS) for a grade b7 fastener. Samples 21 and 23 frcm heat 2-2702, both had marginal YS for grace 07 fastener. AlS3 grade B7 straples from heat #46100 also had marginal niechanical properties. isicrostructural enalysis for samp ks 21 and 22 of heat 2-27CE it.61cated abnormal microstructure for a quench and terapered product (the microstructure was not tui.perc.d marte.nsite). Sample El reveakd 4 microstructure of terspered lower bainite and ferrite. Sen.ple 22 had a microstructure of tempered martensite with smtll islands of ferrite.
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l ORGANIZATION: PRECISION HEAT TREATINC CCliPAf!Y MCNTEBELLO. CALIFORNIA REPORT INSPECTION
.NO.:,99901096/07-01 RESULTS:
PAGE 3 of 7 Based on these observations, it is apparent th6t the fasteners received inadequate heat treatment due to ir+ roper quench rates. PHT was idertified as the heat treater cf these fasteners.
CIPC subcontracted heat treating to PHT as a matter of routine.
(
Prior to the inspection at PHT, the inspector reviewed records with a former efficial of the CIPC's staff at CIPI.
This official concluded that tirae delays occurring between the he.at trcatir.s evens and the quench l
tarik were prebebly the cause of the iroproperly heat treated fasteners, because the batch heat treatnient process was being employed. With the batch heat treating process in the open fire furnace, the material is reanually removed and trar.sferred from the heat treatir.g oven to the quench tank.
- At CIPI, the hRC inspector atterapted to idertify PHT quenched and tempered, cren fire furr. ace, batch processed fasteners by heat number, custon.er, and sizes. Nine different heats were identified as being heat treated using the open fire furnace batch process. The hcat.treatnent certified work cards-(CWC) for two of these heats #99277 and #6094030 were evaluated by the LRC inspector while at PHT.
Both CWC's were complete.
y The CWC for heat 499277, Cardinal Cl #31373-3, PHT PO #24141 (3-8 x 48" SAS46) indicated that the heat treatment temperature was 1550 F for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and 30 mir.utes and subsequently quenched in oil (oil temperature was approximately 130*F). Ten;pering was carried cut in 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> at 1100*F.
A Rockwell huroness spct check yieldec a hardness of 32 RC, the required hardocss for the material is 30/34 RC. The PHT shop card for heat n094C38, Cardinal Cl #32670-2, PHT P0 #24639 (31 x 4 SA 570) indicated the material received a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> 1550*F beat treatrrent followed by an cil l
quench. The terrpering teroperature was 1050 F for C hours.
The Brinell I
hardness spot check yielded a hardness of 3El LHN, the required hardness l
177/SE1 EEN.
1 1
The open fire dual door, box furnace is 13Ca x 120" x 40" in the work j
- ene and capable of a maximum temperature of E100 F.
Two baskets of 1
material per side are usually charged, for a total of four baskets per furnace. Each basket is approximately 24" x 42" x 10" and can typically i
hold approximately E00# of material. These baskets are r.ot stacked.
Furnace temperature uniformity is held within the limits specified by MIL-H-0075 ar.d is surveyed quarterly. The thermocouple used for the ternperature uniformity. surveys are calibrated against a reference standard therr;occuple. The reference standard is calibrated at an l
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ORGANIZATION: PRECISION HEAT TREATING CChPAhY EChTELELLC, CALIFORNIA i
l REPORT IhSFECTI0f NO.: 99901096/67-01 RESULTS:
PAGE 4 of 7
)
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outside source, against a seccrcary standard, Lees E. Northrop Potentio-meter Model 6652-2 Serial //1767181.
l Each week the test thern.cccuples are used to prcbe the furnace and tlx 1
ttsperature readings are ccmpared to the furnace centrol thermocouple.
l Thermocouple repiccer.ent is carried out when the furrace control therr,c-i ccuples differ frcrr,the test thermoccuples by more than 5 F.
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PHT rabintains an oil quench tank of S' x 12' with high speed agitation, i
l and a capacity of 8500 gallons of high speed querch oil, and 6500 gallcns
{
reserve cil. There are a total of f our mechanical agitators, two on I
cach side of the oil quench tank. 1he temperature of the oil is main-l tained in the range of 75-140 F.
I i
When heat treating by the batch process it is conceivable that tiu
{
delays n4y occur frcm huraan error when physically reing the heat 1
treating baskets trora thc furnace to the cil quench tank (i.e.,
{
dropped basket, ctc.). The NRC inspector witnessed the quenchir.g l
operation 6r.d timed the transfer between oven and c,uer,ch tari at 20 I
seconds.
J The quenching medium (egite, tion, reserve oil, cil temperature) should have sufficcd for an 6dce,uate quench. Euc to the routine and relative j
ease of the transfer process, crrcrs would have to be. of a deliberate j
nature (i.c., operator not being att(ntive) or rsechanical prcblems with j
the charge transfer vehicle.
From interviews with Phl personnel the NRC inspector was informed that CIPC on occassions would submit for heat trutir.g. fasteners that hbd beer, previously heat treatec together with faster.crs that had not been heet treated. Rcheat treating of previousi) heat treated prcducts is highly ur. desirable sir.ce there is no guarantee that subsequent heat treattrents will Jit.ld the cesired microstructure or rrechanical prcperties.
Therefore, there is a possibility that the heat trcated fasteners which exhibited low and marginal ri.echanical properties at CECO were subjected to prior heat treatn.ent by CIPC.
All. heat treatruent performed ct PhT is dct6 according tc hlL-E-6875 cr custcII.er process specifications. Procedures for heat treating at PHT are speciiitc cr the shop crder ferrn which accompanies the raterial to be heat treated. At the tir,,e of the inspection, the riuclear related PC's reviewed by the hRC irspector, revcultd that 10 CFR Part 21 ha6 been iraposed on Phl.
92
Gk6ANIZATION: PRECIS 10f! HEAT TREATING COMPANY I4CNTEBELLO, CALIFOR!ilA REF0RT INSPECTION fiO.: 99901096/67-01 RESULTS:
PAGE 5 of 7 10 CFR Part 21 Issues Upon review of the PhT quality assurance docurrentation, it was apparent that a prccedure for evaluating and reporting deviations did nct exist.
Therefere.at the time of the inspection, there was no formalized method.
of evaluating potentially reportable Part 21 items or reporting Part 21 items (see Violation 87-01-01).
Calibration The PHT Quality Assurance manual requires that hardness testing equipment be calibrated every 90 days by an approved instrument laboratory. PHT is in compliance with this commitment.
PHT regularly tests their protcctive atrresphere furnaces for intergranular oxidation and decarburization by approved cutside sources. Also, n,echanical properties test anc microstructural evaluation of quenched and tempered test pieces are perforned er, a it.cnthly basis.
l Calibration icr both the Rockwell anc Erinell hardness testers were up to date. All hardness testing equipment, Rockwell.anc Erinell, was.fcund to be calibrated every 90 days and a stickt.r was' affixed l
to each trachine that inGicated the date of the calibration. The King l
(No. XR-34) and Tinius Olsen Brinell (No. 1E5165) hardness testers were calibrated by Golden State Calibration & Service Ccmpany in acccrdance with National Bureau of Standards (NBS), reference test i;os. SJTD1/lC2533. The Rockwell hardness tester fio. 5JR-16C4, Svk-1074, and 4TT-749 were cc,librated by Golden State Calibration &
Service Company to be within the limits when using a standardized diamcnd and/or ball penetretcr en standardized Rockwell test biccks (six tests were perforned).
The ranges were verifitd tc be in occordar,ce with /,STh specification E18, per MIL-Std-45662.
Harantss Tester fio. 411-745 calibrated ranses 15h-EE, 15N-91 were verified as well as tester No. SJR-1074 calibrated ranges C-62, C-45, C-LE and tester SJR-1684 calibrated ranges C-12, C-4E, C-25.
The Rockwc11 hardness tester inachire ID #2-SJR-EE/5fi 1684 was checked with test blocks and recorded on a daily basis by the shift inspector.
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i ORG/J.IZATION: PRECISION HEA1 1kEATING COMPANY H
MONTEBELLO, CALIFORhIA 1
REP 0kT INSPECTION NO.: 99901096/87-01 RESULTS:
PAGE 6 of 7 i
Constant and accurate calibration of the hardness testers are l
essential since the hardness results are used to verify material prcperties and adjust process controls.
j 2.
Thermocouple used to perform the temperature uniformity survey in j
February 1987 on the gas fired car-botten furnace fCB1 were calibrated l
by the Nelson Instrument Service Corrpany, Incorporated. The Nelson Instrument certification #11948 was revicwed and found adequate. The therraccouple correction was (+)2 F for testing tercperatures 1300, 1500, 1700*F. The standarcs used were NBS test #236962 for S/N 1640392 and hbS test #235727 for S/N 876547. The standard thern,0ccuple was calibrated according to NBS test 6238464.
The calibration of all recording potentiors.ter controllers were up to date and a sticker was affixed tc each instrunent' indicating the date of last calibration.
3.
End quench tests for hurdenability evaluations were perforced in 1987. Thest tests were done acccrding tc the requiremer,ts of MIL-H-6875G and ASTit A-255 for 4340 steel ard evaluated by F.QS Inspection, Incorporated in metallurgical lab report #37171-5-1, dated March 16, 1987. All samples met their require.d test specifications.
Basket Material Precision P0 CWC Date P0 Furnace 410 SS bar PR 18362 02/23/87 H141 T2A 4340 Bar PR 18367 02/26/66 107 T2 17-4 PH PR 18360 02/23/67 H-126-T2A 4340 PR 18367 02/26/87 h-165 T1 4340 bar PR 18356 02/20/87 155 T2A 4340 PR 18358 02/26/87 199 3
4340 PR 163bb 02/26/07 001 0F1 4340 PR 18365 02/26/87 002 0F1 8740 Steel Bar PR 18363 02/28/67 707 2
4.
Decarburization Control and high len'perature Oxidation Test were performed in February 1987 and reported by i< metallurgical Testing Corporation on March 16, 1987. The results indicated that test pieces frcm furnaces T2A,11, T1A, T2, TC1, R2 and 2 were free of decarburization and detrin.cntal intergrar.ular oxidation.
Intergranular high temperature oxidaticr. and decarburization tests were performed per LCP-73-2020A.
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ORGAl'12AT10N: PRECISION HEAT TREAllhG COMPANY N0tlTEBELLO, CALIF 0RilIA 1
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NO,: '99501096/87 RESULTS:
PAGE 7 of'7.
j REPCRT IllSPECTION Plant Tour During the plar,t tour of PhT,-the endotherraic atn.osphere contro11cd furnaces, the cpen fire r.cn-atrespheric furnace, and the: car botton,
.l te'r.,perir,g furnace '(flo. I and ho. 2) were inspected for calibration.
Thc oil quench tank was verified to have ade.quate raechanical agitation.
The transfer time between the cpen fire furnace and quer,ch tank was witnessed to be 30 seccods, l'attrial for shipment with blue acceptance tags 5: crc-verified as having information regarding customer name, number of pieces, caterial type, part number, serial nun.ber, part name and PHT inspector approval.
The f!RC inspcctor also verified proper distribution of QC inspection stamps ho. 1, 2, 6, 7, 9, 10, 11, 1E, 13, 15 to the shift supervisor, shift leadmar. and QC inspectors. Each inspector had proper possession of his 'starrp during working hours.
L Upcn revicw of four furnace charts fcr shop crders #53331, #53318 and l
- 53317 the time in which the. load was charged and removed was not indicated (see Nonconformance 87-01-01).
Shop order #53318 consisted of 422 pieces cf heat treat ASTM A354BD,
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it3/L x 1, HX HD materibi tcr A&G Engineering Cornpany.
Shop order j
- 63317 consisted of 654 pieces cf heat treated ASi1E SA, 546 Grace 630, l
H1100 7/8 livy Hx nut material fcr A&G Engineering. Shop crder #53332
)
ccasisted of 125 '1bs. cf heat treated ASitE SA 564 TY 630, H11001 1/8
};vy hex nut material for A&G Engineering.
f E.
Pt.RSONS C0liTACTED:
1
- Brict. L. Launder, Vice President i
- Janies F. Lastra, QC Manascr 1
- P. C. Scribante, Sales
- Present et the exit Deeting.
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M -ORGANIZATION: -SATIN'AMERICAN' CORPORATION SHELTON, CONNECTICUT INSPECTION-:
INSPECTION REPORT DATES:.07/06-08/87' ON-SITE HOURS:
16 NO.: ' 99901094/87-01
,CORRE5PONDENCE AUUKLbb: b6110 American Lorporan on ATTN: Mr. Joseph Satin President.
Post Office' Box.619 40 Oliver Terrace Shelton, Connecticut 06484 ORGANIZATIONAL CONTACT: Aram Nahabedian
- TELEPHONE. NUMBER:
.203-729-6363.
NUCLEAR INDUSTRY ACTIVITY: Satin American provides new and returDisneo
~i circuit' breakers and switchgear to the nuclear industry.
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ASSIGNED INSPECTOR:
fIvd/h..t,
P//3/f/
'cotison, Special Projects Inspection
'Date Jefffey JFtc on PIS)-
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OTHER INSPECTOR (S'): Richard P. McIntyre, SPIS l
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APPROVED BY:
U b
(L<A/>
9-1-S'T l
Uldis Potapovs, Chief, SPIP, Vendor Inspection Branch Date l
i IHSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR Part 21 end Appendix B to 10 CFR Part 50.
B.-
SCOPE: This inspection was conducted to review records pertaining to the supply of'200 Class 1E circuit breakers to the Farley Nuclear Plant.
Additionally, a review of Satin American's quality assurance program was
~ erformed.
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PLANT SITE AP. PLACABILITY: Farley 50-548/364, flaine Yankee 50-309, Indian Point 2, 50-247.
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ORGANIZATION: SATIN AMERICAN CORPORATION SHELTON, CONNECTICUT j
REPORT INSPECTION HO.-
99901094/87-01 RESULTS:
PAGE 2 of 7 A.
VIOLATIONS:'
None.
B.
HONCONFORMANCES:
1.
Contrary to Criterion VII of Appendix B to 10 CFR Part 50, the Satin American Corporation quality assurance program does.not require that vendors who supply components to be used in safety-related applica-tions, be on the Satin American Corporation Approved Vendors List (APL).
(87-01-01) 2.
Contrary to Criterion IX of Appendix B to 10 CFR Part 50, welders, whose qualification had not been established, were listed as quali-fied in the Satin American quality essurance. manual.
(87-01-02)
C.
UNRESOLVED ITEMS:
Hone.
i D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
None.
)
E.
OTHER FINDINGS OR COMMENTS:
)
1.
Supply of safety-related molded case breakers to the Farley Nuclear l
Plant A recent inspection conducted at the Farley Nuclear Plant by the NRC revealed that a large number (200) of safety-related molded case circuit breakers had been procured from Satin American Corporation for use in Farley's 600 volt motor control centers. Further inspec-tion revealeo the fact that a number of these breakers had a 400 volt rather than a 600 volt U.L. rating and that Satin American had tagged and certified these breakers as being acceptable for use in 600 vac circuits.
During the inspection at Farley, Alabama Power was asked to provide documentation which would support the use of these 480 volt rated breakers in their 600 volt motor control centers. Also in question was the circuit breakers seismic qualification.
Alabama Power could not provide sufficient documentation to support either the seismic qualification or the use of these breakers in 600 volt applications.
98
4 ORGANIZATION:
SATIN MiERICAN CORPORATION SHELTON, CONNECTICUT REPORT INSPECTION Nn.*
99901n94/A7-01 RFSulTS:
PAGF 9 nf 7 1
The circuit breakers in question were originally manufactured by ITE and are of the Type HE molded case design. At one time, these breakers carried the 600 volt U.L. rating, but soine years ago ITE moved their manufacturing facility and at that time decided to drop the 600 volt rating. Type HE breakers manufactured in subse-quent years carried a 480 volt U.L. rating. Review of U.L. standard 489 for molded case breakers indicates that six main tests are required to be performed on sample breakers in order for the breakers to carry the U.L. rating.
The first test is a calibration check done at 135 percent and 200 percent of the breakers current carrying capacity. This check was performed only on selected sample breakers shipped to Farley.
The second test is an overload test to be performed at 600 percent of the breakers current carrying capacity at rated voltaoe.
This test was not performed by Satin American Corporation.
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The thiro test is a tungsten load test to be performed at rated current. This test was not performed by Satin American Corporation.
The fourth test is a tempreature test performed at rated current.
Satin American Corporation did not perform this test.
The fifth test is an endurance test at rated voltage. Satin American did not perform this test.
The sixth test is a dielectric test at 1000 volts plus twice rated voltage. Satin American Corporation did perform this test on all the breakers supplied to Farley.
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In conclusion, the subject breakers were never subjected to tests j
two and five at the 600 volts in which they were to be ultimately used. Additionally, due to the fact that no lot control was maintained by Satin American Corporation (the 200 breakers may have come from many different orignial manufacturing lots), the number one calibration check should have been performed on all breakers.
Also, Satin American Corporation has not audited the original manufacturer and therefore has no assurance that design changes which could have affected the breakers original seismic qualifica-tion were not made. Satin American Corporation procured the subject breakers commercial grade and received no certificate as to the breakers seismic qualification.
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y ORGANIZATION: SATIN AMERICAN CORPORATION SHELTON, CONNECTICUT
-REPORT INSPECTION fin. - QQQOln94/R7-01 RFEUITS-P AfW d nt' 7 Documentation indicating'that Satin American had informed Alabama Power that sone of the 600 volt breakers they had ordered would only carry a 480 volt U.L. listing was presented during this inspec-tion.
i 2.
_ Supply of overcurrent trips to Maine Yankee The inspectors reviewed documentation pertaining to the supply of three General Electric EC-1 Type Circuit. breaker overloads to the Maine Yankee Atomic Power Plant. Although Maine ~ Yankee did not
.l indicate that these overloads were for a safety-related application, they did invoke ANSI N45.2.2 Level.B and the Satin American.
'Corporatio'n QA program. The overloads were ordered under Purchase Order 47714-00 dated June 13, 1987 and were supplied by Satin American Corporation as being suitable for use in safety-related applications.. The inspector reviewed test results that showed the overloads had received a thorough series of tests before being dedicated by Satin American Corporation as safety-related Class lE l
equipment.
Calibration of the overloads to GE calibration curve GES-6000 was performed at 95%, 120%, and 300% of the long time pickup setting and at 250% of the short time pickup setting. Also the long time trip times were checked for pickups set at 80%, 120%,
140%, and 160% of the coil rating. The short time trip times were 1
checked for pickups set at 400% and 1000% of the coil ratings.
Additionally, a reset time test was performed where current was decreased from 300% to 80% of pickup and the overloads were' observed to reset.
In conclusion, the testing done on the GE overloads appeared to adequately serve as basis for a safety-related dedication.
3.
Refurbishment PO from ConEdison The inspectors reviewed a Purchase Order from ConEdison (P0 709397) for refurbishment of a Westinghouse DB-75 breaker installed in their Indian Point 2 Nuclear Power Plant. The purchase order invoked 10 CFk Part 21 and the Satin American Quality Assurance Manual.
This breaker had not yet arrived at Satin American Corporation at L
the time of the inspection.
l 4.
Supply of GE EC-1 overcurrent trip devices to Farley The inspectors revieweo Purchase Order QP-1240 dated October 14, 1986 from Alabama Power Corporation for purchase of ten GE-EC-1 over-current trip devices for use in their Farley Nuclear Plant. The 1
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ORGANIZATION: SATIN AMERICAN CORPORATION l
SHELTON, CONNECTICUT l
l REPORT INSPECTION NO.-
99901n94/A7-01 RFSUITS' PAGF G nf 7 devices.were.specified as QA Code A (safety-related) by Alabama i
Power, however, 10 CFR Part 21 was not invoked.
Requirements of the Purchase Order specified Satin American Corporation provide a copy of their QA manual and a certificate of conformance stating.that the overloads supplied are equal to the original aevices in form, fit, and function, f
Satin tested these overloads at 95%, 120%, and 300% of pickup for longtime and at 1200% of pickup for instantenous trips. Trip times were obtained from GE curve 088980484, Figure 44 and curve GES-6000.
Testing of these overloads appears to have been adequate for installa-tion into a safety-related application.
5.
10 CFR Part 21 Program 1
The inspectors noted that applicable portions of 10 CFR Part 21 were posted as required both in the office and in the shop areas.
A-flow chart indicating Part 21 responsibilities was also posted in these areas.
l 6.
Control of Purchased fiaterial, Equipment, and Services i
Review of Section 6.0 of the Satin Americt.i Quality Assurance manual revealed the fact that vendors supplying components to Satin American were not required to be on their approved vendors list. Additionally, the manual did not state how materials could be procured from vendors not on the approved vendcrs list.
Criterion VII of Appendix 8 to 10 CFR Part 50 requires measures be established to control purchased material.
This deficiency is identified as Nonconformance 87-01-01.
i 7.
. Control of Special Processes
~
The inspectors reviewed Section 9.0 of the Satin American Quality Assurance Manual and verified that special processes such as welding, soldering, brazing, zinc chromate plating, silver cyanide plating and degreasing are controlled and accomplished by qualified personnel using qualifiec procedures.
When reviewing the qualifications for welders at satin American, two personnel were listed as certified welders, when in fact they were not certified. Satin American stated these welders had not performed any inshop welding, and that all welding was performed by outside contractors.
This incorrect personnel qualification is identified as Nonconformance (EC-01-02).
101 L___________________-_____-___-_______.
1
ORGANIZATION:
SATIN AMERICAN CORPORATION SHELTON, CONNECTICUT REPORT INSPECTION NO.!
999nin94/R7-01 RFSill TS; panr A nf 7 All other personnel qualifications and the procedures for controlling special procedures appeared adequate.
8.
Audits of Satin The inspector reviewed an audit of Satin American performed on August 21-22, 1986 by Cygna Energy Servies for the Boston Edison Company. The auditors had documented four findings concerning supplier requalification, QA audit plans, manufacturing procedures, and training of personnel.
Corrective action to the above findings was implemented by Satin American in January 1987 and appears to have adequately addressed the auditors concerns.
Additionally, a survey report conducted August 26, 1986 by Southern Company Services was reviewed.
No deficiencies were noted by the auditors during the survey.
9.
Control of Measuring and Test Equipment The inspectors reviewed Section 12.0 of the Satin American Quality Assurance Manual for the Control of Measuring and Test Equipment.
The inspectors verified that the monthly calibration status for test equipment and metering devices had been documented as required.
The status reports from September 1984 until June 1987 were reviewed.
The inspectors went on a shop tour and randomly verified that pieces of test equipment such as clamp-on ammeters, high potential test sets, high current analyzers, megohmeters and voltometers were not beyond their calibration due date listed on the calibration sticker and on the calibration status report. The inspectors reviewed the calibration dates for 8 measuring and testing devices and none were beyond their due date. No Honconformances were identified during this part of the inspection.
- 10. Exit Meeting Following the inspection, an exit meeting was held where the findings of this inspection were discussed. The following were in attendance of the exit:
R. P. McIntyre, Reactor Engineer, USNRC R. C. Nartouich, QA Supervisor, Satin American Corporation 102
. ORGANIZATION:
SATIN AMERICAN CORPORATION SHELTON,. CONNECTICUT REPORT INSPECTION NO.-
99901094/87-01 RESul TS t par,F 7 nf 7 J. B. Jacobson, Reactor Engineer, USNRC A. Nahahedian, Vice President, Satin Arnerican Corporation J. Satin, President, Satin American Corporation L. Gradin, Dir. Engr / President, EchoTech, Incorporated I
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l ORGANIZATION: SOUTHWESTERN ENGINEERING COMPANY J0PLIN, MISSOURI REPORT INSPECTION INSPECTION l
NO.: 99901083/87-01 DATES: 04/06-08/87 OH-SITE HOURS; 2d CORRESPONDENCE ADDRESS:
Southwestern Engineering Company I
ATTN: Mr. llick Capra I
General Manager Post Office Box 13851 Joplin, Missouri 64801 ORGANIZATIONAL CONTACT: Arnold Davis, QC Manager TELEPHONE NUMBER:
(4171 782-5080 NUCLEAR INDUSTRY ACTIVITY: Mositure separator reheaters (MSR) and condensers.
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ASSIGNED INSPECTOR:
h%
~87 Date fj.'T.Conway,Progr#(DevelopmentandReactive nspection SectioQ PDRIS) f OTHER INSPECTOR:
- 1. Tinkle, Consultant
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4.24 81 APPROVED BY:
Date
!d. {. Stone,' Chief, P RIS, Vendor Inspection Branch i
INSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR Part 50, Appendix D ano 10 CFR Part 21.
j B.
SCOPE: This inspection was made to conduct a programmatic evaluation i
oTtiie implementation of the QA program pertaining to the modifications l
made by Southwestern Engineering Company (SEC) of MSR internals at I
Virginia Power Company's (VPC) North Anna and Surry nuclear facilities.
PLANT SITE APPLICABILITY: North Anna 1 and 2 (50-338/339) and Surry 1 and 2 1
(50-26C/281).
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ORGAhlZATION: SOUTHWESTERN ENGINEERING COMPANY.
J0PLIN, MISSOURI i
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REPORT INSPECTION I.
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PAGE 2 of: 8(h Nn - ocon1nR1/R7-01 RESULTS:
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VIOLATIONS:
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None.
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8.
NONCONFORMANCESi
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v None.
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.C.
UNRESOLVED ITEMS:
None.
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STATUS OF PREVIOUS INSPECTION FINDINGS:
,A jb c
None. This was the first inspection at this facility.
E.
OTHER FINDINGS'AND C0KkENTS:
T i
1.
SEC
'I
<d g
During various discussions, the SEC Quality Control (QC) Manager provided backgrounorand other information concerningiSEC's facilities and nuclear business involvement. A briqi rammary Q-follows: SEC designs and fabricates large and hip'g 1sure i
heat exchanger equipment for the petro-chernical and a e/wic f,
. power generating industry. Over the past twenty years, the i
company's principal business activities have involved design and j
manufacture of MSRs,ifeedwater heaters (FhH2, and steam surface j
condensers. SEC produces equipment for both p ssil and nuclear i
,q electric generating plants. SEC's corporate office, including the corporats, staff, engineering, and long lead procurement h
activities are located in the City of Commerce, California.
e The corporate office is also responsible for contract l.
administration for on-site installation work perfornied by SEC. SEC has two plants in Joplin, Missouri.
engaged in manufacturing condensers is located,Anglder plant St'.ith and Wall Street. A new plant located on 9th Street is er, gaged in manufacturing equipment such as pressure vesselst "SRs, and
- FWHs, SEC also has a facility located in Burlin, ;isconf n i and a field service' office in Florence, Kentucky.
ThetWnconsin facility is planning on getting an "h" stamp for Section III ASME Code work.
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- __ _______ _____ ________ ________ _____t
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s ORGANIZATION:' SOUTHWESTERN ENGINEERING COMPANY J0PLIN4MIE{#dRI 3<
qs
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,INSPECTKN
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J REPORT ui ooon1nP1/A7 n1 RFShlTSp V
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PAGE 3 of 8 y
i Q
Prior to 1983 SEC's manufacturing plant wa's locatgd at its corporate headquarters in California.
In 1983 t'he California manufacturing capability was moved to the 9th Street plant in Uoplin. SEC maintai.ns the Joplin plant at lith Street and Wall, but it is~ currently closed due to insufficient work volume. The new Joplin facility started operati'ng in mid 1983.
The facility qualified for and jobt.ained,"N" and "NPT" stamps for ASME Section III wdrH. Sitece SEC did not receive any orders for ASME Section III C6d& work, they decided not to renew the stamps when they ekpired in December 1986. T4 date, the new Joplin plant has "U," "S" and "R" stamps and has only manufactured i
ASHE Code components to Section VIII. Purchase Order (P01 1
79-5721-12 from Duke Power Company placed in 1979 was the last y
i customb/ order for ASME Section..II)., Code equipment. The order was for fuel pool cooling heat excqvgyrs which were manufactured at the California plant. Job order file; and records for all manufacturing work done at the California plant are maintained in California.
f 2.
OC hanual Various secticN of SEC's QC Manual, Revision 1, dated February 20, 1987 were revfewed. The)QC Manual contains a description of the QA program intended to meet the requirements of Section VIII
" Pressure Vessels" Division I and Section I " Power Boilers" of the ASME Code and the National Board Inspectiun Code, Edition 4 Revision 1.
The QC Manual is applicable for work conducted at both of SEC's facilities located in Joplin.
It was noted that l
NDE personnel and welders are qualified to SNT-TC-1A and Section IX of the AShE Code, respectively.
s
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3.
Domestic Nudlear Customers M
A summary of the 65 MSRs fabricated by SEC. (doplin) to the requirements of Section VIII of the ASME Code from 1983 to the presentdgr domestic nuclear customers follows:
Date Job No ?
Customef Shipped I
8h-28-11 thru 14 Wiscor. sin Electric (Point Beach) 9-83 93-26-01 thrf 04 Rochester' Cas & Electric (Ginna) 3-84
'63-44-01 thru 04 Baltimore 6as & Electric 4-84 y
(Calvert Cliffs) 107 a
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5 ORGANIZATION: SOUTHWESTE R ENGINEERING-COMPANY x_g<0
. 'J0PLIN, MIS 5OURI y
4 N
REPORT s
INSPCCTION En - ooon1mam7-n1 RESULT g
) PAGE 4 of a 3
Job No.
Custbry Shipped
' 'b 83d?-01/02-Connecticu\\ht.nkea(HaddamNeck) i 7-84
<B4-22-01 thru 04 Virginia Electrf4 (North Anna) '
8-84 L
84-16-02 thru 05 Vi rginia JTMctr 9/10-84 84-20-01'thru 04,.a EVirginit ElecW s(c1(Surry)-
c (Sorry) 12-84 85-11-01 thru It. TVA,(Sequoyah) >\\
8/9-85 1
i 85-11-21 thru 32 TVA\\
12-85/1-86 84-42-01 thru 06 Pub' tic Service Electric & Gas 2/3-86 i
f1.
3.(Salem) g x 853 1 0,1/02 and
'TUGC0 (Ccmardhe Peak) 3/7-86
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C6121-0,3 Public Service Electric 6 Gas 8-86 4
i (Salem) 86-14-01.thru 03 CDuke Power (0conee) 10/11/12-86 s
i 3m 86-10-01 thr'u OL?
T1orida Power (Crystal Piver) 9, 1/3-87,
' R 86-24-01.thru 04' Fort 1and General Electrit*(Trojan) 3-87 A-
"86-31-0S"thru 08' Westinghouse (WPPSS) 4-87 A sumrary of the seven condensers fabricated by SEC (Japlin)
E c
to the requirements of Section Vill of the ASME Code from 1983
['
to the present' fe4 domestic nuclear customers follows:
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Date Job N_o_.
Customer Shipped
)
l 4
6750 i
Houston Power & Light (South Texas) 10-63 i '\\
c 84-3510 h
Virginia Pcwer 10-85 84-3520 Vi rginio' rower 1-86 85-2810/2820Y TUGC0 (Comanche Peak) 4/6-86 85-2610/262%'\\
10GC0 (Comanche Peak) 6/11-86 i
j 4
i The inspector selectively reviewed 3cb files for-13 replacement ql NSR tube bundles and components. Tha orders included Job Nos.
!j 84-22-02, fA-16-05, 84-22-01, 82-28L13, 86-14-01, 86-31-06, 86-24-01, 86-10-05, 83-44-04, 83162-02, 85-27-12, 84-42-01, and 85-11-06. Each file contained a technical specification which, among other things % listed applicable codes and standards
?
i' (e.g., Section Vill for design and manufacture,.Section IX for o
welding, Section Il-Part A for titerials, and Sectfon V for NDE I
i and SNT-TC-1A for examiners) and gave the requirements for a-l9 documersted QA program.
In Jnany cases the customer would install
- 4 the replacement tu_be4end14 with SEC providing engineering ls technical AssistanALUpoort, but in others SEC would be responsibl&forremoifngandinstallingthenewMSRcomponents, i
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0 ORGANIZATION: SOUTHWESTERN ENGINEERING COMPANY J0PLIN,.hlS500RI-i qh; hi l
C REPORT:
INSPECTION nn, 04Qn10M /n7-01 '
RESULTS:
PAGE 5 of R H
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.TheLfiles included a Traveler / Inspection Plan and Record!for.
j
.the'tubesheet, bundle, head, tubing and final' assembly; copies of. drawings marked with a QA stamp; Bill of Materials; Procedures
- List; CMTRs from tubing and weld material manufacturers; nonconformance/ corrective action reports; magnetic particle,
@s liquid penetrant, ultraso).ic ano hydrotest reports; inspection reports; radiographic weld maps and reader sheets; heat treatment charts; and shipping papers.
In addition, QA documentation
' packages to: the customer included Form U-2 Manufacturer's Partial
,.%a Data Report, CMTRs, Leak Testing. Report (tube to tubesheet joints),-
MLN' NDE and hydrotest; reports; and SEC certifications'that all: material complies with Section:II of the ASME Code, MSR components comply.
lwith Section'VIII, and the unit was hydrotested in accordance 1.
with the specification requirements.
l f
It was noted that heat-treatment charts dated June 25, 1986 and' July 15, 1986 for Job No. ' 85-27-12 did not identify the affected i
component on.the chart which does not agree with the requirements in Section 8.1 " Heat Treatment Systems Control" of the QC Manual.
0 5.
MSR Modifications L r.
The NRC inspector reviewed the records retained at SEC's Joplin
' N facility for.the MSR internals supplied to North Anna (Job No.
84-22-01 thru 04) and Surry (Job No. 84-16-02 thru 05 and.
84-20-01 thru 04). A review of the documents-(specific types identified in Section D3 of thi.s report) indicated that the MSR internals were designed, manufactured, and tested to the requirements'of Section VIII of the ASME Code and shipped to the North Anna site in August 1984 and the Surry site in September, October, and December 1984.
l During the week' of the inspection, copies of VPC P0s and contract documents were received from SEC's corporate offices lf in Californio. P0 Nos. ET-41600-SC and ET-41586-SC for Surry and ET-41603-SC for North Anna were issued to SEC for materials and engineering services. Each P0 contained several attachments.
. Attachment IB VEPC0 Special Terms and Conditions for Nuclear I
Work" (STC 1-STC 13) was dated February 18, 1983 and was included in each P0.- Section II of Attachment IB stated that the provisions
)
of 10 CFR Part 21 nay be applicable to the work performed, and i
J the supplier shall comply with the requirements of 10 CFR Part 21 to the extent applicable.Section III of Attachment IB stated I
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ORGANIZATION: SOUTHWESTERN ENGINEER'ING COMPANY J0PLIN, MISSOURI REPORT-
- INSPECTION Nn - oqon1nna/R7-01 RESULTS:
PAGE 6 of 8 that the supplier shall adopt and implement a QA program that meets all NRC requirements including 10 CFR Part 50, Appendix B.
Attachment IIA. referenced the' technical specification for each project.
.In a review of SEC' documentation packages sent to VPC, it was noted that SEC did not take any exception to the requirements of the P0s or specifications as referenced in Certificate of Comp 21ances signed and dated.0ctober 17, 1984 (Job No. 84-16-05) and August 23, 1984 (Job No. 84-22-01) by the QA Manager or equivalent.
Contract Nos. PSC-C-0184704 and FHN-378-0228 were'between VPC and SEC.for MSR modifications at Surry, Unit Nos. I and'2 and North Anna, Unit Nos. 1 and 2, respectively. The scope of work in each contract was for SEC to provide services and materials for the removal and installation of the new MSRs at both nuclear 3
sites. Section 5 of PSC r-0184704 and Section 3 of FHN-378-0228 4
addressed the applicability of 10 CFR Part 21, and both stated
. in part, " Contractor understands that the provisions of the Title'10 Code of Federal-Regulations (CFR) Part 21 apply to this contract."
In discussions with the QC Manager, the hRC inspector was told that installation activities performed at nuclear sites are administered by SEC's corporate office.
For the Surry ano North Anna projects, SEC subcontracted ~to Powerplant Specialists (PS) to perform the installation activities.
It was also stated that SEC is responsible for providing the qualified welding procedures and for qualifying the welders. All the records related to the on-site installation' activities are maintained in the corporate office in California. These records include welding procedure specifications, procedure qualification records, and welder qualification tests.
A future inspection of the SEC and PS facilities in California will include a review of welding records, 6.
Nondestructive Examination (NDE)
The NRC inspector reviewed Sections 7.1 and 7.2 of the QCN, eight NDE procedures, and the qualification records for nine NDE. personnel.
Procedure AS-QA-1.5 'tritten Practice for Certification of NDE Personnel" documented SEC's qualification /
certification program. The four procedures addressing magnetic 110
ORGANIZATION: SOUTHWESTERN ENGINEERING COMPANY J0PLIN, MISSOURI
' REPORT.
INSPECTION Nn -
cooninM/R7 01 RESULTS:
PAGE 7 of 8 particle, radiographic, ultrasonic and liquid penetrant testing were prepared by a Level JII examiner and were approved by the QC Manager who is also certified.to a Level'III.
In general, the type of information found in the record files for the nine NDE personnel (four -Level III and five -Level II) included:
educational background, resumes, certificates of training completed, record at qualification, certification statements, examinations (general, specific, practical) and eye exams.
It was notea that an eye exam for individual SECC 16 was not dated, and a signature and date were missing for an eye exam for individual SECO 19. The qualification records appeared to satisfy the requirements of SNT-TC-1A with the exception that the records did not contain a statement indicating that each NDE personnel had satisfactorily completed training in accordance with Procedure AS-QA-1.50, 7.
Calibration of Measuring & Test Equipment (M&TE)
The inspector reviewed 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 intcrvals. The M&TE selected for review include the following, identified by serial number (S/N):
two MT (P-90) (33 and 80567), pressure gauge (42), five temperature gauges (113, 137, 152, 155, and 156), Sonotest (UFD-7) (7119), tor amp probe (AX53099)que wrench (12425), tong tester (AX46503),
and two temperature recorders (HR5-052374-1 I
and MDH-8325-3642).
It was noted that temperature gauge (137) was calibrated by Tulsa Gauge and Instrumentation.
NDT Technology calibrated the Sonotest UT instrument, and General Testing Laboratory calibrated the torque wrench. The amp probe and tong tester were calibrated by Agra Engineering. Calibration of the MT (P-90) (80567) was performed by Venture, and the two temperature recorders were calibrated by Honeywell. Certification records do not reflect traceability of calibration standards used for some instruments (e.g., S/hs 137, 12425, 80567, HRS-052374-1, MDH-8325-3642-F051). Documentation was not available to indicate the traceability requirements contained in Procedure AS-QA 1.60
" Instrument and Gage Calibration" are being met in these cases.
111
ORGANIZATION: SOUTHWESTERN ENGINEERING COMPANY J0PLIN, MISSOURI 1
1 REPORT INSPECTION on - coon 1nn1/97 n1 RFSill TS -
PAGE 8 of 8 Gaps were observed between the calibration due dates and the actual calibration dates _ for some instruments (e.g., S/Ns 42,.113, 137.
-AX53099). This was discussed with the SEC Quality Engineer responsible for calibration. He stated that if instruments are not recalibrates prior to the calibration due date, they are removed from service and kept in his office. There was no documentation to indicate that the requirements for tagging and segregation contained in Section 7.3'" Instrumentation and Gage Calibration" of the QC Manual are being met in these cases.
There was no recent documentation to indicate that the requirements contained in Section 7.3 of the QC Manual and Procedure AS-QA 1.60 are being met with respect to performing surveys of calibration vendors (e.g., Venture for S/N 80567). According to the calibration records and calibration frequency requirements contained in procedure AS-QA 1.60, the Sonotest UT instrument (S/N 7119) has been overdue for outside calibration for over six months.
It was further noted that SEC only has one of these instruments.
The five temperature gauges were being used on drying ovens in the weld material storage area. Procedure AS-330 " Storage and Handling of Shielded Metallic Arc Electrodes" requires that stainless steel, nickel base, and cooper base electrodes be stored in a holding oven between the temperature range of 100*F to 125 F; and low hydrogen, mild steel and low alloy electrodes be stored in holding ovens at 150 F.
All five ovens were operating at high temperatures based on the gauge readings of 100, 178, 168, 166 and 162 F.
F.
PERSONNEL CONTACTED l
- N. Capra, General fianager
- A. Davis, QC hanager E. Cochran, Quality Engineer 1
9ttendeo Exit Meeting 112
1 l
. ORGANIZATION: SOUTHWESTERN LABORATORIES HOUSTON, TEXAS
. REPORT
' INSPECTION INSPECTION.
N0. :t 99901097/87-01 DATE: 08/03-06/87 ON-SITE HOURS: 28 CORRESPONDENCE ADDRESS: Southwestern Laboratories ATTN: Mr. Joseph Jeanes Chief Executive Officer 222 Cavalcade Street Houston, Texas 77009 ORGANIZATIONAL CONTACT: Daniel N. Hanna, Jr., QA Manager.
TELEPHONE NUMBER:
.(713)692-9151 NUCLEAR INDUSTRY ACTIVITY: Analytical services, metals testing, and nondestructive examination.
ASSIGNED INSPECTOR:
NkM
!OM-8}
(
T. Conway, Prog a Development and Reactive Date
.J.}InspectionSecti PDRIS)
V OTHER INSPECTOR (S): Terry. Tinkle (consultant)
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' APPROVED BY: NI
-L-J. C. Stonel Chief, PDRIS, Yendor Inspection Branch Date INSPECTION BASES AND SCOPE:
A.
BASES:
10 CFR Part 50, Appendix B and 10 CFR Part 21.
B.
SCOPE:
The purpose of the inspection was to conduct a programmatic evaluation of the implementation of the QA program in the areas of training / qualifications, control of purchased services, control of
)
measuring and test equipment, audits (internal / external), and reporting of defects.
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PLANT SITE APPLICABILITY: Fort St. Vrain (50-267) and Comanche Peak 1/2
.(50-445/446).
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ORGANIZATION: SOUTHWESTERN LABORATORIES HOUSTON, TEXAS REPORT INSPECTION NO.: 99901097/87-01 RESULTS:
PAGE 2 of 13 A.
VIOLATIONS:
1.
Contrary to Section 21.6 of 10 CFR Part 21, a copy of Section 206 was not posted in three areas where Part El and a procedure were posted.
(87-01-01) 2.
Contrary to Section 21.31 of 10 CFR Part 21, a review of purchase orders (P0) to service vendors revealed that while 10 CFR Part 21 was imposed upon SWL, SWL did not impose 10 CFR Part 21 requirements on P0s 5150 (December 23, 1986), 1810 (July 21, 1987), and 1308 (November 21, 1986) to r fter Laboratories, and 5542 (March 5,1987) o to Page-Wilson and 6183 (June 11, 1987) and 440 (January 23,1985)to Gulf States Balance.
(87-01-02) 3.
Contrary to Section 21.21 of 10 CFR Part 21, SWL's 10 CFR Part 21 Procedure No. 9700-108-87 does not ensure that a responsible officer shall be notified in all cases when a defect is found in a safety-related basic component.
(87-01-03) 8.
NONCONFORMANCES:
1.
Contrary to Criterion V of Appendix B to 10 CFR Part 50. Section 6 of ANSI N45.2 and Section 5.3 of the QAM, instructions and/or proce-dures did not exist for the following activities:
(87-01-04) a)
Establishment of criteria for the qualification of auditors.
b)
Control (e.g., origination, review, approval) of procurement documents.
i c)
Overall individual / group responsibilities from the receipt of an item to the origination, review, and approval of the final report.
d)
Generation and use of the Approved Vendors List (AVL).
2.
Contrary to Criterion VII of Appendix B to 10 CFR Part 50, Section 8 of ANSI N45.2, and Section 7.3 of the QAM, there was no documented evidence that a survey or audit had been undertaken on three vendors who perfonned services for SWL. lhe vendors include Coffer Labora-tories (December 1986 and July 1987), Page-Wilson (March 1987), and Technology & Calibration (March 1986).
(87-01-05) 114
.I ORGANIZATION: SOUTHWESTERN LABORATORIES HOUSTON, TEXAS REPORT.
INSDECTION
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N0.':
99901097/87-01 RESULTS:
PAGE 3 of-13 3.
Contrary 'to Criterion IV of Appendix B to 10 CFR Part 50, Section. 5 of ANSI N45.2, and.Section 7.4 of the QAM, the requirement to have an acceptable QA program was not included or referenced'in P0s to Coffer Laboratories (Nos. 5150, 1308, and 1810), Pa 5542),andGulfStatesBalance(Nos.6183and0440)ge-Wilson (No.
(87-01-06) 4.
Contrary to Criterion XVIII of Appendix B to 10 CFR.part 50, Section 19 of ANSI N45.2, and Sections 18.11 and 18.3.6 of the QAM, there was no documented evidence of written procedures, checklists, or a report for an audit performed of Grady Cook Sales in May 1985.
(87-01-07) 5.
Contrary to Criterion II of Appendix B to 10 CFR Part 50, Section II of ANSI N45.2, Section 1.4.1 of the QAM, and Sections 3.2 and 6.0 of Procedure No. 9700-105-82, a review of training records revealed the following:
(87-01-08) a)
Tests and certifications to ANSI N45.2.6 were missing for F. Plummer and A. Norwood, both from the Metals Testing and Metallurgy-(MTM) department.
b)
Documented evidence did not exist to show that J. O'Neal from the Analytical Services (AS) department and M. Peterson and C. Albritton from the MTM department had been trained or indoctrinated in the QA program.
c)
Indoctrination and QA orientation were not documented on form SWL-105-2 for approximately 31 individuals (11-NDE 4-AS, and 9-MTM).
i 6.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Section 5.4.4 of the QAM, the QA Manager failed to sign:
(a) Report of Analysis and Tests No. 42711-9 (April 18, 1986) to Bechtel and Nos. 947 (June 18, 1986), 830-842 and 744-842 (June 11, 1986),
6938-6939(December 3, 1985), 5958-5962 (October 22, 1985) and 3671-3683 (June 28, 1985) to Texas Utilities Generating Company; and (b) approximately 55 NDE reports (7 to Gulfalloy in 1982 and 1983, 7 to WFI in 1986, 41 to Ladish in 1984 and 1985).
(87-01-09) 7.
Contrary to Criterion I of Appendix B to 10 CFR Part 50 and j
Section 3 of ANSI N45.2, there was no documented evidence of the authority and responsibility of all the persons and organizations which are a part of the QA department.
(87-01-10) 115
ORGANIZATION: SOUTHWESTERN LABORATORIES HOUSTON, TEXAS REPORT INSPECTION NO.:
99901097/87-01 RESULTS:
PAGE 4 of 13 8.
Contrary to Criterion IV of Appendix B to 10 CFR Part 50 and Section 5 of ANSI N45.2, there was no documented evidence that adequate measures were established for the issuance, review, and approval of P0s including changes to the P0.
(87-01-11) i 9.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Section 11.4.4 of the QAM the identity of the person conducting testing was not documented on some test documentation.
(87-01-12)
- 10. Contrary to Criterion XII of Appendix B to 10 CFR Part 50 and Section 5.2.1 of the QAM, some measuring and test equipment (M&TE) were not calibrated at the specified frequency.
(87-01-13)
- 11. Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Section 5.7 of Procedure No. 9700-102-75, calibration vendors have not always supplied the proper traceability on certification documentation.
(87-01-14) l
- 12. Contrary to Criterion XVIII of Appendix B to 10 CFR Part 50, Section 18.4 of the QAM, and Section 4.3.2 of Procedure No. 9700-104-82, it was noted that internal audits of some departments were not conducted at least once every 12 months, and a checklist was missing for one internal audit.
(87-01-15)
C.
STATUS OF PREVIOUS INSPECTION FINDINGS:
None. This was the first inspection at this facility.
D.
OTHER FINDINGS AND COMMENTS:
1.
Southwestern Laboratories (SWL)
The headquarters office of SWL is located in Houston, Texas, and the company has 13 branch offices throughout Texas and Louisiana.
SWL provides test and inspection services for the following areas:
Geotechnical Engineering Construction Material Er.gineering Wood Product Inspection and Testing Nondestructive Examination Metallurgical Engineering Mechanical Testing Environmental Services Analytical Services 116
y ORGANIZAT'ON: SOUTHWESTERN LABORATORIES HOUSTON, TEXAS REFORT INSPECTION NO.: 99901097/87-01 RESULTS:
PAGE 5 of 13
'According to the Director of Metals Testing approximately 10 years ago, SWL adopted a single QA program for all their work. According to statements in Section 1 of the QAM, Revision 4, dated July 15, 1983, the SWL quality program meets the intent of ANSI N45.2 and complies with the requirements of the ASME Code. The QAM further states the SWL quality program covers the full scope of 10 CFR Part.50, Appenaix B.
2.
Control of Perchased Services The inspector reviewed Section 7.0 of the QAM, the AVL and external audits to assure that qualified venders were being used by SWL. The AVL dated August 8, 1985 was signed by the QA Manager, and it listed eight companies as calibration service vendors.
It was noted that the issuance and control of the AVL was not documented in the QAM, procedures, or instructions (See Nonconformance 87-01-04).
l The external audit report contained a check list for 11 categories:
Organization; QC Program; Industries and Procedures; Identification and Control of Materials; Control of Special Processes; Control of Inspection Examination and Testing; Control of M&TE; Process Routing Sheet; Nonconforming Materials and Equipment; Records; and Audits.
The check lists for the four audits in 1985 and the four audits in f.
1986 did not contain a discussion statement for any item, but cnly a j
check mark in a."Yes, No, or N/A" category. The audits were l
performed by the former QA Manager, but there was no documented evidence to show that the individual was qualified to performed audits.
In addition, SWL had not established any criteria to qualify individuals for the performance of audits (See Noncon-formance 87-01-04).
i For the May 1985 audit conducted of Grady M. Cook, who calibrated l
the optical comparator (S/N 7018), the only documentation was a i
statement signed by the QA Manager that the vendor would be retained on the AVL for another year (See Nonconformance 87-01-07).
Of the eight calibration service vendors on the AVL, SWL had the QA l
Manual for only Southern Calibration and Services (SCS). Four of 1
the companies were audited on an annual basis, and the remaining i
four were audited every two years. Audits were overdue on SCS, Gulf i
States Balance (GSB), Grady Cook Sales, Gulf Coast Calibrating, and j
Hildebrandt Engineering.
It was also noted that SWL did not conduct audits / surveys of the service vendors-Coffer Laboratories, Page-i Wilson and Technology & Calibration.
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u ORGANIZATION:. SOUTHWESTERN LABORATORIES HOUSTON, TEXAS REPORT INSPECTION
-N0.:
99901097/87-01 RESULTS:
PAGE 6 of 13 Selected SWL P0s to vendors for M&TE calibration services were reviewed. P0s'6183 (June 11, 1987) and 440 (January 23, 1985) to GSB and P0 5542 (March 5, 1987) to Page Wilson did not invoke quality requirements nor 10 CFR Part 21..(See Violation 89-01-02 and Nonconformance 87-01-06)
SWL P0 1308 dated November 21, 1986 to Coffer Laboratories requested I
a chemical analysis for eight samples per ASTM B61. The PO was j
signed by SWL purchasing, but there was no. documented evidence that the P0 was reviewed by quality assurance. Although a Coffer l
lab report identified the work as nuclear,' the SWL P0 did not
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document. the work as nuclear, and the requirements of 10 'CFR Part 21 werenotinvoked(SeeViolation 87-01-02).
3.
Control of M&TE j
The inspector reviewed Section 12.0 of.the QAM, Procedure No.
I 9700-102-15 " Control of M&TE," and control cards to assure that' M&TE is calibrated at regular scheduled intervals. The results i
of this review are summarized below.
(See Nonconformance 87-01-13)
MTE CALIBRATED NO..
DESCRIPTION g
H
.2054K001 Gas Chromatograph Gulf Nuclear 1
2054K002 Elec. Cap. Detector Gulf Nuclear 2054K020
'Hitachi Spectrophotometer SWL 2054K029 Meter Box No. 1 SWL 2054K030 Meter Box No. 2 SWL 2054K031 Meter Box No. 3 SWL 20548002 Mettler Bal (AE160)
GSB
-20548003 Mettler Bal GSB 20540004 Torbal Bal (EA-IAP)
GSB l
20088001 Leco Elec Bal GSB i
2008P001 Sate Univ Tester SCS j
2008P008 Instron Tester SCS 2008P017 Furnace Temp Indic.
Thermo-temp i
2008P019 Tinius Olsen Tester Army.Res. Ctr 2006P030 Furnace Temp Rec.
Thermo-temp 2008P031 Wilson Hardness Tester SCS 2009P089 Rockwell Tester SCS 2008L001 Tucon Microhardness Tester Page Wilson lla t
ORGANIZATION: SOUTHWESTERN LABORATORIES HOUSTON, TEXAS REPORT INSPECTION N0 : 99901097/87-01 RESULTS:
PAGE 7 of 13 The control card for No. 2054KC30 does not show an annual calibration entry fer the July 1987 date indicating it is currently out of calibra; ion.
The control card for No. 2054B002 does not show the requ m d six month calibration entries for October 19.84 and
- Jan,
/ 1986 indicating the item was out of calibration from Oct
.r 1984 to January 1985 and January 1986 to July 1986. The card for No. 20548003 is missing a six month calibration i ntry for July 1984 indicating the item was out of calibration from uly 1984 to August 1984. The control card for No. 20548004 is missing the six month calibration entry for January 1986 indicating the item was out of calibration from January 1986 to July 1986.
For the remaining M&TE, the control card indicates calibration of the item is current. A spot check of recent calibration history on the card indicates the item had been-calibrated as required. The control card for No. 2054K020 indicates the item is calibrated prior to each use. The certification documents (e.g., document dated June 11, 1987 from GSB for the calibration of the Mettler AE160 balance) provided by SWL calibration vendors do not contain-certification statements.
(See Nonconformance 87-01-14).
i 4.
10 CFR Part 21 SWL Procedure No. 9700-108-87, " Notification Procedure 10 CFR 21" dated January 28, 1987, was reviewed and found to be inadequate in that it does not ensure that a responsible officer with executive authority will be notified in all cases when a defect is found in a safety-related basic component (see Violation 87-01-03). During an inspection of.the facilities, it was noted that the procedure was l
posted along with the regulation in three locations, but Section 206 of the Energy Reorganization Act was not posted (See Violation 87-01-01).
5.
Audits The inspector reviewed Section 18.0 of the'QAM and Procedure No. 9700-104-82 " Auditing of QA Program" pertaining to internal audits. Selected annual audits in 1985, 1986, and 1987 by a SWL vice president of the QA departnient were reviewed. The results of a review of internal audits conducted by the QA Manager are as follows; I
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ORGANIZATION:
SOUTHWESTERN LABORATORIES' HOUSTON, TEXAS' REPORT.
' INS;ECTION N0.1 99901097/87-01 RESULTS:
PAGE 8 of 13.
AUDITED-
-AUDIT D
DATE DEPARTMENT FINDINGS 01/24/86 Analytical (20-01)
Four 01/24/85 Analytical (20-01)
Five.
I 06/03/86 Fechanical (20-08)
Eight 09/18/87 NDE-20-09 One 09/19/85 NDE-20-09 Seven A checklist could not be found for the audit performed in January.
l 1986. The audit checklist did not identify the auditor fer the June l
1986 and September 1987 audits, and the audit checklist for the June 1986 audit was not dated. Corrective action was complete for all'five audits.
The QA Manager indicated internal audits had _not been performed during 1987. As a. result, a number of SWL departments have not.been
< audited within 12 months of the previous audit (See Nonconformance' 87-01-15). The file of annual QA reports to the SWL President was scanned for the past few years, and a report was found for each
- year, i
a 6.
' Documentation Packages (DP)
$WL was asked to provide a list of customers using SWL services for commercial. nuclear work.
In response, the QA Manager provided a list of. customers who require quality statements on reports and i
other test documentation.
He stated that SWL may have perforned nuclear related work for the following companies:
Lone Star Screw Gulf States Utilities NPS Industries Taylor Forge Engineered Systems (TFES)
Anderson Greenwood Texas Utilities Generating Company (TUGCO)
Stainless Products B&B Insulation WFI Nuclear Products Bechtel Power Corporation Ladish Dresser Industrial Capitol Manufacturing Southern Bolt & Fastener Public Service Company of Colorado (PSC) 120
ORGANIZATION: SOUTHWESTERN LABORATORIES HOUSTON, TEMS REPORT INSPECTION NO.: 99901097/87-01 RESULTS:
PAGE 9 of 13 It was noted that the SWL contract files are organized by customer name, and both nuclear and non-nuclear documentation were located in the same file.
' preliminary review of DPs indicated SWL performs work on blanket P0s as well as individual P0s.
A DP consisted of the customer's P0, SWL's Report of Analysis and Tests, billing summary, and invoice. The report was initialed by the department manager or supervisor ana was signed by the QA Manager if it contained a quality statement. Approximately 110 DPs were reviewed.
Fifty-five of the DPs addressed NDE activities and the remaining 55 were for chemical analysis, mechanical testing, or I
metallurgical evaluation. The NDE work was performed for Gulfalloy (seven reports), Ladish (41 reports), ano WFI (seven reports). A summary of the activities performed for nuclear customers is as follows:
Eight P0s from Anderson Greenwood and Company for mechanical testing referenced Appendix B to 10 CFR Part 50, 10 CFR Part 21 and ANSI N45.2. The eight test reports from SWL contained a quality state-ment and indicated that 10 CFR Part 21 applied.
A blanket P0 N1566 dated January 5, 1987 from WFI was to perform chemical analysis, mechanical testing, metallurgical testing, corrosion evaluation, and NDE for the calendar year 1987 in accordance with SWL's Quality Program, and 10 CFR Part 21 is applicable.
Eighteen reports from SWL contained a quality statement and that Part 21 applied.
Bechtel's P0 14926-BA-6612 dated January 15, 1986 and succeeding revisions through Revision 7 dated February 17, 1987 was to provide metallurgical and chemical testing, and Part 21 and applicable portions of Appendix B to 10 CFR Part 50 applied. The 12 SWL reports were for testing items for the South Texas Project Electric Generating Station. All the reports contained a quality statement and were signed by the QA Manager with the exception of one report (see Nonconformance 87-01-09).
P0 CPF-10881-5 dated June 17, 1983 from TUGC0 and numerous supple-ments dated tnrough June 1986 for chemical analysis referenced ANSI N45.2, Appendix B to 10 CFR Part 50, Section III of the ASME Code and imposed Part 21.
P0 CPF-13174-S dated January 31, 1986 also referenced the same documents.
Eight SWL reports which documented results of work in 1985 and 1986 for these two P0s were reviewed.
Six of the reports did not contain a quality statement and were not signed by the QA Manager (see Nonconformance 87-01-09).
121
k ORGANIZATION: SOUTHWESTERN LABORATORIES HOUSTON, TEXAS
' REPORT INSPECTION N0 : 99901097/87 RESULTS:
PAGE 10 of 13 SWL Reports 19874-1, -2 dated April 10, 1987, document results of chemical analysis of stainless steel specimens on P0 2699 from TFES. The report contains the following quality statement: " Tests have been performed per Quality Assurance Program Revision 4, dated July 15, 1983 as approved by Taylor Forge Engineered Systems, October 10, 1984 10 CFR 21 applies." The report was initialed by a reviewer and signed by the Department Manager and the QA Manager.'
P0 No. N7622 from PSC dated December 19, 1986 required a certificate of conformance for chemical analysis of specimens in accordance with ASTM standards 0608, D1662, and D1179. The P0 identified that the work-would'be related to a nuclear safety application, and 10 CFR Part El-was invoked.
SWL Report 2035 dated January 21, 1987 docu-mented the chemical analysis results. The report was signed by the Analytical Lab Supervisor and the QA Manager and contained the following quality statement:
" Services were furnished in accordance with Quality Program, Revision 4 dated July 15, 1983, approved on March 19, 1986. Public Service shall be advised of any changes in the program." Analyses Data Sheet documented the test results, time, date and contained the initials of the person performing the tests. The equipment used for the test was not identified.
PSC P0 N7043 R-04 dated August 23, 1986 requested ASME weld qualifica-tion for seven hardfaced samples. The P0 identified technical requirements for heat treatment, including a limitation of 600*F/hr over 600*F for heatup and cooldown, but it did not identify whether the work was nuclear or non-nuclear, and it did not invoke 10 CFR 21.
SWL Report 86610-3 dated September 26, 1986 for this work contains a quality statement. A review of the heat treat oven temperature chart for 5HF21-2 (specimen No. 3) indicated the heatup rate limita-tion of 600*F/hr over 600 F was exceeded. Analysis Data Sheet documented the results for the Rockwell "C" hardness tests, but the following information was not provided: date of test, name of tester, equipment and procedure used for test.
PSC P0 7043 R-08 dated November 19, 1986 requested chemical analysis of eight samples to ASTM B61. The P0 identified the work as nuclear, but it did not invoke 10 CFR Part 21. SWL Report 19295 dated December 2, 1986 identified the date, tester, test procedures, specification and equipment; and it contained a -quality statement.
The report was signed by the Department Manager and the QA Manager.
122
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ORGANIZATION: SOUTHWESTERN LA8 ORATORIES HOUSTON, TEXAS REPORT INSPECTION NO.: 99901097/87-01 RESULTS:
PAGE 11 of 13 PSC P0 N7340 dated September 22, 1986, requested laboratory services for halide testing. The P0 invokes 10 CFR 50, Appendix 8 and 10 CFR
)
- 21. SWL Report 1805 dated November 5, 1986 contains a quality statement and was signed by the Analytical Lab Supervisor and the QA Manager. Analysis Data Sheets document results of the chemical testing, but there war no information to identify the tester and the
)
equipment used.
PSC P0 N7043 R-05 dated September 16, 1986 requests analysis of l
eleven different specimens. Specimens No. 9 and 10 were identified i
as nuclear. SWL Reports 86696-1 through 86696-11 dated October 27, 1986 document results of the work. The QA Manager signed ten of the eleven reports.
PSC P0 N7043 R-04, dated July 31, 1986 requested chloride analysis of samples and was identified as nuclear. SWL Report 1308-1 dated August 13, 1986 contained a quality statement and was signed by the Analytical Lab Supervisor and the QA Manager. Analysis Data Sheet documented the analysis results, but there is no information identifying the tester ur the equipment used.
NUS P0 AUS-7674 dated December 3,1986 requests an evaluation of a weldment.
The P0 invokes SWL QA Program Revision 4, dated July 15, 1983 as well as 10 CFR Part 21. SWL Report No. 44592 dated January 23, 1987 was signed by the Department Manager and the QA Manager and contains a quality statements.
7.
Examination and Testing The inspector reviewed Sections 10 and 11 of the QAM relating to inspection, examination, and testing.
Seven NDE procedures for ultrasonic, magnetic particle, radiographic, and liquid penetrant testing to the requirements of Section V of the ASME Code were also reviewed.
The procedures were UT-109-84, MT-203-84, MT-204-85, LP-2-84, LP-101-84, LP-202-86, and RT-201-84.
Approximately 55 NDE reports to Gulfalloy (seven), WFI (seven),
and Ladish (41) were reviewed.
Three reports to WFI in 1986, seven reports to Gulfalloy in 1982 and 1983, and six reports to Ladish in 1984 and 1985 did not contain a quality statement.
In addition, a number of NDE reports for testing to Section III of the ASME Code j
were not signed by the QA Manager (See Nonconformance 87-01-09).
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1 123 I
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l ORGANIZATION: SOUTHWESTERN LABORATORIES HOUSTON, TEXAS REPORT INSPECTION NO.: 99901097/87-01 RESULTS:
PAGE 12 of 13 i
Some of the test documentation did not include all the 'rformation required by the QAM. Complete traceability for testing *cs not provided because equipment used for the testing was not recorded:
Analysis Data Sheet No. 2035 dated January 21, 1987; Analysis Data Sheet No.1805 dated November 5,1986; and Analysis Data Sheet No. 1308-1 dated August 13, 1986.
The name of the individual performing the test was not recorded for Analysis Data Sheet No.1805 dated November 5,1986; and Analysis Data Sheet No.1308-1 dated August 13, 1986.
(See Nonconformance 87-01-12.)
It was noted that test report Nos. 2035 (January 21,1987),1805 (November 5, 1986), and 1308-1 (March 13, 1986) issued by the Analytical Laboratory were reviewed and signed by the Analytical Lab Supervisor rather than the Department Manager.
8.
Training The inspector reviewed Section 1 of the QAM and Procedure No. 9700-105-82, " Indoctrination of Personnel in Company Policies and Quality Assurance Procedures."
It was noted that all personnel engaged l
in testing and inspection activities are required to participate l
in QA policy and procedure review sessions on an annual basis. The training and certification records for selected personnel in the AS, MTM, and NDE departments were reviewed.
Records for six personnel from AS,11 from MTM, and 11 from NDE indicated that one individual from AS and two individuals from MTM were not trained or indoctrinated in the QA Program.
Certifications and tests to the requirements of ANSI N45.2.6 were missing for two individuals from MTM.
It was also noted that form SWL-105-2 was not used to document the 1 indoctrination and QA training of 11 individuals for NDE, four from AS, and nine from MTM.
(See Nonconformance 87-01-08.)
The qualification records for the 11 NDE personnel indicated that each individual was certified to the requirements of SNT-TC-1A.
In addition, four examiners who were certified to perform radiography had received radiation safety training.
Each file contained a radiation safety examination, a certificate signed by the Radiation Protection Officer, and a letter from the Texas Department of Health indicating that the individual had passed the industrial radiography examination required by the Texas Regulations for Control of Radia-tion.
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l 0 ORGANIZATION: SOUTHWESTERN LABORATORIES HOUSTON, TEXAS REPORT INSPECTION l
N0.: 99901097/87-01 RESULTS:
PAGE 13 of 13 f
The. Radiation Protection Officer, who is also the NDE QC Manager,
.i had conducted NDE and Radiation Safety Audits on an annual basis from 1983 of the four SWC field offices located in Austin, Beaumont, i
Dallas, and Texas City.
j E.
PERSONNEL CONTAClED:'
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- J. Jeanes, Chief Executive Officer l
- D. Hanna, QA Manager
- R. Richter, Director of Metals Testing M. Tipton, Supervisor, Analytical Laboratory L. Lilly,.QA Manager (prior to March 1987)
- Attended' exit meeting.
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I u______________________
SSINS No.: 6835 IN 87-32 UNITED STATES NUCLEAR REGULATORY C0t941SSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C.
20555 July 10, 1987 NRC INFORMATION NOTICE NO. 87-32: DEFICIENCIES IN THE TESTING OF NUCLEAR-GRADE ACTIVATED CHARC0AL Addressees:
All nuclear power reactor facilities holding an operating license or a con-estruction pennit.
Purpose:
This information notice is provided to call attention to deficiencies found in the testing of nuclear-grade activated charcoal used for accident mitigation in l
nuclear facilities.
It is expected that recipients will review the information for applicability to their ncilities and consider action, if appropriate, to l
preclude a similar problem at their facilities. However, suggestions contained in this information notice do not constitute NRC requirements; therefore, no specific action or written responso is required.
Background:
The ASME Comittee on Nuclear Air and Gas Treatment (CONAGT) first identified a problem with testing nuclear-grade activated charcoal when the comittee conducted an interlaboratory comparison and found that seven U.S. and eight foreign testing companies obtained vastly differing results from testing samples of the same charcoal. After efforts to resolve the differences failed, the NRC contracted with the Idaho National Engineering Laboratory (EG&G) to investigate the problem. The contractor has conducted independent hboratory studies and has worked with the testing companies to identi#y the principal problems. Serious problems were found with the capabilities of the testing companies and with the testing standard fASTM Standard D-3803-1979, " Standard Methods for Radiotodine Testing of Nuclear-Grade Gas-Phase Adsorbents"1 Specific suggestions were made to improve the capabilities of the testing companies; a new testing protocol was developed to correct shortcomings found in the standard; and a final interlaboratory comparison was conducted.
The results, which were received in~ November 1986, indicate a substantie1 improvement over the original CONAGT results, but some companies still did not report acceptably accurate results. The contractor's technical evaluation report, EGG-CS-7653, " Final Technical Evaluation Report for the NRC/INEL Activated Carbon Testing Program," has been published and has been placed in the NRC Public Document Room.
8707070003 127
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IN 87-3?
July 10, 1987.
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Page 2 of 2 Discussion-1 Engineered safety features, including charcoal, are provided ~at ' nuclear power j
plants to protect employees and the public from accidentally released radio-.
]
active materials. 'It has been recognized that only certain charcoals would; meet the special needs of nuclear plants (see NUREG/CR-3990. " Charcoal. Perform-i
. ance Under Accident Conditions"). ASTM Standard D-3803-1979 was developed to i
specify the requirements for testing charcoal and was accepted by the NRC
_l
.(' Regulatory Guide 1.52,1979). The NRC investigation, which followed the f
CONAGT interlaboratory comparison, identified serious shortcomings in the j
standard and found that it had never been verified. The. standard is currently-theing revised. ' Nowever, until it is available the protocol developed by EG&G j
is one possibility for consideration by~ testing companies.
,1 Although shortcomings in testing capabilities were identified by EG8G, defi-
-f ciencies can be corrected only by the individual companies. The failure to upgrade equipment.so the test parameters car, be adequately controlled is-the principal reason for the unacceptable-results in the recent-tests.
Additional in*ormation on test accuracies and changes made to improve the l
accuracy of test results may be sought by direct contact with the. individual testing companies.
No specific action or written response is required by this in'ormation notice.
If you have any cuestions about this matter, please contact the Regional Administrator of the appropriate regional office or this office.
M Charles E. Rossi, Director Division of Operational Events Assessment.
Office ~of Nuclear Reactor Regulation Technical Contacts: Charles A. Willis, NRR I301)492-8340 Charles R. Nichols, NRR (301) 492-9416
Attachment:
List of Recently Issued NRC Information Notices I
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128 b
IN 87 July 10, 1987 1
' t.!$T;0F'RECENTLY !$$UED 5
'INFORMATION NOTICES 1987.
j l
Notice No.
Subject.
Date of Infomation Issuance -
Issued to
- 31 Blocking, Bracing, and.
7/10/87
-All NRC licensees.
- Securing of Radioactive
' Materials Packages in l
Transportation.
87-30
. Cracking of Surge Ring-7/2/87 All nuclear power Brackets in Large General.
reactor facilities-Electric Company Electric
. holding an OL or Cp.
. Motors 87-29.
Recent Safety-Related
.6/26/87 All NRC licensees.
i Incidents at Large.
authorized to' possess
.Irradiators, and use sealed sources in large irradiators.
1 87-28 Air Systems Prob 1rms at 6/22/87 All nuclear power l
U.S. Light Water Reactors reactor facilities holding an OL-or CP.
87-27 Iranian Official Implies 6/10/87 All nuclear power Vague Threat to U.S.
febetor facilities Resources holding an OL or CP, research and nonpower i
D reactor facilities, and fLe1 fabrication
- and processing facilities using or possessing formula quantities of special nuclear material.
87-26 Cracks In Stiffening Rings 6/11/87 All uranium 'uel on a8-Inch Diameter UF fabrication and 6
Cylinders.
conversion facilities.
87-25 Potentially. Significant 6/11/87 All nuclear power i
Problems Resulting from-reactor facilities
{
Human Error Involving holding an OL or CP.
Wrong Unit, Wrong Train, or Wrong Component Events.
87-74 Operational Experience 6/4/87 All nuclear power Involving Losses of reactor facilities
.l Electrical Inverters.
OL = Operating License CP = Construction Permit i
129
SSINS No.: 6835 IN 87-33 UNITED STATES NUCLEAR REGULATORY C09981SSION 0FFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C.
20555 July 24, 1987 NRC INFORMATION NOTICE NO. 87-33:. APPLICABILITY OF 10 CFR PART 21~
TO NONLICENSEES
-Addressees:
All NRC licensees.
Purpose:
L This notice is provided to inform addressees of a potential generic problem.
. concerning the applicability of 10 CFR Part 21 to certain fabrication, erec-tion', installation, modification, inspection, testing, and training services It is expected that recipients will review the informa--
provided to licensees.
tion for applie-bility to their facilities and consider actions, if appropri-However, suggestions contained in this ate, to prec1: c timilar. problems.
information notice do not constitute NRC requirements; therefore, no specific
^
action or written response is reoufred.
Description of Circumstances:
Durthg' recent NRC inspections at the Byror, Station, Units 1 and 2, it was noted.
that the licens:e, Commonwealth Edison Company-(CEcol, had failed to appropri-ately apply the requirements of 10 CFR Part 21 to onsite contractors.. These contractors provided fabrication. emetion,' installation, modification, mafate-nance, inspection, and testing services for the Byron facility. During inter-views with senior CECO management, the NRC was infomed that CECO had made the j
decision not to specify applicability of the requirements of 10 CFR Part 21 for any of their onsite contractors. During a separate NRC inspection, it was also
-f noted that Iowa Electric Light and Power Company (IELP) had failed to appropri-stely apply the requirements of 10 CFR Part 21 to a local vendor that had-repaired / rewound Class IE electrical equipment for the Duane Arnold Nuclear Power facility. IE Information Notice 85-101, " Applicability of 10 CFR 21 to
)'
Consulting Fims Providing Training," was previously issued on December 31, 1985, to infom licensees and consultants of a potontial generic problem concerning the applicability of 10 CFR Part 21 to certain training activities provided by consultants. ' The notice was issued after an NRC vendor inspection of a company which provided consulting services, including training, to the nuclear industry disclosed that licensees had failed to appropriately apply the requirements of 10 CFR Part 21 to the consultant providing the training.
8707200145
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5 IN 87-33 July 24, 1987.
.Page 2 of 3 Discussion:
The Commission is taking this opportunity to emphasize the responsibilities of licensees,under the requirements of 10 CFR Part 21, " Reporting of Defects and Noncompliance." 10 CFR Part P1 establishes procedures and requirements for the
' implementation of Section 206 of the Ener3y Reorganization Act of 1974,'as
-' amended. - Section 206 imposes reporting responsibilities on' directors and-responsible officers of. firms constructing, owning, operating or supplying '
the components of any facility or activity which is licensed or otherw',se-regulated pursuant to the Atomic Energy Act of 1954, as amended, cr the Energy.'
Reorganization Act of 1974, as amended. ' The focus of Section 206 goes beyond those entities licensed or regulated by the Comission to all entities that engage in the activities described in the regulation.
10 CFR'21:3(c) s^ates that the terms " constructing" or " construction" include the design, manufacture, fabrication, placement, erection, installation, modification,l inspection,' or testing of a facility or activity that is sub.iect to 10 CFR Part 21 and consulting services which are related to,the facility or activity-that are important to safety. 10CFR21.3(1)statesthattheterms
" supplying" or." supplies".means contractually responsible for a basic component used or to be used in a facility or activity sub,4ect to 10 CFR Part 21, 10 CFR 21.3(a)(3) states that "in all cases ' basic component' includes design, inspection, testing,' or consulting services important to safety that are i
. associated with the component hardware, whether these services are performed by the component supplier or others." Hence, onsite and offsite construction services are subject to the provisions of 10 CFR Part 21 whenever these services j
are associated with a basic component as defined in 10 CFR 21.3(a)fli. Although 10 CFp Part 21 obligations are applicable, whether or not the contractor has been contractually oblicated to the provisions of 10 CFR Part 21, licensee
. procurement of such services should nevertheless specify the applicability of 10 CFR Part 21 as stated in' 10 CFR 71.31 unless such services fall under the definition of "comercial grade item" as defined in 10 CFR 71.3(a)(41 Any deviation discovered following the suitability for application review performed to dedicate a' commercial grade item for:a safety-related application would be reouf red to be evaluated by the dedicating or subsequent user organi-zation and, if appropriate, reported pursuant to the requirements of 10 CFR l
Part 21.
Further discussion and guidance on this matter is provided in NUREG-030?,
Revision 1, " Remarks Presented (Questions / Answers Discussed) at Public Regional
. Meetings to discuss Regulations (10 CFR Part 21) for Reporting of Defects and Noncompliance," published in October 1977.* This publication provides NRC staff
- remarks on 10 CFR Part 21 as well as some legislative and legal discussions-on 10 CFR Part 21 and its impact upon reactor, material, fuel cycle, and export licensees and related suppliers.
- NUREG-0302, Rev. 1, is available through the NRC/GPO Sales Program, U.S. NRC, Washington, D.C.
20555, (20?) 275-2060.
4 i
132 4
v.
IN 87-33 July 24, 1987 Page 3 of 3 No specific action or written res ponse is required by this information notice.
If you have any questions about tiis matter, please contact the Regional
' Administrator of the appropriate regional office or this office.
i f4 $ k_~'J' Charles E. Rossi Director o
Division of Operational Events Assessment Office of Nuclear Reac'ar Regulation Technical Contacts: Jaime Guillen NRR (301) 492-8933 M. P. Phillips RIII (312) 790-5530 R. S. Love, RIli (312) 790-5593
Attachment:
List of Recently Issued NRC Infomation Notices I
i l
133
I SSINS No.: 6835 IN 87-35 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C.
20555 July 30, 1987 NRC INFORMATION NOTICE NO. 87-35: REACTOR TRIP BREAXER, WESTINGHOUSE MODEL DS-416. FAILED TO OPEN ON MANUAL INITIATION FROM THE CONTROL ROOM Addressees:
All nuclear power reactor facilities holding an operating license (OL) or construction permit (CP) employing Westinghouse 05-416 reactor trip breakers.
Purpose:
This notice is provided to alert recipients to a potentially significant safety problem associated with a reactor trip breaker (RTB). The NRC expects that recipients will review this notice for applicability to their facilities and consider actions, if appropriate, to preclude a similar problem. However, suggestions contained in this notice do not constitute FDC requirements; therefort, no specific action or written response is recuired.
~
Description of Circumstances:
On July 2,1987 McGuire Nuclear Station Unit 2 was perfonning control rod drop tests after its recent refueling outage. This test was in progress with the plant in mode 3 (hot shutdown). With all control rods inserted and the RTBs closed for testing the next bank of control rods, station personnel smelled smoke in the vicinity of the RTBs. A manual trip of A and B train RTBs was initiated from the control room. Only the A train RTB opened. The B train RTB was eventually tripped manually at the breaker panel. The smoke had come from the B train breaker shunt trip coil, which had burned and shorted while trying to open the breaker. The coil is designed for intermittent duty and to carry current only until the breaker opens. Failure of the breaker to open resulted in a prolonged and damaging current. Operators in the control room stated that open indications for both the A and B train redundant RTBs were observed for all attempted breaker opening evolutions during the control rod drop testing process. However, the event recorder indicated that the B train pTB failed to open on a previous manual trip attempt (approximately 4 minutes bare) when operators were r,etting up for the control rod drop test on the Itu bank of rods.
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An.NRC Augmented Inspection Team (AIP evaluated the ifcensee's investigation i
into the reactor trip breaker problem. ~ Abnormal wear and a < broken weld were-Figure. Attachment 1).. The Lroken weld was on the main drive link between the
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found in this early vintage of Westinghouse 05-416 breaker (seb1mtinghouse-o center pole lever and the pole shaft.,Except for the shunt trip coil that had t
V burned and shorted while trying to open 'the breaker, the breaker's electrical 6
1 controls and auxiliary contacts wre verified to,be properly wired and operat-m ing as designed. The cause fot: the anomalous breaker status indication is g; y still under investigation.
y Attempts to' repeat the condition, where the breaker was mechanically binding in the closed position, were minimally successful. Preliminaryconclusionsofthe
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ATT. are that.the breaker's mechanical bindin wear (greater than 2000 cycles of ~ operation)g was caused by a combination.of
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early vintage breaker, and the broken weld. These factors may have combined to allow sufficient lateral movement of the main linkage to cause it to jam at or o
near full breaker closure _ and thus prevent the breake? frcm opediing. Since the k
6 control room operating personnel stat'ed that they obser' ed the open indication h
v on the closed B train RTB, the field wiring is being verified by the licensee i
to ensure that wiring is as designed. The shorted shunt trip coil had allowed
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(i 125 volts de between the positive terminal and the chassist a "sg'eak" circuit is possible,
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Discussion:
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Final conclusions for the cause of this event have not been reached. Further 9 investigation and dismantling of the breaker will be conducted in Westinghouse laboratory facilities. The licensee and NRC will participate in this investi-t gation. If the results indicate findings different than the above preliminary e
conclusions, a supplement to this notice will be issued.
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The licensee is inspecting all of the RTBs Jor signs ef'oknonnal wear, cracks 1
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in welds, and excessive lateral play.(greater than.1/0 in'ch) in the roller end t
of the main drive link where it contacts the close cain. This measurement had
^
not previously been part of the periodic preventive maintenance for the RTB.
P Moreover, following any reactor trip, the licensee is ensuring the open posi-tion of both RTBs by inspecting the breaker before reclosure. These are short-term corrective actions until the detailed analysis of'the deficiencies is completed.
l A significant number of generic communications have been issuedmithregard to reactor trip breakers (RTBs) and similar c3rcuit breakers used in safety-b' related systems. Such connunications that may be related toihe matter in this information notice are listed in Attachment ?.i,
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If ycu have any questions about this matter, please contact the Regional f\\
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L-. j /' f Administrator of the appropriate NRC. regional office or this office.
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s my harles E. Ross, Director Division of Operational Events Assessment
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. Ezr.arie Communications on Reactor Trip Breakers and Similar Circuit i2.
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July 30, 1987 j
Page 1 of 1 j
i GENERIC COMMUNICATIONS ON REACTOR TRIP BREAKERS AND SIMILAR CIRCUIT BREAKERS Information Notice 86-62, " Potential Problems in Westinghouse Molded Case Circuit Breakers Equipped With a Shunt Trip," July 31, 1986.
t Infomation Notice 85-93, " Westinghouse Type DS Circuit' Breakers, Potential Failure of Liectric Closing Feature Because of Broken Spring Release Latch lever," December 6, 1985.
Bulletin 85-02, "Undervoltage Trip Attachments of Westinghouse 08-50 Type Reactor Trip Breakers," November 5,1985.
Information Notice 85-58, " Failure of a General Electric Type AK-2-25 Reactor Trip Breaker," July 17, 1985.
Supplement 1, November 19, 1985.
Information Notice No. 83-76, " Reactor Trip Breaker Malfunctions (undervoltage as Trip Devices on GE Type AK-2-25 Breakers)," November 2,1983.
l
'[
Generic Letter 83-28, " Required Actions Based on Generic Implications of Salem l
ATWS Events," July 8, 1983.
b~'
Information Notice 83-18. " Failures of the Undervoltage Trip Function of I
Reactor Trip System Breakers," April' 1,1983.
{l Bulletin 83-04, " Failure of the Undervoltage Trip Function of Reactor Trip Breakers," March 11, 1983.
g Bulletin 83-01, " Failure of Reactor Trip Breakers (Westinghouse DR-50) to Open on Automatic Trip Signal," February 25, 1983.
Circular 81-12. " Inadequate Periodic Test Procedure of PWR Protection System,"
j July 22, 1981.
Bulletin 79-09, " Failures of GE Type Circuit Breaker in Safety Related System,"
April 17, 1979.
1 139
- o
IN 87-35 July 30, 1987 LIST OF RECENTLY ISSUED INFORMATION NOTICES 1987 Information Date of Notice No.
Subject Issuance Issued to 87-34 Single Failures in Auxiliary 7/24/87 All holders of an Feedwater Systems OL or ~a CP for l
pressurized water reactor facilities.
87-33 Applicability of 10 CFR 7/24/87 All NRC licensees.
Part 21 to Nonlicensees 87-3?
Deficiencies in the Testing 7/10/87 All nuclear power of Nuclear-Grade Activated reactor facilities Charcoal.
87-31 Blocking, Bracing, and 7/10/87 All NRC licensees.
Securing of Radioactive Materials Packages in Transportation.
87-30 Cracking of Surge Ring 7/2/87 All nuclear power Brackets in large General reactor facilities Electric Company Flectric holding an OL or CP.
Motors.
87-29 Recent Safety-Related 6/26/87 All NRC licensees Incidents at Large authorized to possess Irradiators.
and use sealed sources in large irradiators.
87-28 Air Systens Problems at 6/PP/87 All nuclear power U.S. Light Water Reactors.
reactor facilities holding an OL or CP.
87-27 Iranian Official Implies 6/10/87 All nuclear power Vague Threat to U.S.
reactor facilities Resources, holding an OL or CP, I
research and nonpower l
reactor facilities, and fuel fabrication and processing facilities using or possessing formula quantities of special nuclear material.
OL = Operating License CP = Construction Permit 140
f 1
SSINS No.: 6835.
' i '
UNITED STATES l
NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON. D.C.
20555-August 31, 1987 NRC INFORMATION NOTICE N0. 87-40: BACKSEATING VALVES ROUTINELY TO PREVENT PACKING LEAKAGE Addressees:
All nuclear power reactor facilities holding an operating license or a con-struction permit.
Purpose:
This information notice is provided to alert recipients to potentially sig-i nificant safety problems that could he caused by.backseating valves routinely to prevent packing leakage. It is expected that recipients will review the information for applicability to theif facilities and. consider actions, if '
appropriate, to preclude a similar problem. However, suggestions contained g
in this notice do not constitute NRC requirements; therefore, no specific l
action or written response is required.
Description of Circumstances:
On June 12,.1987, Virginia Electric and Power Company reported (Licensee Event Report [LER] 87-011-00) that on May 16, 1987, a low flow reactor trip occurred at its Surry Power Station, Unit 1. 'The c'ause of the low flow was the failure of the stem of the A hot leg loop stop valve. The stem failure permitted the disk to drop and partially block flow in the A loop. The licensee is performing j
a detailed metallurgical examination to determine the failure mode'and mechanism of the valve stem. The preliminary report indicated that failure was due to stress or fatigue.
Before this event, the licensee had routinely backseated the loop stop valves as part of its Containment Checklist Procedure before startup. In accordance with this procedure, the valves were manually torqued onto their backseats to 1/16-inch deflection while the unit was in cold shutdown. This value was reverified when the unit reached a hot shutdown condition. To reduce the stress on the valve stem, the licensee is revising the operating procedure so that the valves are normally operated off the backseat.
A similar event had occurred with the 8 hot leg loop stop valve on December 1, 1973. This failure was evaluated by Westinghouse, and a failure report, "Surry Unit No. 1 Reactor Coolant Isolation Valve Stem Failure Report," was issued on March 7, 1974. In the report, the f ailure mechanism was ' identified as a high 8708250140 i
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4 IN 87-40 August 31, 1987 Page 2 of 4' r
strain', low-cycle failure with little-deformation. The' high strain was at-
~tributed.to'the licensee's practice of routinely ~ electrically backseating the valves on torque during plant startup. 'In this method, the. valve's motor
. operator:is used to drive the valve open until'the forces resulting from the disk-pressing'against the backseat'are high enough to cause.the open torque switch to open. This' practice caused the valve stem to'be subjected to'high-stresses each time ~the valve was opened. Depending on the gearing within the motor operator, these stresses may.have remained 'until.the valve was closed-
- during the next plant shutdown.
The report recommended that the valves not be electrically backseated on torque.
It further recommended that if.'backseating became necessary during maintenance, it should be done manually by the use of the handwheel with minimum applied load r
s
.and without exceeding the compensating spring pack deflection specified in the manufacturer's' revised instruction manual. This manual, " Instruction Manual.
' Motor Operated Reactor Coolant 30" Loop Stop Valves for Reactor Coolant System i
Westinghouse WNES 546-CAK-704978 Darling Valve S.0.
E-5004,"' states in a cau-l tion that manual backseating is permissible only if the open deflector indicator:
I reading does not' exceed 1/16-inch maximum and that' manual.backseating may be used~only when the packing needs replac,ement. As noted above, contrary to this recommendation, it had been the.llcensee's recent policy to routinely manually
.l
. backseat these valves'during plant startup.
~
q Before the.1973. valve. stem failure,;it had been theilicensee's. practice tc' routinely electrically backseat the valves during plant startup.. Although' j
Virginia Electric and Power Company had stopped this practice in 1974, the NRC staff is' aware that other licensees routinely electrically backseat valves.
The most common reason for this is to stop valve leakage from around the stem
.on valves that are not readily accessible during plant operations.
The following~ inspection reports reflect current practices and problems related to electrical backseating.-
(1); NRC Inspection Report 50-321/85-34 provides the results of an inspection
'at Edwin I. Hatch Nuclear Plant, Unit 1, performed between November 10 and
. December 20, 1985.
In the report, the inspectors noted observing informa-tion tags that indicated that the reactor core isolation cooling (RCIC) system inboard steam line isolation and the reactor water cleanup (RWCU) system inboard isolation valves had been' electrically backseated. The operating personne1~ indicated that they were not aware that any testing had been done to verify that the valves would close within the required time limits. A violation was issued when subsequent testing of the RWCU valve demonstrated that it could not close within the required time limits.
Additional testing of the RWCU valve showed that it did meet the closure time limits when it was not starting from a backseated condition.
NRC Inspection Report 50-321/86-22 and 50-366/86-22 provides the results of an inspection performed at Hatch Units 1 and 2 between July 28 and August'1, 1986.
In the report, the inspectors noted that the licensee routinely electrically backseated containment isolation valves in the high pressure coolant injection, RCIC, RWCU, and recirculating pump systems.
{
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IN 87-40 August 31, 1987 i
I Page 3 of 4 Review of maintenance records indicated that the valves had been electri-cally backseated as many as 18 times during the previous 2 years. The procedure used by the licensee to electrically backseat the valves con-sisted of bypassing the open limit switch and then driving the valve disk onto the backseat until the locked motor current of the motor was ap-proached.
l (2) NRC Inspection Report 50-277/86-25 provides the results of an inspection j
at Peach Bottom Atomic Power Station, Unit 2, performed between December 8 1
and 19, 1986. In the report, the inspectors noted the licensee's practice of electrically backseating valves suspected of having excessive packing leakage. The procedure used by the licensee involved an operator manually J
closing the motor contacts at the motor control center. The contacts were
]
held closed until the operator noticed an increase in the motor current j
shown on a clamp-on amp meter, q
Discussion:
General Electric Company's Service Information Letter (SIL) 385 issued November 1982 discusses potential valve damage. and provides recommendations for motor-operated valves that are normally backseated or are subjected to excessive backseating torque. This describes the types of damage that could be the result of backseating as valve stem failure, valve stem elongation, backseat damage, cracking of the stem nut, and other related component distress 'It also notes that most damage progresses slowly to the point where valve oper-ability is uncertain or valve failure occurs, before the damage is apparent.
Among the recommendations were:
2 (1) identification of the valves that are normally backseated or that may have had excessive backseating torque applied and establishment of a program for evaluation, inspection, and repair of these valves (2) consultation with valve and motor operator vendors to establish any procedures or modifications that could uinimize damage, including the use of torque switches or other vendor-recommended alternatives to backseating motor-operated valves.
The inspectors also indicated that the licensees for the Hatch and Peach Bottom plants are taking actions to prevent recurring packing' leaks. Georgia Power Company is instituting a program at the Hatch units to replace the packing on these valves at regular intervsis and also plans to live load the packing to j
help prevent packing leakage. At Peach Bottom, Philadelphia Electric Company is modifying the packing gland area, replacing the existing asbestos-based packing with graphite packing, installing carbon bushings, and live loading the packing.
j The information herein is being provided as an early notification of a possibly significant matter that is still under consideration by the NRC staff. If NRC evaluation so indicates, further licensee action may be requested.
)
143
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4 IN 87-40 August 31, 1987 Page 4 of 4 h
No specific action or written response is' required by this information notice.
If you have any questions about this matter, please contact-the Regional Administrator of.the appropriate regional office or this office.
1
/
Charles E. Rossi, Director 1
Division of Operational Events Assessment Off. ice of Nuclear Reactor Regulation Technical Contacts: George A. Schnebli, Region II (404) 331-5582 Larry E. Nicholson, RI, Surry Power Station l
(804) 357-2102 Keith Poertner, RI, ldwin E. Hatch Nuclear Plant
~
(615) 842-8001 Richard J. Kiessel, NRR (301) 492-9605
Attachment:
List of Recently-Issued Information Notices i
4 144
SSINS No.: 6635-IN 87-41 UNITED STATES HUCLEAR REGULATORY COMMISSION
. OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C.
20555 August 31, 1987 f
HRC INFORMATION NOTICE NO. 87-41: FAILURES OF CERTAIN BROWN 00 VERI ELECTRIC CIRCUIT BREAKERS j
[
Addressees:
All nuclear power reactor facilities holding an operating license or a con-l struction permit.
1
Purpose:
l This information notice is provided to inform recipients of failures of certain j
Brown Boveri Electric (BBE) circuit breakers.
It is expected that recipients will review the information for applicability and consider actions, if appro-priate, to preclude similar problems from occurring at their facilities. Sug-gestions contained in this notice do not constitute NRC requirements; therefore, no specific action or written response is required.
Description of Circumstances:
Concerns regarding BBE circuit breakers have been recently reported to the NRC.
On April 20, 1987, Duquesne Light Company, the Beaver Valley Unit 2 licensee, notified the NRC in accordance with the reporting requirements of 10 CFR 50.55(e) of the failure of a BBE Type SHK Class IE 4-KV circuit breaker. When the circuit breaker was racked onto the bus and 125-V DC control power was applied to the breaker's control circuit, the closing spring charged and the circuit breaker immediately closed and opened several times before the control power could be turned off. The licensee determined by field testing that the closing coil was not being energized.
Another problem with BBE circuit breakers occurred at River Bend and was reported in Licensee Event Report (LER)87-004, dated March 6, 1987. On February 6, 1987, with the unit at full power, the Division I diesel generator 4.16-KV output circuit breaker (Gould-Brown Boveri Type SHK) failed to close during a weekly surveillance test. The licensee's inspection of the output circuit breaker revealed that a mounting bolt had fallen out of the closing spring charging motor, rendering the motor inoperable. Further investigation
{
revealed several other circuit breakers that contained loose or missing charg-ing motor mounting bolts. The licensee also stated that the River Bend circuit breaker preventive maintenance program, which the licensee believes to be in accordance with the vendor's recommendations, did not detect this problem. The licensee believes the root cause of the problem to be insufficient torquing of the charging motor mounting bolts by the vendor.
l 8708250191 l
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IN 87-41 August 31, 1987 Page 2 of 2 Discussion:
With regard to the repeated closing and opening of the breaker, this problem
(
was also reported to the NRC in accordance with the reporting requirements of 10 CFR 50.55(e) by Philadelphia Electric Company, for timerick Unit 1, in 1983.
The circuit breakers involved were returned to BBE for analysis, and BBE transmitted its fiadings to the NRC in a letter. dated April 20, 1983 BBE stated that the repeated closing and opening of the circuit breaker can be corrected by the addition of a light spring to the close latch in the ci,rcuit breaker operating mechanism. 'BBE also stated that, depending upon the year of manufacture, circuit breakers may or may not have been originally supplied I
with this spring. BBE recommended that these springs be installed in circuit breakers that were not originally supplied with one. This problem may also affect BBE Type 7.5HK, 15HK, and 38HK circuit breakers. Although information regarding this problem has been distributed by BBE and the Institute for Nuclear Power Operations (INPO), the Beaver Valley event indicates that this problem still exists. Licensees and applicants may wish to contact BBE to obtain ad-ditional information regarding this problem and appropriate corrective actions.
4 Failure of a circuit breaker to fully close on demand, inadvertent closure of a circuit breaker, or loose or missing closing spring charging motor mounting j
bolts could result in a circuit breaker not performing its intended function.
I This, in turn, could result in the loss of a power supply (such as a vital bus or a diesel generator).
No specific action or written response is required by this information notice.
If you have any questions about this matter, please contact the Regional Administrator of the appropriate regional office or this office.
{
arles E. Rossi, Director 1
Division of Operational Events Assessment Office of Nuclear Reactor Regulation Technical
Contact:
Jack Ramsey, NRR (301) 492-9091
Attachment:
List of Recently Issued NRC Information Notices I
146
,r
['
1 Attachment IN 87-41 August 31, 1937 1
LIST OF RECENTLY ISSUED INFORMATION NOTICES 1987 f
Information Date of Notice No.
Subject.
Issuance Issued to-40' Backseating' Valves. Routinely L8/31/87 All nuclear power f o Prevent. Packing Leakage reactor facilities t
87-39.
Control of Hot Particle.
8/21/87 All nuclear power Contamination at Nuclear reactor facilities.
Power Plants'.
and spent fuel storage facilities holding an NRC license or CP.
87-38
< Inadequate or-Inadvertent 8/17/87 All nuclear power Blocking'of Valve Movement reactor facilities holding an OL or CP.
87-37 Compliance'with'the General 8/10/87 All persons spec'ifi-License' Provisions of cally. licensed to 10 CFR Part'31 manufacture or to
' initially transfer devices containing radioactive material to general licensees, as defined in 10 CFR Part 31.
87-35 Significant' Unexpected 8/4/87 All nuclear power Erosion of Feedwater Lines reactor facilities holding an OL or CP.
87-35 Reactor Trip Breaker, 7/30/87 All nuclear power Westinghouse Model 05-416, reactor facilities
{
Failed to.0 pen on Manual holding an OL or i
Initiation from the Control CP employing W D5-416 Room reactor trip breakers, j
l 87-34 Single Failures in Auxiliary 7/24/87 All holders of an Feedwater Systems OL or a CP for j
pressurized water reactor facilities.
1 87-33 Applicability of 10 CFR 7/24/87 All NRC licensees.
Part 21 to Nonlicensees OL = Operating License CP = Construction Permit 149
i SSINS No.: 6835 IN 87-42 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C.
23555 September 4, 1987 NRC INFORMATION NOTICE NO. 87-42: DIESEL GENERATOR FUSE CONTACTS-Addressees:
All nuclear power reactor facilities holding an operating license or a con-struction permit.
Purpose:
This information notice is being provided to alert recipients to potential failures of emergency electrical power supplies resulting from misalignment and/or. degradation of fuse contacts. It is expected that recipients will review the information for applicability to their facilities and consider actions, if appropriate, to avoid similar problems. However, suggestions contained in this ratice do not constitute NRC requirements; therefore, no l
specific action or written response is required.
Description of Circumstances:
At Browns Ferry Nuclear Plant, Unit 3, during routine surveillance testing of an espergency diesel generator on April 20, 1987, an explosion occurred in the electrical control cabinet. The explosion resulted from a phase-to phase short in a. cable bundle in the potential transformer (PT) fuse compartment.
The electrical control cabinet was supplied by Power Systems Division of General Motors, the diesel generator manufacturer. The licensee's investiga-tion revealed that the cables routing power from the PT fuses to the trans-formers shorted after their insulation failed because' of an overtemperature condition. This condition is believed to have been caused by an excessive air gap and/or poor contact between the spring finger contact arrangement of-the.PT fuses. The spring finger contacts were manufactured by Allis-Chalmers.
The moveable contacts are mounted to the PT fuse compartment door so that when the door is opened, the contacts disconnect, thereby preventing anyone from pulling the fuse while it is carrying current. When the door is closed, the moveable coratact finger connects with the stationary contact. Although the contacts were destroyed by the event, the licensee concluded that an air gap due to misalignment or contact degradation from corrosion, pitting, or burning, caused arcing and eventual failure. Although the cause has been attributed to inadequate maintenance of the contact, the licensee is considering design changes as a permanent solution.
8708310006 149
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IN 87-42 :
' I September 4, 1987 l
Page 2 of 2 Discussion:
The licensee inspected the other 1reven diesel generator control cabinets and found three other instances'of poor contact. Visual inspection for' burning, 1
pitting, corrosion and. improper alignment and mating of contact surfaces were J
~
performed as w il as micro-ohameter, checks. Since 1973, four previous onsite f
events involving this type of pressure contact arrangement have occurred on
{
non-safety-related electrical boards designed by General Electric Co.-
Because of this type of problem,' General Electric modified similar contacts on its--
electrical boards to knife switch contacts. The' licensee considers the problem to be generic.and has initiated a visual-inspection of all PT fuse compart-ments.
1 1
No specific' action or written response ~is required by this information notice.
4 If'you have any questions about this matter, please contact the Regional
]
' Administrator of the appropriate regional office or this office.
harlesE.Ro[ssi, Director Division of Operatior.a1 Events Assessment Office of Nuclear Reactor Regulation
' l Technical Contacts:
C. Brooks, RII-
.(205) 729-6196 i
l Samuel D. MacKay, NRR (301) 492-8394
Attachment:
List of Recently Issued NRC Information Notices l
Lo 150
+-
(,.
e r
r p
t' 1
1 L
Attachment-
~
IN 87-42~
.i o4
. 1.
(September 4, 1987 LIST OF RECENTLY ISSUED X
' INFORMATION NOTICES 1987-
-t :
i f
~
Date of 1
- p.
,-Information
. Subject
= Issuance
' Issued to:
7
' Notice No.
1 - !
. B
/ /7 3 31 8 '
All nuclear power
- 87-41'-
Failuresvof Certain rown
- Bovari Electric Circuit
' reactor facilities:
l
,t.
s 87-40 Backseating Valves 1 Routinely '8/31/87 A11. nuclear power p
. to Prevent Packing Leakage' reactor facilities-holding an~0L or CP.
'l 87-39 Control ~of' Hot Particle 8/21/87.
All nuclear power Contamination at Nuclear reactor facilities-Power Plants and-spent fuel
-i storage facilities holding an NRC.
license or CP.
d
'87-38 Inadequate or Inadvertent 8/17/87 All nuclear. power
' Blocking of Valve Movement reactor facilities h"
.. Compliance with the General 8/10/87-All persons specifi-License Provisions of cally licensed to j
10 CFR Part 31 manufacture or to initia11y' transfer-devices containing radioactive material to general licensees, J
as defined in 10 CFR'
'f Part 31.
j 1
87-36' Significant Unexpected 8/4/87 All nuclear power j
Erosion of Feedwater Lines reactor facilities holding an OL or CP.
87-35 Reactor Trip Breaker, 7/30/87 All nuclear power Westinghouse Model DS-416, reactor facilities Failed to Open on Manual holding an OL or Initiation from the Control CP employing W 05-416 I
1 Room reactor trip breakers.
87-34 Single Failures in Auxiliary 7/24/87 All holders of an Feedwater Systems OL or a CP for pressurized water reactor facilities.
.OL = Operating License CP = Construction Permit' o
151 l
1 I
1 SSINS No.: 6835 IN 87-43 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON D.C. 20555 September 8, 1987 1
1 i
NRC INFORMATION NOTICE NO. 87-43: GAPS IN NEUTRON-ABSORBING MATERIAL l
IN HIGH-DENSITY SPENT FUEL STORAGE j
i RACKS I
I f
Addressees:
l All nuclear power. reactor facilities holding an operating license or a con-struction permit, j
Purpose:
This notice is to alert recipients to a potentially significant problem per-taining to gaps identified in the neutron absorber component of the high-density spent fuel storage racks at Quad Cities Unit 1.
The safety concern is that certain gaps might excessively reduce the margin of nuclear subcriticality
{
I in the fuel pool The NRC expects that recipients will review this notice for applicability to their facilities and consider actions, if appropriate, to preclude a similar problem occurring at their facilities. However, suggestions in this notice do not constitute NRC requirements; therefore, no specific action or written response is required.
l
)
Description of Circumstances l
On May 1, 1987, Commonwealth Edison Company (CECO), the licensee at Quad Cities
)
l'and 2, presented data to the NRC regarding gaps measured in Boraflex, a l
neutron-absorbing material used in the high-density fuel storage racks manufac-tured by the' Joseph Oat Corporation (OAT). Boraflex is a trade name for a boron carbide dispersion in an elastomeric silicone matrix manufactured by Bisco Products, Inc. (BISCO). Data pertaining to the gap size and distribution had been obtained by National Nuclear Corporation (NNC) under contract to CECO.
The licensee had retained Northeast Technology Corporation (NETCO) to interpret the data. NETCO prefaced their assessment as preliminary, noting that avail-
.able data was limited, but concluded that the gap formation mechanism may be related to large local stresses in the Boraflex from fabrication-induced re-straint within the rack and to tearing and shrinkage of the material.
The average gap size is 1-1/2 inches, with the largest 4 inches. The gaps accur in the upper two-thirds of the cell length.
8709010085 153
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IN 87-43 September 8, 1987 y
Page 2 of 3' l
These gaps are inferred from-anomalies in " blackness". testing results by NNC.
J
- The existence of a gap in the Quad Cities neutron absorber panel has been confirmed by underwater neutron radiography conducted by Nusurtec, Inc.
CEC 0'also discussed the effects'these gaps might have on the approved safety q
L analysis for the-spent fuel storage racks. ' CECO used conservative assumptions for gap size, gap location, and' fuel burnup. Considerable margin in k-eff I
.' appears to'be available before the licensing limit of 0.95 would be approached.
In July 1986, Wisconsin Electric-Company, the licensee at Point Beach I and 2,-
reported to the NRC that. test coupons of Boraflex material'nad shown consider.
able. degradation under high radiation. However, the licensee asserted that
-this' result did not represent the actual condition of Boraflex used in its spent fuel storage racks because of differences in methods of encapsulation, sample geometry, and' handling frequency. Additionally, the coupons had been
. subjected to about 5' times more radiation than is associated with the. average fuel rack position. Subsequent examination of full-length panels disclosed i
two results: in one pane) examined for effects of the water environment but exposed to negligible gamma radiation, there was no degradation of the Boraflex.
In another panel exposed to significant gamma radiation, 1-2 percent of the surface.3nowed a gray discoloration at the edges, similar to the degradation i
of the coupons, j
'l Discussion:
The concern is that separation of the neutron-absorbing material used in high density fuel storage racks might coinpromise safety. Although Quad Cities reports that its racks, even with 'g'aps in the Boraflex as large as 4 inches, can mcet the criticality criterion of k-eff less than or equal to 0.95, this may not'be the case for larger gaps or for other plants. A list of the 31 sites using Boraflex is given in Attachment 1.
Related information is given in "Behavict of High-Density Spent-Fuel Storage Racks," EPRI NP-4724. Electric
. Power Research Institute, August 1986.
Efforts'to understand the gap formation have revealed seversi topics on which information is needed. Accordingly, the material supplier (BISCO) and the Electric Power Research Institute (EPRI) have undertaken research programs tp collect this information. Some of their objectives are described below.
The BISCO program aims to establish with increased accuracy the relationship between radiation dose and size changes. The program also evaluates the potential effects of handling and restraint, during and subsequent to the fuel rack fabrication, on gap formation.
The EPRI program will correlate data from utilities' neutron absorber coupon surveillance programs. EPRI will further examine data obtained from CECO, as well as from BISCO and other sources, to improve the understanding of possible or actual gap formation models, including the effects of rack fabrication methods and irradiation damage mechanisms. The EPRI Program will also attempt 154
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[x.,
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" i, IN87-43l 4
September 8, 1987
>~,
Page 3.ofL3 J o model the specific Quad Cities experience considering absorbed gamma ' dose as a^
tfunction of axial elevation, neutron absortsing sheet restraint, and fractional change.in length.
The effect of' rack design and manufacturing methods on the consequences of stress, temperature, and chemical environment to irradiated Boraflex is.uncer-tain. Recent blackness test results.'at Turkey. Point,'who uses a Westinghouse-
' spent fuel storage rack, did not indicate the presence of gaps in the Boraflex.
The research programs are designed to evaluate each consequence and,;in partic-ular, to' improve the understanding of stress caused by method of attachment of the Boraflex panel to the stainless steel wall of the cell.
l Together, these programs are designed to improve the industry understanding of-
,the safety implications of the observed gaps'in the Boraflex neutron absorber L
component of the OAT high. ensity spent-fuel storage racks at Quad Cities.
d L
No specific action or written response is required by this information notice.
If. you have any questions about this matter, please contact the ' Regional
' Administrator of the appropriate regional office or.this office.
' fQ harles E. Rossi, Director' Division of Operational Events' Assessment.
~ "
Office of Nuclear Reactor Regulation
- Technical. Contacts
- Vern Hodge (301) 492-8196 Albert D. Morrangiello (309) 654-2227 I
Attachments:
1.
List of' Plants Using Boraflex
$$$^**
Structures in the Spent Fuel Pool s.pte.oer s, 1987
- 88' l 1 2.' List of Recently Issued NRC Information Notices LIST OF PLANT $ WITH 80RAFLEX STRUCTURES IN THE SPENT FUEL POOL 1.
Arkansas 1,2
- 16. Peach Bottom 2,3 j
2.
Beaver Valley 1
- 17. Pfigria 3.
Diablo Canyon 1,2 18.
Pt. Beach 1,2 4.
Calvert Cliffs 2 19 Pr. Island 1.2 5.
Farley 1,2
- 20. Quad cities 1,2*
6.
Fermi 2*
21 Rancho Seco".
7.
ft'Calhoun.
- 22. River 9end 8.
Ginna
- 23. Rooinson 2 9.
Grand Gulf 1,2*
24.
Summer"'
- 10. McGuire 1,2 25 Trojan
'11.
Mt11 stone 1,2,3
- 26. Turkey Pt. 3,4
- 12.'
Ntoe Mt. Pt. 1,2
- 27. Waterford 3
- 13. North Anna 1,2
- 29. Seabrook 1,2 j
+
14.
Oconee 1,2,3
- 29. watts Bar 1,2
- 15. Oyster Creek *
- 30. Comanche Peak 1,2 31 Herrfs 3
aPlants having spent fuel storage racks fabricated by Joseph Oat Corporation.
155 l
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i SSINS No.: 6835 j
IN 87-44 j
UNITED STATES NUCLEAR REGULATORY COMMISSION l
OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C.
20555 j
i September 16, 1987 I
4 NRC INFORMATION NOTICE NO 87-44: THIMBLE TUBE THINNING IN WESTINGHOUSE l
REACTORS Addressees:
+
All pressurized water reactor facilities employing a Westinghouse nuclear l
steam supply system (NSSS) holding an operating license or a construction l
permit.
l
Purpose:
\\
This information notice is being provided to alert addressees to potential It j
problems resulting from thimble tube thinning in Westinghouse reactors.
I is expected that recipients will review the information for applicability j
to their facilities and consider actions, if appropriate, to avoid similar problems, However, suggestions contained in this information notice do not l
constitute NRC requirements; therefore no specific action or written response is required.
Description of Circumstances:
During the recent refueling outage at North Anna Unit 1, eddy current (EC) testing identified wall thinning on approximately 23 out of 50 thimble tubes.
The wall degradation occurred on the thimble tubes just above the lower core Several plate, between the lower core plate and the fuel assembly guide tubes.
thimble tubes with greater than 35% wall thinning were identified, with one thimble tube thinned as much as 49%.
{
Discussion:
The movable incore neutron detectors travel within retractable thimble tubes.
l" The thimble tubes normally extend (as indicated in Attachment 1) from a 10-path transfer device, through the seal table, through the bottom of the reactor vessel, and into selected fuel assemblies. The thimble tubes are supported j
by guide tubes within the lower vessel region and the fuel assemblies, and by high pressure conduits between the reactor vessel and the seal table.
The thimble tubes are sealed at the leading (reactor) end, but are open at the 10 path transfer device to allow insertion of an incore, eutron detector.
f n
8709100056 f
157 l
- --_------- ----_ _ _ j
IN 87-44 September 16., 1987 Page 2 of 3 Hechanical high pressure seals, located at the seal table, are used to seal the area between the thimble tube and the high pressure conduit. This seal serves as a reacter coolant system (RCS) pressure bcundary since the area between the thimble tube and the high pressure conduit is at RCS pressure.
Consequently, a leak in a thimble tube results in degradation of the RCS pressure boundary by creating a path for reactor coolant to bypass the mechanical seal.
In order to halt the flow of leaking reactor coolant, the manual isolation valve must be closed.
As indicated, the thimble tubes are supported over most of their length.
However, a small portion of the thimble tube is directly exposed to RCS flew. This exposed por tion is between the top of the lower core plate and the bottom of the fuel assembly..This region is approximately 18.4 to 34.8 mm in length, depending on the reactor type.
'It is believed that flow-induced vibration on this exposed portion causes fretting at the adjacent guide tubes.
Undetected thinning of a thimble tube could lead to the development of a non-isolable leak and a corresponding loss of reactor coolant. As discussed I
previously, the manual isolation valve would have to be cloe.ed to halt the
{
flow of leaking reactor coolant. The leaking coolant may create an environ-
{
. ment in the vicinity of the isolation valves too hazardous for personnel to
{
- enter, j
Leaking thimble tubes could result in degradation of the incore neutron moni-toring system.
If not isolated, reactor coolant from leaking thimble tubes Cdn flow into the 10 path transfer device, allowing coolant to flood the other thimble tubes originating from that device. This could result in rendering l
inoperable more than just the leaking tube.
)
L
}
In addition to North Anna Unit 1, incore thimble tube thinning and leakage has been detected at facilities in France and Belgium. In this country, leaks in thimble tubes are known to have occurred at Salem Unit 1.
In Licensee Event Report (LER)81-028, Public Service Electric & Gas Co. (PSE&G) reported that three incore thimble tubes were known to have developed leaks
)
because of fretting. One of these leaks resulted in the flooding of all six
{
10 path transfer devices, partially or completely flooding all the thimble i
tubes in the reactor.
In addition, thinning has been det.ected on the Farley l
l thimble tubes.
l At North Anna Unit 1, the proposed corrective action was to retract selected l
thimble tubes approximately 2 inches. This would move the thinned area out l
of the region of high turbulence. In addition, the thimble tube that experi-enced the most degradation will be taken out of service by closing the corre-I sponding isolation valve.
l l
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158 I
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IN'87-44 September 16,.1987 Page 3 of 3 q
'i Nonspecific-action or written: response'is required'by'this information notice.
k If you'have any questions about this' matter, please contact the Regional i
Administrator of the appropriate regional office or_this office.
f
.c J
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g Charles E. Rossi,' Director Divisior, of Operational Events Assessment.
1 Office of Nuclear Reactor Regulation-
~
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Technical Contr.ct: Jack Ramsey, NRR.
a,,
(301) 492-9081<
' Attachments:
l 1.
Typical Westinghouse Incore Neutron' Monitoring System'
- 2.. List'of Recently. Issued NRC.Information Notices 4
. j
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1 159
\\
a TYPICAL WESTINGHOUSE INCORE NEUTRON MONITORING SYSTEM SAFETY SWITCHES DRIVE UNITS LIMIT SWITCHES 8-PATH TRANSFERS INCORE NEUTRON 10-PATH TRANSFERS -
INCORE NEUTRON s DETECTOR INSIDE THIMBLE TUBE ISOLATION I
SEAL TABLE b
I MECHANICAL
^
II HIGH PRESSURE CONDUlTS I
I f
i REACTON e
CORE THIMBLE TU9E INSIDE FUEL ASSEMBLY OulDE TUBE UNQU10ED THIMBLE TU8E MAY EXPERIENCE THINNING.
IN THl8 REGION N
l THIMSLE TUBE INSIDE A
LOWER VESSEL RE0 LOP!
LOWE R I
OulOETUBE CORE I
PLATE HIGH PRESSURE b
CONDUlTS e D 2
n THlMBLE TUBES I
INSIDE HIGH PRESSURE CONDUlTS s
160
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/
h' Attachment'2 IN 87-44
-~ Septembe r 16, 1987 ci-
. LIST OF.RECENTLY ISSUED-
,,, ~
INFORMATION NOTICES 1987 l,
q
.Information.
Date of ENotice No; Subject'
' Issuance Issued to J
87 43-
^
'-Gaps'in Neutron-Absorbing
~ 9/8/87 Allinuclear' power
' Material'in High-Density reactor' facilities Spent Fuel Storage' Racks holding an OL or CP.
87-42.
' Diesel. Generator Fuse 9/4/87 All nuclear power-reactor facilit'ies Contacts.
87-41 Failures of Certain Brown 8/31/87 All nuclear powerJ Boveri-Electric Circuit reactor facilities.
L<4 Breakers holding an OL or.CP.
) 40I Backseating Valves Routinely. 8/31/87 All nuclear power E
to' Prevent-Packing Leakage.
reactor facilities holding'an OL or CP.
L 87-39' Control of Hot Particle' 8/21/87 All nuclear' power Contamination at Nuclear reactor facilities L
and spent fuel-l.
Power Plants storage facilities holding'an NRC
-license or CP.
87-38 Inadequate or Inadvertent 8/17/87 All nuclear. power Blocking of. Valve Movement reactor-facilities-.
- 1 87-37 Compliance with the General 8/10/87' All persons speciff--
1 License Provisions of cally licensed to'
)
10 CFR Part 31 manufacture or to
'l initially transfer.
I devices containing 1
radioactive material to general licensees, as defined in 10 CFR
.,.H Part 31.
87 Significant Unexpected 8/4/87 All nuclear power Erosion of Feedwater Lines reactor facilities 1
j j
OL y Operating License CP = Construction Permit l
161 1
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INDEX a
LFACILITY:
REPORT NUMBER
'PAGE
- Allied Nut &' Bolt Company, Inc.-
King;of Prussia, Pennsylv6nia 99901093/87-01 1-Colt Industries Beloit, Wisconsin.
99900300/86-01
.9-
- Cooper Energy Services.
Grove City,' Pennsylvania' 99900317/87-01 17-Frank Electric Corporation
. York Pennsylvania 1 99901091/87-01 29 General Electric Company Philadelphia, Pennsylvania' 99900219/87-01
.39' Limitorque Corporation q
Lynchburg, Virginia 99900100/87-01 49 c.;
Namco Controls,xInc.
Mentor Ohio 99900378/87 ' 57!
[-
L Patel. Enterprises. Inc.
Huntsville,-Alabons 99900931/87-01 65 Posi-Sea' l, ' International North Stonin9 ton, Connecticut-99900886/87-01 75~
Powerplant Specialists, Inc' -
Costa Mesa,' California
.99901086/87-01 85 Precision Heat Treating Company -
- Montebello, California
.99901096/87-01 89-I' Satin American Corporation Shelton :-Connecticut 99901094/87-01 97
-Southwestern Engineering Company
-)
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163
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I-APPLIES TO ALL PLANTS DOCKETNO.- APPLIES ONLY-T0 THE IDENTIFIED UNIT 166
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BIBLIOGRAPHIC DATA SHEET NUREG--0040 Vol. 11, No. 3 ses taverio.
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July 1987 - Se enber 1987
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