ML20237L115
| ML20237L115 | |
| Person / Time | |
|---|---|
| Issue date: | 08/31/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0040, NUREG-0040-V11-N02, NUREG-40, NUREG-40-V11-N2, NUDOCS 8708270427 | |
| Download: ML20237L115 (150) | |
Text
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,an. tg NUREG-0040 Vol.11, No. 2 LICENSEE CONTRACTOR AND VENDOR INSPECTION l STATUS REPORT QUARTERLY REPORT l APRIL 1987 - JUNE 1987 l l
UNITED STATES NUCLEAR REGULATORY COMMISSION f.~<m,,,
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Available from )
Superintendent of Documents U.S. Government Printing Office Post Office Box 37082 l Washington, D.C. 20013-7082 A year's subscription consists of 4 issues for this publication.
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Single copies of this publication I are available from National Technical ,
! Information Service, Springfield, VA 22161 l
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NUREG-0040 Vol.11, No. 2 LICENSEE CONTRACTOR AND VEND 0R INSPECTION STATUS REPORT QUARTERLY REPORT ..
APRll1987 JUNE 1987 s
$ te fuNfded72u*g'uIt 1d517 Division of Reactor inspection and Safeguards
- Office of Nuclear Reactor Regulation .-
U.S. Nuclear Regulatory Commission Washington, DC 20555
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CONTENTS l
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- 1. Preface ..................................................... 111 l
- 2. Reporting Format ........................................... V
- 3. Inspectors Reports .......................................... 1
- 4. Selected Information Notices ................................ 129
- 5. Index ....................................................... 145
- 6. Table of Vendor Inspection Reports Related to Reactor Plants ................................. 147 l 1
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i PREFACE A fundamental premise of the Nuclear Regulatory Commission's (NRC) nuclear facility licensing and inspection program is that licensees are responsible for the proper construction and safe operation of their nuclear power plants.
The total government-industry system for the inspection of nuclear facilities has been 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 ruies and regulations. The NRC inspects to determine whether its requirements are being met by a licensee and his contractors, while the great bulk of the inspection activity is performed by the industry within the framework of ongoing 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 l 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 nucicar 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, firms designing nuclear steam supply systems, architect engineering firms designing nuclear power plants, and certain selected major equipment vendors were inspected on a regular basis by NRC to ascertain through direct observation of selected activitics 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
vendor interfaces, environmental qualification of equipment, and equipment problems found during operation and corrective action. If nonconformances with NRC requirements and regulations are found, the inspected organization is requested to take appropriate corrective action and to institute preventive measures to preclude recurrence.. If generic implicai'ons are identified, NRC assures that affected licensees are expeditiously informed.
8 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 ir.dicate that NRC inspections have confirmed satisfactory implementation of the accepted QA programs. Licensees and applicants could, at their option, use the letters to fulfill their obligation under 10 CFR 50 Appendix B, Criterion VII, that requires them to perform initial source evaluation audits and subsequent periodic audits to verify QA program implementation. However, based on the above described change in nuclear plant design and construction activities, NRC will no longer issue confirming letters to principal contractors since future NRC vendor program inspections will focus on selected areas rather than addressing the implementation of their respective QA programs. Therefore, confirming letters that have already exceeded their three year effective period will not be renewed. Confirmir.g letters issued less than three years ago will remain in effect until the stated effective period expires. Therefore, as the confirming letters expire, licensees and applicants will no longer be allowed to take credit for the NRC acceptance of the implementation of a principal contractor's QA program. Licensces 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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l ORGANIZATION: COMPANY, DIVISION CITY, STATE REPORT INSPECTION INSPECTION N0.: Docket / Year / Sequence DATE: ON-SITE HOURS:
CORRESPONDENCE ADDRESS: Corporate Name Division ATTN: Name/ Title Address City, State Zip Code ORGANIZATIONAL CONTACT: Name/ Title TELEPHONE NUMBER: Telephone Number NilCLEAR INDUSTRY ACTIVITY: Description of type of components, equipment, or services supplied.
ASSIGNED INSPECTOR:
Name/ Vendor Program Branch Section Date OTHER INSPECTOR (S): Name/ Vendor Program Branch Section APPROVED BY:
Name/ Chief - Section/ Vendor Program Branch Date INSPECTION BASES AND SCOPE:
A. BASES: Pertain to the inspection criteria that are applicable to the activity being inspected; i.e., 10 CFR Part 21, Appendix B to 10 CFR Part 50 and Safety Analysi. Report or Topical Report commitments.
B. SCOPE: Summarizes the specific areas that were reviewed, and/or identi-fies plant systems, equipment or specific components that were inspected.
For reactive (identified problem) inspections, the scope summarizes the problem that caused the inspection to be performed.
PLANT SITE APPLICABILITY: List plant name and docket numbers of licensed facilities for which equipment, services, or records were examined during the inspection.
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ORGANIZATION: ORGANIZATION CITY, STATE REPORT INSPECTION NO.: 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 nonconformance with applicable commitments to NRC requirements.
In addition to identifying the applicable NRC requirements, the specific industry codes and standards, company QA manual sections, or operating procedures which are used to implement these commitments may be referenced.
C. UNRESOLVED ITEMS: Shown here are inspection results about which more information is required in order to determine whether they are acceptable items or whether a violation or nonconformance may exist. Such items will be resolved during subsequent inspections.
D. STATUS OF PREVIOUS INSPECTION FINDINGS: This section is used to identify the status of previously identified violations, items of nonconformance, and/or unresolved items until they are closed by appropriate action.
For all such items, and if closed, include a brief statement concerning action which closed the item. If this section is omitted, all previous inspection findings have been closed.
E. INSPECTION FINDINGS AND OTH.? COMMENTS: This section is used to provide significant information concerning the inspection areas identified under
" Inspection Scope." Included are such items as mitigating circumstances concerning a violation or nonconformance, or statements concerning the limitations or depth of inspection (sampic size, type of review performed and special circumstances or concerns identified for possible followup).
For reactive inspections, this section will be used to summarize the disposition or status of the condition of event which caused the inspection to be performed.
F. PERSONS CONTACTED: Typed, Name, Title
- present during exit meeting SAMPLE PAGE (EXPLANATION OF FORMAT AND TERMIN0 LOGY) vi
INSPECTORS REPORTS l
ORGANIZATION: THE AMERACE CORPORATf0N i UNION, NEW JERSEY INSPECTION INSPECTION {
REPORT DATES: 8/25-29/86 ON-SITE HOURS: 85 )
NO : 99900296/86-01 i CORRESPONDENCE ADDRESS: The Amerace Corporation Control Products Division ATTN: Mr. H. Schulte, Vice-President and General Manager j i
1065 Florei Avenue !
Union, New Jersey 07083 ORGANIZATIONAL CONTACT: Mr. Joseph Ferguson, QA Manager TELEPHONE NUMBER: (201) 289-8200 i
NUCLEAR INDUSTRY ACTIVITY: The Amerace Corporation manufactures and distributes Agastat E-7000 series timing, magnetic latching and general purpose relays, and Buchanan terminal blocks for the nuclear industry.
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A ASSIGNED INSPECTOR: ,
/( fg e !
- 4. J/ ptrosi6o, Reactive Inspection Section, (RIS)
OTHER INSPECTOR (S): E. Yachimiak, RIS D. King Progra ao nation Section ;
APPROVED BY: / /o/uk(. {
E. W. Verschoff, Ch RIS, Vendor Program Branch Date j INSPECTION BASES AND SCOPE:
A. BASES: Appendix B to 10 CFR Part 50 and 10 CFR Part 21.
B. SCOPE: 1) Verify QA program implementation as a followup of a recent l U.S. Nuclear Regulatory Commission (USNRC) safety system modification l l inspection at the Dresden nuclear power plant. 2) Followup of a USNRC j inspection at Control Products Corporation that identified a potential ;
10 CFR Part 21 violation at the Amerace Corporation. !
PLANT SITE APPLICABILITY: All nuclear power stations that use Agastat electrical !
control relay series E-7000, EGP, ETR or EML. l i
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ORGANIZATION: THE AMERACE CORPORATION UNION, NEW JERSEV REPORT INSPECTION NO.: 99900296/86-01 RESULTS:
PAGE 2 of 14 A. VIOLATIONS:
Contrary to Section 21.31, " Procurement Documents," of 10 CFR Part 21, the Amerace Corporation Control Products Division (CPD) failed to impose the provisions of 10 CFR Part 21 on the Control Products Corporation of Grafton Wisconsin during the period of June 3,1985 until Jurie 29, 1986. The ,
Control Products Corporation manufactures nuclear grade Agastat relay series EGP, ETR, and EML for the Amerace Corporation (86-01-01).
B. NONCONFORMANCES:
1.
Contrary to Criterion I, " Organization," of Appendix. B to 10 CFR Part 50, the CPD ouality assurance (QA) manager and the OA/QC organization lack sufficient authority and organizational freedom to identify quality problems, and are not independent from production cost and schedule considerations. The QA manager reports to the production faci.lity manager, which does not assure that effective execution of the QA program will be performed as required (86-01-02).
2.
Contrary to Criterion 111. " Design Control," of Appendix B to 10 CFR Part 50, CPD failed to adequately translate several design drawing requirements for its Agastat E-7000 series relays into inspection characteristics to assure that the design parameters were within tolerances. Specifically (86-01-03):
a.
CPD failed to translate screw design torque parameters for the Agastat relay assembly into inspection verification attributes to assure that the minimum torque values are reached and that the maximum torque values are not exceeded.
b.
CPD failed to translate several required dimensional charac-teristics from CPD engineering change notice number 083924 onto CPD quality control (QC) inspection verification documents to assure that the safety-related components complied with the design requirements.
3.
Contrary to Criterion V, " Instructions, Procedures, and Drawings," of Appendix B to 10 CFR Part 50, and Section 1.3, of the CPD QA Manual, CPD failed to establish quality control inspection procedures or instructions for receipt inspection and in-process inspection activities for the Agastat E-7000 series relays (86-01-04).
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ORGANIZATION: THE AMERACE CORPORATION UNION, NEW JERSEY REPORT INSPECTION RESULTS: PAGE 3 of 14 NO.: 99900296/86-01
- 4. Contrary to Criterion VI, " Document Control," of Appendix B to 10 CFP Part 50, and Section 13, of the CPD QA Manual, as of August 29, 1986 (86-01-05)
- a. CPD failed to assure that the QC inspection department super-visor prepared the material receipt inspection checklists. ,
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- b. CPD QA/QC management failed to review and approve QC material i receipt inspection checklists that were prepared by the receipt j inspection QC inspectors who performed the inspections. 1
- 5. Contrary to Criterion VII, " Control of Purchased Material, Equipment, f and Services," of Appendix B to 10 CFR Part 50, CPD procured the calibration services of a vendor who was not on its approved vendors .
list. The unapproved vendor was the Sheffield Measurement Division of the Warner and Swassey Company (SMD). SMD calibrated the CPD's receipt inspection departments Cordax 1000 dimensional measurement machine (86-01-06). y 1
- 6. Contrary to Criterion X, " Inspection," of Appendix B to 10 CFR Part 50, and Sections 5 and 6 " Receiving Inspection" and "In-Process Inspection," of the CPD QA Manual, CPD failed to adequately establish and execute certain quality . inspection activities. Specifically -
(86-01-07): ,
- a. Two examples were noted where QC receipt inspection personnel verified only a portion of the stated QC inspection lot popu-lation for certain characteristics,
- b. Written QC instructions were not established for the QC receipt inspection and in-process Agastat relay QC inspection personnel as required. ;
- c. CPD route sheets (" Travelers") were not established and trans-mitted to the QC in-process' inspection personnel for the Agastat E-7000 series work activities. The CPD route sheets are
" travelers" to record the in-process QC inspectors signature and date of work activity verifications.
- d. Final QC inspection practices did not assure that the quality records were completed. As an example, the required route sheets were not completed but the nuclear orders were still signed off as being in compliance, i
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c ORGANIZAT10N: THE AMERACE CORPORATION UNION, NEW JERSEY REPORT INSPECTION NO.: 99900296/86-01 RESULTS:
PAGE 4 of 14 7.
Contrary to Criterion XII, " Control of Measuring and Test Equipment (M&TE)," of Appendix B to 10 CFR Part 50, CPD failed to adequately control its measuring and test equipment as evidenced by the NRC inspector finding 17 out of 32 M&TE items that were either not con-
) trolled or were out of calibration (86-01-08).
- 8. Contrary to Criterion XV, " Nonconforming Materials, Parts, or Components," of Appendix B to 10 CFR Part 50 and Section 9. " Control of Nonconforming Materials," of.the CPD QA Manual CPD failed to control 7,800 out of 68,500 defective Agastat E-7000 contact arm assemblies (Part Number 32356-01). Three CPD defective material reports (DMR) were written to control a total of approximately 60,660 contact arm assemblies, but approximately 7,800 of these were not controlled to prevent their inadvertent use. Additionally, the three DMR's were not written until 9/20/85, and the nonconformance was first noted on 1/20/85(86-01-09).
- 9. Contrary to Criterion XVIII, " Audits," of Appendix B to 10 CFR Part 50, CPD has failed to establish a planned system of periodic audits to verify its QA program implementation and effectiveness (86-01-10).
C. UNRESOLVED ITEMS:
None.
D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
This was the first NRC inspection that was performed at this Agastat 7000 series relay facility.
E. OTHER FINDINGS OR COMMENTS:
- 1. 10 CFR Part 21.
The CPD 10 CFR Part 21 procedure was reviewed for its adequacy to provide compliance with 10 CFR Part 21 reporting requirements. In addition, the procedural implementation of the 10 CFR Part 21.6 posting requirements were evlauated by inspecting the manufacturing areas. This review found that CPD has an adequate 10 CFR Part 21 procedure and has posted the required documents.
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ORGAN 12AT10N: THE AMERACE CORPORATION )
UNION, NEB JERSEY REPORT INSPECTION ,
NO.: 99900296/86-01 RESULTS:. .PAGE 5 of 14 1 During a recent NRC inspection at the Control Products Corporation (CPC) in Graf ton, Wisconsin, it was revealed that the Amerace Corporation had not imposed the provisions of 10 CFR Part 21 on CPC for nuclear grade component manufacturing. CPC of Grafton, Wisconsin is the manufacturer for three series of nuclear grade Agastat relays for Amerace. A review of CPD purchase orders determined that CPD had not imposed the requirements of 10 CFR Part 21 on CPC during a one year period, between June of 1985 and June of 1986. (see Violation 86-01-01)
- 2. Quality Assurance Organizational Establishment.
Discussions were conducted and a review was performed of the CPD QA Manual Organization Section to verify that adequate QA department
" access to management" had been established, delineated, and executed as required for resolving quality problems.
The review revealed that the QA organization did not have sufficient authority and organizational freedom from the production facility management to identify and resolve quality problems. In addition, discussions with the QA/0C personnel determined that they were not adequately independent of production cost and schedule concerns (see Nonconformance 86-01-02).
- 3. Design Specifications.
NRC inspectors determined that CPD failed to translate several design specifications into QC inspection verification instructions. .The NRC inspectors querried CPD personnel concerning its fastener instal-lation methods that were observed in the assembly areas regarding Agastat relays. It was noted that each assembly work station where fasteners were used had at least one pneumatic (air) screwdriver that was used for screw fastener installation. There were no obvious torque ranges specified on the air screwdrivers by the tool manu-i facturer or by CPD. None of the air screwdrivers were controlled by l CPD under their M&TE system. Therefore, CPD was cuestioned as to how they verify or monitor that its typical inch pound screw fastener torque parameters are-not exceeded. The NRC inspector noted that the air screwdrivers being utilized by CPD at the work stations were capable of exceeding the CPD torque values.
It was also noted that the assembly personnel would use a calibrated l
torque screwdriver to verify that the fastners had obtained specific minimal torque, but no measures were performed to verify if the design torques were exceeded. Design specification PS41, Revision H, l
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i ORGANIZATION: THE ANERACE CORPORATION UNION, NEW JERSEY REPORT INSPECTION NO.: 99900296/86-01 RESULTS: PAGE 6 of 14 dated October, 1984 states that blade retainer screws (design drawing !
- 32650) shall be within a torque range of 5-9 inch-pounds. However, j this design torque parameter had not been translated into a quality '
verification document. The CPD QC inspectors were checking for minimum torque by using a break-away torque wrench set at 6 inch-pounds. The NRC inspector determined that CPD had no inspection instructions or verification records to assure that the 8 retainer screws had received the required minimum torque, or to assure that they had not been overtorqued. A further review of this crea deter-mined that CPD has not translated any of the other Agastat relay assembly screw design torque values into QA/CC inspection verification attributes.
Another aspect within the same design requirement area was reviewed and a second example of inadequate translation of design requirements into inspection verification attributes was revealed. Specifically, CPD failed to translate the dimensional measurement criteria from a CPD engineering change notice number @3p for an electrical contact strap, (Part Number MQQ6,4), into 0A/QC inspection verifi-cation documents (see Nonconforman,ce 86-01-03).
4 Instructions, Procedures and Drawings.
Discussions were conducted with CPD QA/QC personnel to determine if their QA/QC instructions contained adequate qualitative and quanti-tative acceptance criteria for determining that important activities
, had been accomplished. This was done in the receipt inspection and in-process Agastat inspection areas. Deviations were found in both areas (see Nonconformance 86-01-04). Specifically:
- a. Material receipt inspection QC personnel were found to be using inspection checklists for all incoming components.
The QA/QC inspection checklists delineated each specific characteristic that was to be inspected. The QC inspectors i indicated that items had be~en inspected by putting a check- l mark and their unique inspection stamp adjacent to the check-mark. The NRC inspectors queried the CPD CA/0C inspection personnel regarding the type of instructions that were used, what attributes were required to be on the checklists, and why there were no supervisor approvals or review signatures on the inspection checklists. At this point, it was revealed that: 1) there were no required inspection checklist approvals being performed; 2) only the inspector that made up the checklist and inspected the l
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ORGANIZATION: THE AMERACE CORPORATION UNION, NEW JERSEY REPORT INSPECTION RESULTS: PAGE 7 of 14 NO.: 99900296/86-01 material reviewed the inspection characteristics for l i
adequacy; 3) there were no QC receipt inspection. procedures or instructions that delineated what the inspector was l required to do, and to ensure that the appropriate quali- {
tative and quantitative acceptance and rejection criteria f were included on the checklists; and 4) the inspection )
supervisor did not prepare the checklists as required by the QAM.
b) In-process inspection activities were observed in the E-7000 series Agastat relay manufacturing area. The NRC inspector l queried the CPD QC inspector as to what documents instructed !
her to perform the in-process inspection attributes that were being performed and what type of document she used to document ;
her inspections.
It was revealed that CPD had also failed to establish proce-dures or instructions in this area. The CPD QA Manual requires specific forms to be used for the QC instructions and the component " travelers." These deviations are discussed additionally in Section E.7.
i 5. Document Controls.
Discussions were conducted with the QC inspectors and a review was performed to determine if the CPD receipt inspection documents were reviewed for adequacy and approved by authorized personnel. It was found that CPD failed to comply with its QAM requirements in the receipt inspection area regarding the preparation and review of inspection checklists (see Nonconformance 86-01-05). j Six out of six OC receipt inspection checklists that were reviewed !
showed that the QC receipt inspector had established what inspection characteristics were to be inspected without any review for adequacy f of the documents by someone other than the person who prepared it.
Specific component checklists that were reviewed and found to be ;
without review and approval are as follows: l
- a. Block Terminal Assembly, CPD drawing 32279-00, rev. N.
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- b. Coil Assembly, Magnet AC, CPD drawing 32274-00, rev. E.
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- c. Core, Magnet AC, CPD drawing 32526-01, Rev. K.
- d. Arm Assembly, contact, CPD drawing 37356-00, Rev. B.
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ORGANIZATION: THE AMERACE CORPORATION' l UNION, NEW JERSEY j 1
REPORT INSPECTION NO.: 99900296/86-01 RESUt.TS : . PAGE 8 of 14
- e. Diaphram Assembly, CPD drawing 32372-00, Rev J.
- f. Binding head screw, CPD drawing 7002730-00, Rev J.
The CPD practice of allowing the same individual to write the check-list, review the checklist for adequacy, and inspect the hardware with the checklist is not consistent with Appendix B to 10 CFR Part 50.
Additionally, as discussed below, CPD takes exception to outside vendors source inspections and surveys based on a' satisfactory "first piece" inspection and history of acceptable shipments.
However, based.on the above, the CPD QA/QC management has not adequately controlled the implementation of its QAM requirements to assure themselves that they have been receiving " acceptable shipments."
- 6. Control of Purchased Material, Equipment and Services.
A review of the CPD area of purchased material, equipment, and services determined that CPD does not normally perform vendor source inspections or pre-award surveys at the vendor's facility. Instead, the CPD practice is to accept or reject the vendor on the basis of the outcome of the components "first piece" inspection.
During the review of Measuring and Test Equipment, a list of the calibration laboratory service vendors was reviewed to verify that all of the vendors were on the approved vendors list. It was deter-mined that one calibration service vendor was not on the approved vendors list. The vendor is Sheffield Measurement Division of the Warner Swassey Company (SMD). This vendor provided calibration services for CPD's Cordax 1000 measuring machine. The Cordax 1000 is used by CPD as a quality control tool for receipt inspection by measuring the dimensions of purchased ccanponents. SMD calibrated this machine initially on April 9, 1985 and subsequently recalibrates it on December 3,1985 and on' May 17, 1986. At the time of the inspection, the NRC inspector determined that SMD had not been surveyed and was not on CPD's approved vendor's list (see Nonconformance 86-01-06).
The NRC inspectors could not determine if the calibration services that were provided were adequate, and if the Cordax 1000 machine was within acceptable tolerances. -Since no records existed at CPD regarding vendor surveys, it is not known whether or not SMD has calibration standards that are traceable back to the National Bureau of Standards, I
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i ORGANIZATION: THE AMERACE CORPORATION UNION, NEW JERSEY REPORT INSPECTIOW NO.: 99900296/86-01 RESULTS: - PAGE 9 of 14 i
- 7. Receipt and In-process Inspections.
The receipt and Agastat relay in-process inspection areas were selected for inspections and review to determine if the Appendix B QA program was adequately " executed" and to verify CPD's conformance with the documented instructions, procedures, and drawings and if appropriate inspections were being performed. It was revealed that 1 CPD has not adequately executed their quality activity verification 4 inspection program (see Nonconformance 86-01-07). Specifically:
- a. Two examples of CPD allowing QC personnel to perform material receipt inspections without instructions or procedures, to approve the inspection checklist characteristics that they established, and to inspect smaller samples or fewer attributes than the QC personnel documented as having been inspected.
Specifically:
(1) A receipt inspection checklist for a magnetic coil assembly l (Part #32274-00), was checked as having been tested for dielectric test per TP-TRE-01. No records were available to indicate that this test had been performed at receipt.
inspection for lots received between 5/29/85 and 8/1/86, but the inspection checklist attributes had been checked and stamped by QC. Discussions determined that the dielectric tests were not performed by 00, but they were )
performed by manufacturing personnel after the fact.
I (2) A receipt inspection checklist for binding head screws j (Part #700?730 00), was checked as having the zine plating )
thickness verified for all samples in the specified lots received between 6/20-771786. The inspection supervisor and QC personnel stated that this was not done and that only a small sample of the inspectT6n T6Is were actually tested for plating thickness. This inspection checklist also indicated that OC verification was performed and had been checked and stamped by QC.
- b. Section 13 of the CPD QA manual requires that detailed written instructions be supplied to all manufacturing areas. outlining the steps to be taken to assure conformance to the requirements.
These detailed written instructions were not provided to the t in-process OC personnel for the Agastat E-7000 series relay area.
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ORGANIZATION: THE AMERACE CORPORATION UNION, NEW JERSEY
.c REPORT INSPECTION NO.: 99900296/86-01 RESULTS: PAGE 10 of 14 In addition, the CPD QA manu61 also requires on all nuclear orders that a- traveler accompany each E-7000 series relay order. The purpose of the traveler is to indicate by QC inspector stamp and date that all the important work activities have been performed. However, no travelers have been established or utilized to control the Agastat E-7000 series relay quality activities. .Several in-process nuclear orders were observed to establish.if travelers are used as required.
To date no travelers have been used as determined by discussions and.
observations'.
- 8. Measuring and Test Equipment (M&TE) Control.
A review was performed of the M&TE area af ter observations deter-mined that at least six M&TE items were not being controlled regard-ing unique identification and calibration. Further review determined that a total of 17 of 32 M&TE items were not adequately controlled as required by the CAM (see Nonconformance 86-01-08).
In addition, e, Tha111um-204 radioactive source was found by the NRC inspector inside of a storage cabinet located in the QC receipt inspection office. The QA/QC management stated that the source was used for a beta back scanner that has since been returned to the manufacturer. The USNRC Region I Nuclear Material Section was .
informed and will follow-up this matter.
Observations of CPD's M&TE controls at various work stations were performed. The NRC inspector determined that CPD failed to properly assure that M&TE are properly identified, controlled, and calibrated at the required intervals. CPD also failed to assure that M&TE computer records correspond to the actual M&TE calibration status of the as found items.
NRC inspectors discovered a .197" core-gap gage in an operating work station that was past due for calibration and was unidentified. This gage had no visible permanent ID number etched on it,-no ID # on its calibration sticker, and the sticker indicated that the gage had not been calibrated since August 31, 1985. CPD computer calibration records indicated that three of the .197" gages existed and were all properly calibrated. After looking at all of CPD's core-gap gages; excluding the uncontrolled .197" gage found by the NRC inspector, only two of the .197" gage size could be located. The following day the NRC inspector found that three .197" gages were identified and '
properly calibrated. The NRC inspector determined that the uncali-brated and unmarked gage found by the NRC inspector the day before 10
ORGANf2AT10N: THE AMERACE CORPORATION j UNION, NEW JERSEY I
I REPORT. INSPECTION NO.: 99900296/86-01 RESULTSr PAGE 11 of 14 )
l was the third of the .197" gages even though the M&TE computer.
record indicated it was in control. After CPD made that determin- ;
ation, the item was-apparently recalibrates and returned to its i proper work station. .q Specific examples of uncontrolled M&TE items ars as follows:
Item Name CPD-ID Problem
- a. Anvil micrometer SN-41 Red nonconformance tag not used, found out of calibration.
- b. .001" Caliper NO. 505-629 Red nonconformance tag not i used, out of calibration, j
- c. .197" core gap gage None No permanent ID # and I calibration records contradicted actus1 missing.
and uncalibrated status.
- d. .165"/.187" core gap DP-0317 No permanent ID f.
gage
- e. .183"/.190" core gap DP-0320 No permanent ID f.
gage i
- f. .187"/.165" core gap DP-0316 No pennanent ID #.
gage 9 .112"/.124" core gap DP-0311 No permanent ID # and gage has same sticker # as j item h. ;
- h. Core gap gage DP-0311 Samesticker#(DP-0311) as. item g and has same permanent # (DP-0689) as item 1.
i
- 1. Core gap gage DP-0312 Same permanent ID #.(DP-0889) as item h sticker ID # is DP0889.
- j. .192" cone gap gage DP-0323 Sticker ID # (DP-0323) doesn't match permanent ID # (DP-0324).
11 l
l . , . ..
ORGANIZATION: THE AMERACE CORPORATION UNION, NEW JERSEY REPORT INSPECTION NO.: 99900296/86-01 RESULTS: PAGE 12 of 14 Item Name CPD-ID Problem
- k. .192" cone gap gage DP-0324 Sticker ID # (DP-0324) doesn't match permanent ID f I
(DP-032.).
- 1. Micrometer DP-1029 Found in work station, out of calibration no red tag.
- m. Micrometer None Found with no ID # and not properly calibrated no red tag.
- n. Defective micrometer S/N 37 Red nonconformance tag not used and item was not found in box,
- o. .583" Plug gage S/N 61639 Obsolete. Gage found in same bin as calibrated thread plug gages.
- p. Torque screwdriver DP-1093 Calibration records contradicted the uncalibrated status indicated on the sticker.
- q. Torque screwdriver DP-1096 No permanent ID #
- 9. Control of Nonconforming Parts.
A review of the circumstances surrounding a CPD documented problem with an Agastat E-7000 relay sub-component was performed to assess the adequacy of the CPD control of nonconforming parts. The review determined that CPD had not adecuately controlled an entire lot of <
nonconforming components (see Nonconformance 86-01-09).
The above component is a contact arm assembly (CAA), Part Number 32356-01, and was received from one of the two vendors who manufac-tured the component for CPD. It was determined by the NRC inspector and CPD that all but 1,500 of the CAA's could be accounted for by a review of records. CPD stated that the 1,500 assemblies were not used in production, but were retained for samples, etc. The failure of CPD to control the parts was established by the OA/QC DMR, and lack of QA records to indicate the final disposition of the CAA components.
12
ORGANIZATION: THE AMERACE CORPORATION :
UNION, NEW JERSEY REPORT INSPECTION NO.: 99900296/86-01 RESULTS:' PAGE 13 of 14 The first sample parts were rejected on 1/22/85 and the second sample )
was conditionally accepted on 1/24/85 even.though the parts deviated j from the dimensional requirements. The first production parts were accepted'on 6/4/85 even though the previous two samples failed the. ,.
dimensional requirements. CPD inspection records did not have all 4 of the required inspection dimensional attributes. A total of four production lots were received between 6/4/85 and 9/20/65 for an approximate total of 68,500. However. J l
- a. The first DMR written was on 9/19/85 for 30,400 pieces, I the second DMR was written on 9/20/85 for 20,600 pieces, and the third and last DMR also was written on 9/20/85 '
for 9,600 pieces. This makes a total of 60,600 pieces that QC formally controlled after 9/19/85.
- b. Approximately 7,800 pieces were not controlled as required.
In addition, the first DMR was not written until the last shipment was received on 9/19/85, even though defective '
CAA components were received as early as 1/24/85.
- c. CPD documents state that "the nonforming parts were discovered in the E-7000 series production line and that the terminal block where the defective CAA components were used "did not weigh properly due to damaged and bowed contact arm assemblies."
- d. CPD's corrective action (not a QA DMR documented item) was i to remove all known parts from production and replace them i with the original vendor's CAA components, return all the-deficient CAA's to the other vendor, and not use that vendor again.
- 10. Internal Audits.
A review of the CPD internal audit program was performed. One CPD audit report was in evidence, but no audit plan or schedule had been established (seeNonconformance 86-01-10).
13 u_______
ORGANIZATION: THE AMERACE CORPORATION UNION, NEW JERSEY REPORT INSPECTION NO.: 99900296/86-01 RESULTS: PAGE 14 of 14 F. PERSONS CONTACTED Name Title Organization
- Joseph Ferguson Quality Assurance Mgr Amerace Corporation i
- Kenneth Beckett OC Supervisor Amerace Corporation 1 Don Alexander Customer Service Mgr Amerace Corporation
- Ed Leszczak Product Engr Amerace Corporation
- Albert F. Schulfen Plant Manager Amerace Corporation
- William Waddington Chief Engineer Amerace Corporation E. Hill QC Inspector Amerace Corporation l A. Perez QC Inspector Amerace Corporation ;
- Attended exit meeting i
j l
. I I
l 14
)
CRGANIZATICK: BRAND INDUSTRIAL SERVICES, INCORPORATED PARK RIDGE, ILLIN0IS REPORT INSPECTION INSPECTION hc.: 99901020/87-01 DATES: 01/20-23/87 GN-SITE HOURS: 71 i
CORRESPONDENCE ADDRESS: Brand Industrial Services, Incorporated ]
ATTh: Mr. C. W. Brown, President Construction Group 1420 Renaissance Drive Park Ridge, Illinois 60068 ORGANIZATIONAL CCNTACT: Mr. Clayton Brcwn, President 1 TELEPHONE NUMBER: (312)298-1200 l NUCLEAR INDUSTRY ACTIVITY: Brand Industrial Services, Incorporated (BISCO) has been in the nuclear plant fire protection business for 16 years and has provided materials and/or installed fire penetration seals in over 50 domestic ;
nuclear power plants. '
l ASSIGhED INSPECTOR: _ Fev . 5//[f7 l J. J. Pe1frosino, Program Development and Reactive ate Inspection Section (PDRIS)
OTHER INSPECTORS: T. L. Tinkel, Brookhaven National Laboratory J. M. Ulie, Reac or Inspector, Rlli I APPROVED BY: b ///P;7 Oate i
_J. C. tone," Chief, PDRIS, Vendor Inspection Branch INSFECTION BASES AND SCOPE:
A. BASES: 10 CFR Part 21 and Appen<tix f, to 10 CFR Part 50.
I B. SCOPE: 1) Obtain generic technical and testing information regarding silcone foam and silcone elastomer fire barrier penetration seals. 1'
- 2) Evaluate BISCO's QA program implementation.
PLANT SITE APPLICABILITY: Arkansas 1 & 2 (50-313/368); Callaway (50-483);
Clinton (50-461); Comanche Peak 1 & 2 (50-445/446); (continued on next page) 15
ORGANIZATION: BRAND INDUSTRIAL SERVICES, INCORPORATED PARK RIDGE, ILLINCIS REPORT INSPECTION N0.: 99901020/87-01 RESULTS: PAGE 2 of 10 PLANT SITE APPLICABILITY: (continued) Cook I & 2 (50-315/316); Cooper Station 323);Dresden 2 & 3 (50-237/249);Lavis-Besse (50-298); Crystal River (50-302); (50-346); Diablo Canyon (50-275/
Enrico Fermi (50-341); Fort St. Vrain (50-267); Ginna (50-244); Hatch (50-321); Hope Creek (50-354); LaSalle 1 & 2 (50-373/374); Limerick 1 & 2 (50-352/353); Maine Yankee (50-309); McGuire 1 & 2 (50-369/370); Millstone 1, 2, & 3 (50-245/336/423); Nine Mile Point 1 & 2 (50-369/370); Oyster Creek 1 (50-219); Palo Verde 1, 2, & 3 (50-520/529/530);
Peach Bottom 2 & 3 (50-277/278); Perry (50-440); Pilgrim (50-293); Quad Cities 1 & 2 (50-254/265); Rancho Seco (50-312); Robinson (50-261); Salem (50-272);
San Onofre 1, 2, & 3 (50-206/361/362); Seabrook (50-443); Shoreham (50-322);
St. Lucie 1 & 2 (50-335/389); Summer (50-395);Susquehanna 1 & 2 (50-387/389);
Three Mile Island 1 & 2 (50-289/320); Trojan (50-344); Turkey Point 3 & 4 (50-250/251); Vermont Yankee (50-271); Vogtle 1 & 2 (50-424/425); Washington Nuclear (50-397); Waterford 3 (50-382); Watts Bar (50-391); Wolf Creek (50-482); and Zion (50-295/304).
A. VIOLATIONS:
Contrary to Section 21.21, " Notification," of 10 CFP Part 21, EISCO failed to establish a written procedure to implement the regulations of 10 CFR Part 21 as imposed by Section 206 of the Energy Reorganization Act of 1974.
B. NONCONFORMANCES:
None.
C. UNRESOLVED ITEMS:
None.
D.
STATUS OF PREVIOUS INSPECT _ ION FINDINGS:
1.
Section 0.3 of BISCO report ho. 99901020/85-01, mentions a 1976 test that was conducted and accepted by American Nuclear Insurers (ANI) for various BISCO fire bcrrier penetration design configurations.
ANI later withdrew its acceptance of the test in a August 20, 1965 ANI letter transmitted to BISCO and several nuclear generating stations (NGS). This issue was reviewed during this NRC inspection.
Discussed below are the background and conclusions of this review.
Discussion - The technical basis for the ANI acceptance of BISCO's October,1976 9-inch silcone foam (SF) fire penetration seal testing was not fully documented and did not substantiate that the generic )
i 16
. . _ _ _ _ _ _ _ _ 1
ORGANIZATION: BRAND INDUSTRIAL SERVICES, INCORPORATED PARK RIDGE, ILLIN0IS REPORT INSPECTION N0.: 99901020/67-01 RESULTS: PAGE 3 of 10 -
ASTM-E119 and USNRC requirements had been met. Specifically, a review of the test reports and discussions with ANI and BISCO personnel determined that:
- a. Although no specific mention was made of the 9-inch SF penetration test failing the required hose stream test, statements within the report indicate that it did fail.
Therefore, ANI should not have accepted the particular configuration;
- b. ANI and BISCO have stated that a successful hose stream test was conducted on a second 9-inch SF specimen, as allowed by ASTH E-119, but according to ANI, "was never formally do'tu-mented," in 1976 when it was performed;
- c. ANI discovered the lack of objective evidence of the hose stream test in 1985, contacted BISCO and attempted a make-up test that failed; ANI subsequently notified ten NGS facilities of their acceptance withdrawal. The letter stated that the failed test was acceptable to the ANI for " insurance purposes only," but also stated their 2-1/2 hour rating acceptance may not be acceptable where a strict 3-hour rating is required; and
- d. Current industry practice allows fire barrier installer to utilize other installers tested design configurations to substantiate their installations. This practice creates the possibility that there are additional NRC licensees that utilized the failed October 1976 BISC0/ANI design configuration for their installed 9-inch SF installations, and are not aware of the prcLlem, since ANI may or may not be their insurance agency.
A related concern is the design parameters of other BISCO test reports. A review of three penetration seal test reports that BISCO stated would substantiate their recinded 1976 test acceptance by Ahl, '
were found to be very restrictive in their parameters; however, CISCO's statement in their ANI follow-up letter could imply that the NRC licensee's recinded penetraticr. seals were adequate and no additional review was required. The test report numbers are:
61500 Report No. 3001-03-B, dated May 19, 1960 .
BISCO Report No. 740-134, dated May 14, 1984 BISCO Report No. 748-165(3), dated August 9, 1985 17
_ _ _ _ _ _ _ ____ _______A _ - _ _ - _ - -_ _ _ _ _ _ _
ORGANIZATION: BRAND INDUSTRIAL SERVICES, INCORPORATED PARK RIDGE, ILLIN0IS REPORT INSPECTION h0.: 99901020/07-01 RESULTS: PAGE 4 of 10 In letters dated August 20, 1985 to various NGS facilities, and in another letter dated August 20, 1985 to BISCO, ANI advised of its withdrawal of three previously accepted BISCO fire barrier penetration seal design configurations. The ANI letter indicated that the subject designs involved penetrations for cable and pipe sealed with 9-inches of BISCO SF-20. ANI indicated that its acceptance, which was previously issued based on 1976 testing, was being withdrawn because a review indicated available evidence was insufficient to support the 2-hour and 3-hour ratings for these particular designs. SISCO and ANI stated that only one design was of a concern. The particular design configuration identified by ANI and 81S00 is for cable tray penetrations filled with 9-inches of SF with no permanant dams installed. These are typically wall penetrations and are 3-hour rated. The ANI review was performed following a request by Rancho Seco personnel in 1984 or 1985 for ANI to provide them additional documentation of Ahl's 1976 technical basis for its subsequent acceptance of SF penetration seals at their plant.
The original test of this cable tray configuration was one of a number of configurations tested at the same time for BISCO by ANI at the Portland Cement Association (PCA) Laboratories in Skokie, Illinois. This particular test is discussed in a BISCO report dated October 1976. The test report is referred to as PCA-76 for most applications. 1he test specimen passed the 3-hour fire endurance test of ASTM-E119 and its unexposed surface did not exceed the allowable ASTM-E119 temperatures. However, " flame through" occurred at 3-hours 1-minute and PCA-76 does not mention whether or not the hose stream test was performed, but references in other report sections indicate that it failed the requirea hose stream test. Both BISCO and ANI have stated that a hose stream test was performed on a second 9-inch SF specimen as allowed by ASTM-E119, and passed. ANI additionally states that the test was performed on the second specimen but that "nc formal documentation was e,er generated for the test."
Following the Rancho Seco request for additional test documentation to ANI, BISCO and ANI conducted a more severe fire endurance test for a make-up test. This test failed after 2-hours and 35 minutes and was the basis for ANI's August 20, 1985 letter. The test was more severe because the total cross sectional area of the cable was greater than the 1976 tests and additionally all of the cables had jackets of PVC which are more combustible than the original test cable jackets. The test was conducted at PCA on August 6,1985.
l 18
ORGANIZATION: BRAND INDUSTRIAL SERVICES, INCORPORATED PARK RIDGE, ILLINCIS REF0RT INSPECTION NO.: 99901020/87-01 RESULTS: PAGE 5 of 10 BISCO and ANI witnessed the test. The results are presented in CISC0 Test Report 748-183 (Specimen 2), dated August 6, 1985. The test asserrbly was subjected to the ASTM E-119 standard 3-hour fire endurance test and developed a burn through after 2-hours anc' 35 minutes. This was a failure and consequently no hose stream test was performed.
Therefore, as a result of no formal documentation to show that the 1976 hose stream test was performed, and the subsequent August 6, 1985 fire test failure, ANI issued their August 20, 1985 letter to ten nuclear plants that BISCO stated were affected, see list below. The August 20, 1985 ANI letter states, in part:
- a. "Since 1975. American Nuclear Insurers has reviewed and accepted for property insurance purposes only, over 200 fire stop systems...."
- b. "We recently found insufficient evidence to support the ANI acceptance of testing form issued to BISCO for test data, various (1976)
- c. "In an effort to verify the proper rating of this system, we conducted a fire test of a sample of the subject system at an independent test laboratory. The sample withstood the fire exposure prescribed in ASIN E-119 for a duration of 2-hours, 35 minutes...."
- d. "This fire stop system presents a substantial barrier to the passage of fire between fire areas and except for some unanticipated fire loading hazard, all existing systems are considered by ANI to be acceptable for property insurance purposes for the separation of areas...."
- e. "However, this system may not be adequate where there is a strict requirement for a 3-hour rated barrier for other than insurance purposes...." and
- f. "If it is necessary to upgrade this fire stop system for a 3-hour fire rating, the following suggested methods may be adequate:" c) " Apply a ANI/MAERP acce to the bottom side of vertical (floor)ptedseals damming and onboard both sides of horizontal (wall) seals;" or b) " Apply a protective coating over the foam that has been fire tested in accordance with the ANI/MAERP fire test standard for penetration seals...."
19
ORGANIZATION: BRAND INDUSTRIAL SERVICES, INCORPORATED PARK RIDGE, ILLINOIS REPORT INSPECTION NO.: 99901020/87-01 RESULTS: PAGE 6 of 10 ANI transmitted their August 20, 1985 letters to the following NGS facilities:
SEAL INSTALLATION SEAL INSTALLED FLOOR DAMS INSTALLED NANE IN FLOORS LEFT BY BISCO IN WALLS
- VC Summer Yes Yes Yes Rancho Seco Yes Yes Yes Davis Besse Yes Yes Yes
- Susquehanna Yes Yes Yes Hanford Yes Yes Yes k
Shoreham Yes Yes Yes
- Comanche Peak Yes Yes Yes
- Clinton Yes Yes Yes
- Palo Verde Yes No Yes Trojan Yes Yes Yes
- Plants receiving BISCO's followup letter Lfter the ANI August 20, 1965 letter, as stated by BISCO.
Note: No wall dams were left in place for the above plants.
Subsequent to ANI issuing its August 20, 1985 acceptance withdrawal letters, blSCO issued a followup form letter. BISCO stated that to the best of their knowledge only four of the plants noted above responded to the ANI letter. BISCO's form letter states, in part:
- a. Recently you received correspondence from ANI recinding their acceptance of fire testing conducted on two cable tray blockout designs by BISCO."
- b. "The reason behind this action had to do with the large scale fire test that was conducted a.id because of its size requiring separate hose stream tests that were originally submitted to ANI and received their certified acceptance."
l 20
ORGANIZATION: BRAND INDUSTRIAL SERVICES, INCORPORATED PARK RIDGE, ILLINOIS REPORT INSPECTION N0.: 99901020/87-01 RESULTS: PAGE 7 of 10
- c. " Subsequent to the test, BISCO experienced a fire at their facilities...and apparently during this time the subject hose stream test was either destroyed or lost. This was brought to light when ANI, unable to locate their file copy, requested ;
a replacement from BISCO."
- d. "ANI requested BISCO to conduct a new test to their present Y standards, which experienced a burn-through at 2-hours and 35 minutes. However, ANI has stated they will accept all fire barriers of this BISCO design..." and;
- e. " BISCO has testing documentation that substantiates the fire seal design that failed the ANI fire test standards, does meet and surpass the test standards of ASTM-E119 and the NRC."
In summary, it is perceived that some NRC licensees may be relying on ANI for overall acceptance; though ANI is looking at the fire barriers only in regard to insurance purposes and not NRC licensing requirements.
- 2. NRC previously reviewed a 6-inch SF fire barrier issue regarding the Salem Unit 1 NGS facility. NRC report 99901020/85-01 concluded that the test data satisfactorily supported, as required by ASTM-E119, the installed subject seals at Salem; however, the previous NP,C report did not address the required ASTM-E119 hose rtream tests.
Discussion - It was revealed that an NRC fire protecticn staff revicw had approved a Salem deviation request to its Appendix A to the NRC #
Branch Technical Position 9.5-1 guidelines. The deviation acceptance 4 by the NRC exempted Salem from having to perform a hose stream test following its fire endurance test. Therefore, the Salem Unit 1 NGS f acility appears to have an adequate technical basis for its installed 6-inch SF seals and are acceptable according to the NRC criteria.
However, similar to issue 1. above, if these two test reports, SEMC0 PR-55 (6/76) and PSE&G (AISCO 6/76), were adopted by use for seal qua#ications at another facility, the potential would exist for an unquaified basis. Specifically: -
- a. The ASTM-E119 required hose stream test would have to be performed or exempted by the hRC; 21
ORGANIZATION: BRAND INDUSTRIAL SERVICES, INCORPORATED PARK RIDGE, ILLIN0IS REPORT INSPECTION NO.: 99901020/87-01 RESULTS: PAGE 8 of 10
- b. The 6-inch SF seal as built configurations must have the damming boards installed on both sides as shewn in the tested configurations.
E. OT_HER FINDINGS OR COMMFNTS:
- 1. 10 CFR Part 21 ,
Observations determined that BISCO did not have the required proce-dure posted that they adopted pursuant to the provisions of 10 CFR Pcet 21. Discussions were conducted with 51500 concerning their requirements under 10 CFR Part 21 and it was revealed that they had not established the required procedure.
Within three weeks after the inspection was completed, 81500 provided the inspector with a copy of their 10 CFR Part 21 procedure that was generated in compliance with the regulations.
- 2. Facility Tour BISCO provided the NRC inspectors a tour of their warehouse facilities. During the tour, it was noticed that some of BISCO's M&TE devices were not currently valid regarding calibration due dates.
However, a review of their QA manual determined that the specific devices were not required to be calibrated for use by their QC personnel. It was noted to the QA manager that the area's that we noted as needing calibraton contrcl appeared to.be an important link in the BISCO process control; however, BISCO stated that the final verification was performed by QC.
- 3. V.C. Summer Nuclear Plant FaciSity The South Carolina Electric and Gas Company's (SCEG) V.C. Summer nuclear plant facility was visited on February 25-27, 1987 as a result of this inspection. Some of the information obtained during this inspection, in conjunction with previous NRC information, indicated that the potential exists for a nuclear plant to have penetration seal test reports and documentation that will not substantiate the validity of the particular plants installed penetration seal configurations. Therefore, the NRC performed a follow-up inspection at the SCEG facility, i
22
ORGANIZATION: BRAND INDUSTRIAL SERVICES, INCORPORATED PARK RIDGE, ILLIh0lS
\
REFORT INSPECTION NO.: 99901020/87-01 RESULTS: PAGE 9 of 10 It was determined, from a review of te: t report documentation, in-plant installed penetration seal dimensional measurements of blockouts, and cable tray / conduit sizes that some design parameter values in installed penetration seals exceed the values or range of ,
values validated by the test report being cited to cualify these 's particular insttlled penetration seal design configurations.
SCEG is taking corrective action and performing a review of their fire prctection system. This effort is being coordinated through l
the NRC Region II cffice.
F. PERSONS CONTACTED:
BISCO:
- Clayton Brown, President
- Thomas Gilmore, Vice-President
D. Kubicki, NRR/FBPE L. Whitney, IE/0PRB J. Wermiel, NRR/PBPE
- Attended Exit Meeting ANI:
P. Giaccaglia, Senior Staff Engineer W. Holmes, Director / Technical Review J. Carney, Vice President / Technical Review PCA:
R. Hall, Engineer ASTM:
R. Sansont, Staff Engineer '
23
ORGANIZATION: ' RAND B INDUSTRIAL SERVICES, INCORPORATED PARK RIDGE, ILLIN0IS REPORT INSPECTION NO : 99901020/87-01 RESULTS: PAGE 10 cf 10 G. D,0_C,UMENTS EXAMINED:
- 1. ANI letter, August 20, 1985 - withdrawal of previous acceptance.
- 2. BISCO QA Manual - certain secticns.
- 3. BISCO Procedure, dated February 5, 1987 - 10 CFR Part 21 Procedure.
- 4. BISCO letter, dated September 16, 1985 - to TUGCO, following the ANI August 20, 1985 letter.
5, BISCO letter, dated November 13, 1904 - BISCO response to TUGCC letter No. CPPA-41,594.
i "f . NRC Appendix A - to Branch Technical Position 9.5-1, dated August 23, 1976.
- 8. NRC Standard Review Plan - 9.5-1, dated July 1981.
- 9. NRC Generic letter 10, dated April 25, 1986,
- 10. BISCO Test Report " Fire Endurance Test in BISCO Penetration Seal Systems in a Cdncrete Floor usirig BISCO Systems SF-20 ano SF-150L Silicone Polymers," dated October 1976 (PCA-76).
- 11. 8I5C0 Test Report #748-183-(Specimen 2 and 3), "3-hour Fire Test of Two Cable Tray Seal Configuration," dated August 6, 1985 and August 9, 1985, respectively.
- 12. BISCO Test for Public Service Electric and Gas Com (PSE&G)-
undated, regarding six inch SF-20 seals at Salem (pany).
1976
- 13. SEMC0 Test Report "SEMC0 PR 855 RTV Silicone foam Sealant in Concrete Floor," dated June 28, 1976.
- 14. BISCO Test Report #748-134, dated May 14, 1984. Overall size 2.5' x 2.5' that was divided in half. One side with SF-20 and the other with SE-Foam, both sides having one cable tray and one conduit.
- 15. BISCO Test Report #3001-003, dated May 19, 1980. Overall size 2.5' x 2.5', wall, with 9" of SF-20 and no dam, 24
ORGANIZATION: COMBUSTION ENGINEERING INCORPORATED WINDSOR, CONNECTICUT REPORT INSPECTI0f! INSPECTION N0.: 99900002/86-01 DATES: 7/21-25/86 ON-SITE HOURS: 76 CORRESPONDENCE ADDRESS: Combustion Engineering Incorporated ATTN: Mr. H. V. Lichtenberger Vice President - Nuclear Fuel 1000 Prospect Hill Road Windsor, Connecticut 06095 ORGANIZATIONAL CONTACT: Mr. P. Ferwerda TELEPHONE M.lMBER: (203) 285-3352 g NUCLEAR INDUSTRY ACTIVITY: Nuclear fuel assembly supplier for Combustion Engineering (CE) and Westinghouse designed reactors.
4 ASSIGNED INSPECTOR: b N, hN
[p93. R. L. Cilimberg, Special Projects Inspection k!l6!T(P Date Section (SPIS)
OTHER INSPECTOR (51) : C. M. Abbate, SPIS
/
C. Hirayama Consultant APPROVED BY: M N6 hn W. Craig, Chief, SPIS, Vef. dor Program Branch ate INSPECTION BASES AND SCOPE:
A. BASES: 10 CFR 50, Appendix B and 10 CFR 21.
- 8. SCOPE: This inspection was made to review fabrication, inspection, and testing activities.
PLANT SITE APPLICABILITY: PWR facilities with fuel supplied by Combustion Engineering.
25
ORGANIZATION: COMBUSTION ENGINEERING INCORPORATED WINDSOR, CONNECTICUT REPORT INSPECTION NO.: 99900002/86-01 RESULTS: PAGE 2 of 15 A. VIOLATIONS:
None.
B. NONCONFORMANCES:
- 1. Contrary to Section 6.6.2 of the CE Quality Assurance Manual (0AM),
Revision 1, dated January 22, 1986, ink changes were made to Sections B.4.0 and B.5.0 of Operation Sheet (0.S.) Number 925, " Leak Test,"
Revision 32, dated June 1,1984, without the proper approvals.
- 2. Contrary to Section B.7.0 of 0.S. Number 945, " Leak Test," Revision 32, dated June 1, 1984, the helium leak test was not being conducted for a minimum of 30 seconds as required.
C. UNRESOLVED ITEMS:
None.
D. STATUS OF PREVIOUS INSPECTION FINDINGS:
Not applicable.
E. OTHER FINDINGS AND COMMENTS:
- 1. Entrance and Exit Meetings The Combustion Engineering (CE) representative was informed of the scope of the inspection during the entrance meeting on July 21, 1986.
The inspection findings and observations were summarized during the exit meeting on July 25, 1986.
- 2. Helium Leak Test a) Procedure Review The inspectors reviewed 0.S. Number 945, " Leak Test," Revision 32, dated June 1, 1984. The procedure contains the steps used for start-up and shut-down of system equipment, calibration of the equipment, testing of fuel rods, processing and retesting of leaking rods, cleaning of test chambers, and processing of pape rwork. During the review, the NRC inspectors noted that Sections B.4.0 and B.S.O of the 0.S. hSd been changed from loading two fuel rods into the test chamber to loading four 26
'0 ORGANIZATION: COMBUSTION ENGINEERING 1 INCORPORATED WINDS 0R, CONNECTICUT REPORT INSPECTION l N0.: 99900002/86-01 RESULTS: PAGE 3 of 15
-l fuel rods into the test chamber. This change was made with 1 an ink pen and did not have the approval of Manufacturing, l Engineering, and Quality Control Engineering, as required by -l Section 6.6.2 of the CE QAM. When CE was informed of the j unapproved ink changes by the NRC inspectors, the proper
- approvals were obtained.
Nonconformance 86-01-01 was identified in this area.
b) Observation of Testing l
The helium leak test consists of putting two to four fuel. rods into the test chamber. A vacuum is then exerted on the test chamber. The actual monitoring for helium leaking from the inside of a fuel rod occurs after a vacuum is achieved and the
" Start Test" button is pressed. The leak detector meter measures helium leakage and is to be monitored by the operator. The required duration of the test is a minimum of 30 seconds. The meter must be stable before accepting or rejecting a reading and the reading must be below the specified value in order for the test to be acceptable. The NRC inspectors observed that the operator was not performing the leak test for the j required 30 seconds during multiple tests. When this matter was brought to the attention of the QC Manager and the Inspection Supervisor, the operator was instructed to assure that the specified minimum test duration of 30 seconds elapsed before accepting or rejecting the rods. The'QC Manager also indicated that a Deviation Notice (DN) would be written, the entire fuel rod cart would be retested, and other operators would be l
instructed on the importance of assuring that the test is l
performed for a minimum of 30 seconds.
Nonconformance 86-01-02 was identified in this area.
- 3. Leaking Fuel The NRC. inspectors met with Ian Rickard, the CE Project Manager who incidents of leaking fuel at a number was responsible of pressurized for reactors water evaluating (PWR) that use CE fuel. San Onofre 3 (SONGS-3), St. Lucie 1, Maine Yankee, Calvert. Cliffs 1, and Arkansas.
One-Unit 2 havea ~ ll exhibited incidents of 1eaking fuel in varying degrees from June, 1982, to the present.
l i
27
ORGANIZATION: COMBUSTION ENGINEERING INCORPORATED WINDSOR, CONNECTICUT REPORT INSPECTION NO.: 99900002/86-01 RESULTS: PAGE 4 of 15 CE has performed an in-depth evaluation of SONGS-3 cycle 1 which contained 105 leaking fuel rods in 23 fuel assemblies. The evaluation covered a number of possible fuel failure mechanisms such as pellet-cladding interaction, corrosion, fabrication defects, handling damage, spacer-grid fretting, and debris in the primary system. An in-depth analysis of recent fuel failures at other facilities has not been performed. Almost 50% of the leaking fuel rods in SONGS-3 were in Assemblies A-004 and A-005 which prompted CE to look at the fabrica-tion of these assemblies. The leaking fuel rods were believed to be a result of primary hydriding because Iodine 131 was detected early in the fuel cycle and failures occurred from the inside of the rods based on visual examination. Records of hydrogen analysis for the fuel pellets in the failed rods indicated that the pellet hydrogen content was considerably below the maximum limit in the CE specifica-tion. A similar examination of pellet density values indicated adherence to the CE specification.
As a result of their evalua' tion, CE concluded that the failed fuel rods in SONGS-3 were caused by moisture inside the fuel rods due to improper drying and a low vacuum during fabrication. This resulted in hydriding of the zircaloy and subsequent perforation of the cladding during the start of cycle 1. Based upon this evaluation CE has implemented several corrective actions: Vacuum gage calibra-tion procedures were revised, a lock box was installed on vacuum gages at the end cap welding stations, and a new dryer was installed to dry tubes after cleaning.
- 4. End Cap Welding a) Lower End Cap Weld The lower end cap weld is described in 0.S. Number 500, " Weld 1st Cap," Revision 42, dated hay 29, 1986. The procedure outlines how the welding equipment should be set-up and shut-down, the daily check's that are to be made on the equipment, and the steps for making the weld. An addendum to the 0.S.
lists the amperage limits and gage and control levels for each contract number and tooling required for each contract.
The process starts by placing a lower end cap and zircaloy tube in the welding chamber. The chamber is closed and pumped down to the vacuum specified in the welding procedure. In order to prevent the operator from lowering the specified vacuum, a locked cage has been placed over the vacuum gage.
This was done as a result of the leaking fuel problem identified at San Onofre (as discussed in E.3). When the specified vacuum 28
ORGANIZATION: C M MTION ENGINEERING INCORPORATED W M SOR, CONNECTICUT REPORT INSPECTION NO.: 99900002/86-01 RESULTS: PAGE 5 of 15 is reached, the chamber is backfilled with helium and the end cap is welded in the tube. This process is performed on each
~o f the tubes in the cart and on a specified number of sample rods. At the conclusion of welding the lot, en inspector is I notified and the welds are visually examined.
b) Upper End Cap Weld The upper end cap weld is made after 'the fuel _ pellets have been loaded into the tubes and is described in 0.S. Number 852,
" Weld 2nd Cap," Revision 55, dated June 10, 1986. This procedure is similar to 0.S. 500 and also outlines how the welding eauip-ment should be set-up and shut-down, the daily checks that are to be made, and the steps for making the weld.
The upper end cap weld procedure follows the same basic ste, as the lower end cap weld procedure. The differences are that a spring is inserted into_ the tube before the weld is made and 6 strip recording chart which records room temperature and chamber pressure must be checked by the operator. A cage has also been installed around the vacuum gage to eliminate tampering.
Upon completion of observations of the end cap welding and review of the two procedures, the inspectors determined that the proce-dures were being followed by the operators.
No items of nonconformance or unresolved items were identified in this area.
- 5. Zircaloy Tubing a) Tubing Specifications CE utilizes zircaloy tubing to manufacture fuel rods. CE Specification Number 00000-FMD-301, " Specification for Zircaloy-4 Fuel Rod Cladding Tubes," Revision G1, dated '.
October 27, 1981, outlines the scope, the responsibilities of the seller and purchaser, the applicable references, the ,
chemical, mechanical, hydride orientation, surface finish, and dimensional requirements, the quality assurance program and test requirements, packing, marking and shipping requirements and documentation required from the seller. The specification references ASTM 353-77A and identifies wrought Zircaloy-4, ASTM designation UNS AR60804, as defined in ASTM 353-77A, as the material to be used in the fabrication of the tubes.
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ORGANIZATION: COMBUSTION ENGINEERING INCORPORATED WINDSOR CONNECTICUT REPORT INSPECTION NO.: 99900002/86-01 RESULTS: PAGE 6 of.15 b) Purchase Orders The NRC inspectors reviewed purchase order (P0) 9530077, dated May 16, 1985, with respect to the zircaloy tubing. This P0 was from Baltimore Gas and Electric Co. (BG&E) to CE Nuclear >
Products Manufacturing for Batch L fuel bundle assemblies to be used for the Calvert Cliffs 1 reload. The P0 included the final design drawings and material specification numbers to be used and included additional instructions to be followed during the
-manufacturing of the fuel assemblies'. CE Specification Number -;
00000-FCD-0100, " Standard Engineering Specification for 14 x 14 ;
Fuel Assemblies," Revision'01, dated November 26, 1979, was referenced as the basic specification for the manufacturing of 4 the fuel assemblies. This specification was reviewed with respect to the tubing and was found to include the scope, the seller's and purchaser's responsibilities, the applicable documents, the manufacturing requirements and reports, records, and documents to be furnished by the seller. Specification ]
1 Number 00000-FCD-0100 references and requires the cladding to l be fabricated in accordance with Specification Number 00000- '{
FMD-301(describedabove). j P0 9530077 imposed 10 CFR Part 21 on CE and also required CE to obtain mean and upper and lower 95/95 confidence limit rheasurements of the tubing outside diameter (00), wall thick-ness and ovality on at least 300 tubes.
P0 Number 9690036-05303, dated March 19, 1986, to Sandvik i Special Metals (SSM) from CE for Zircaloy-4 seamless tubing '
was also reviewed. The P0 was issued for the BGSE fuel assembly order and outlines the dimensional tolerances of the tubing, the special requirements and/or exceptions to the specification, certification requirements, delivery .
requirements and SSM procedure submittal requirements. The }'
P0 stated that 10 CFR Part 21 was applicable and described the 10 CFR Part 21 reporting requirements. The P0 also ,
required SSM to provide CE with the 00, wall thickness, and ovality measurements of 300 randomly selected tubes (as stated above).
This PO was written in accordance with Section 4.0, " Procurement Document Control," of the CE QAM and included the appropriate ;
specifications, standards, and drawing requirements.
i 30 ,
1 i
l ORGANIZATION: COMBUSTION ENGINEERING INCORPORATED WINDSOR, CONNECTICUT REPORT INSPECTION NO.: 99900002/86-01 RESULTS: PAGE 7 of 15 c) Material Certifications .
The requirements outlined in CE Specification Number 00000- ]
I FMD-301 and imposed on SSM are listed on SSM Form PS-02-04, Revision 7. SSM takes random samples from each lot and performs tests on those samples to ensure the lot meets the requirements.
I These chemical and mechanical tests consist of chemical impurity l
testing, corrosion resistance testing, grain size testing,
' hydride orientation testing, ingot chemical analysis. (taken at i top, middle and bottom), and mechanical property testing. For j each chemical and mechanical test performed on the samples the '
specification reference, acceptance criteria, and results are listed. Ultrasonic testing is performed on all tubes, and visual i
! inspection is performed on the tubes as an overall quality check.
Other information contained on SSM Form Number PS-02-04 includes the supplier and customer names, specification numbers, ingot vendor name, heat number, lot number, PO number, certi'fication number and date and quantity shipped. Also on the form is an area for CE to review the release checks and accept or reject the submittals made by SSM.
In a letter dated June 9, 1986 to the CE QA Manager from the ;
Quality Assurance Engineering Manager, the required dimensional I data submitted by SSM was documented. The letter contained the j dimensional tolerances as specified in the P0, the sample size, the mean, the standard deviation, and the 95/95 confidence limits for the wall thickness, OD and 00 ovality.
l Other information related to the tubing which was reviewed !
during the inspection included Quality Control Material Release Number 3690036-4, dated June 16, 1986, the Standard Surveillance Operation Sheet and Trip Report, dated June 6, 1986, the Surveillance Checklist for Zircaloy-4 Cladding, and the Shipping Release-Quality Control (Vendor). This information was filled out by both CE and SSM and recorded the lot numbers, heat numbers, ingot numbers, the tests performed on the tubes and if the results were acceptable.
Although CE does not perform receipt inspection (except for shipping damage) or chemical and mechanical tests, they do perform scheduled audits of SSM's quality program and perfort surveillance of SSM during the fabrication of a CE order.
During.the release of material from the supplier to CE, the tubing and paperwork are reviewed and controlled in order to 31
ORGANIZATION: COMBUSTION ENGINEERING INCORPORATED WINDSOR, CONNECTICUT REPORT INSPECTION NO.: 99900002/86-01 RESULTS: PAGE 8 of 15 maintain traceability. Traceability is achieved through heat, ingot, and lot numbers. The NRC inspectors observed that the forms which contained acceptance test results were filled out with the proper identification numbers and were reviewed and approved by the appropriate personnel, e) Cleanliness of Tubing After the tubes cre received at CE and are receipt inspected for shipping damage, they undergo a cleaning process. This process is described on Shop Traveler " Tube Cleaning,"
Revision 4, dated June 12, 1986, and includes six separate steps which are subsequently described in 0.S. Numbers 106, 107, 103, 104, 2402, and 1242.
0.S. Number 106, " Vapor Degreaser," Revision 1, dated December 12, 1969, 0.S. Number 107, " Detergent Clean,"
Revision 1, dated April 7, 1970, 0.S. Number 103, " Process Water Rinse," Revision 1, dated April 7, 1970, and 0.S.
Number 104, " Demineralize Water Rinse," Revision 2, all have steps which describe how to perform equipment shut-down, and how to run the process. 0.S. Number 2402, "0ven Dry,"
Revision 3, dated June 19, 1906, outlines the safety precautions, set-up parameters, includin and the production process.g 0.5.
temperature Number 1242, and time settings,
" General Inspection Instructions," Revision E3, dated June 18, 1986, outlines the minimum requirements for each inspecticn speration.
The areas delineated include the computer system in use for traceability of fuel and poison rods, review of shop / work instructions, and inspection procedure general instructions.
The general instructions include part acceptance, repair procedures, part rejection, and sampling plans based on lot size.
Although an entire cleaning process was not observed, the inspectors did observe that the clean, dry tubes were covered with cheesecloth, that both ends were plugged to prevent contamination from outside sources and that the proper revision of the 0.S. Number was present at each work station.
f) Audit of Tubing Supplier The inspectors reviewed a CE Vendor Quality Audit Report of SSM, dated January 16, 1986. The report consisted of a checklist, a comment area, and attachments, which specifically listed the areas audited in more detail than the checklist and problems 32
ORGANIZATION: COMBUSTION ENGINEERING INCORPORATED !
WINDS 0R, CONNECTICUT REPORT INSPECTION RESULTS: PAGE 9 of 15 N0.: 99900002/86-01 1 which were identified. The inspector determined that the audit was performed and the report was written in accordance with the requirements prescribed in Section 18.0, " Audits" of the CE QAM and that SSM met the requirements. set forth in QC-14-09, " Quality Assurance Program Requirements for Suppliers to Combustion Engineering Nuclear Products Manuf acturing-Windsor."
I No items of nonconformance or unresolved items were identified in this area.
- 6. Spacer Grids a) Spacer Grid Assembly The inspector observed the assembly of the spacer grids. A spacer grid is assembled by connecting four zircaloy plates and forming a square. Additional plates are added and the grid is turned over and thirteen or fifteen more plates are attached in rows at the 90 angles to the first set of rows. The different pieces used in the process are segregated in a marked. holder to assure <
that the assembler uses the correct part in the assembly. 0.S.
Number 1799, " Bench Assemble and Weld,14 x 14,16 x 16 Spacer Grids," Revision 18 dated November 27,'1985, describes this l process. The inspector verified that the 0.S. was present at the work station, and the assembler was following the procedure, b) Grid Intersection Veld 0.S. Number 1799 also outlines the steps'to be followed when welding the grid intersections on the' spacer grids. The welds are performed using a computerized welding machine which, after being programmed, welds each intersection.
The operator oversees the welding and watches a computer screen to assure that the weld is made at the correct loca- ,
tion and is of the correct dimensions.
c) Welder Qualifications The inspector reviewed the qualifications of three welders. Each welder is classified as either Level I, II, or III; Level III being the most qualified. The welder's level of qualification, the Welding Procedure Specification (WP3) number which he is qualified to perform and his requalification dates are listed 33
ORGANIZATION: COMBUSTION ENGINEERING INCORPORATED WINDSOR, CONNECTICUT REPORT INSPECTION NO.: 99900002/86-01 RESULTS: PAGE 10 of 15 on the " Welder Qualification" form. A " Weld Qualification Log" and " Engineering Notices" were reviewed for each of the three welders. These forms contained the welder's name, level, WPS number to which he is qualified, the date of the original qualification, requalification dates (yearly), and the 0.S.
number and lot number used for the weld. This is done to assure that the welder is qualified to perform ' specific welds ,
and to assure that he remains qualified. I
.1 The inspector determined that the three welders were qualified I to perform the welds specified on their qualification forms and '
through interviews with the welders, and observation of their work, determined that they followed the 0.S. and WPS forms.
No items of nonconformance or unresolved items were identified in this area, i 1
- 7. Fuel Bundle Assembly During the inspection, the final assembly of the fuel bundle was observed. The procedure used during the process was 0.S. Number 1524,
" Fuel Bundle Loading and Inspection," Revision 43, dated April 10, 1986.
This process began with one row of fuel rods being unloaded from the prestack tray and placed on the pushing table. The pushing table ,
has an indentation on it for each fuel rod. The rod serial numbers 1 on the bottom end cap were read off and entered inco the computer )
to assure that the correct rod was being placed in the proper grid 1 location, if the rod is in an incorrect location, the computer, which has been programmed for the acceptable rods, will display an error. The operator punches the number in again and if an error appears again, a DN is written and must be dispositioned before the l
fuel bundle is released.
l After the rods in the row were indicated as being in the correct ';
locations, they are wiped clean and pushed through the assembly and spacer grids. At the conclusion of each row being pushed into the bundle, the rods are checked for proper fit.
The entire process is repeated until all the fuel rods have been loaded into the bundle. The fuel bundle is then visually inspected and placed in a cleaning chamber which rinses and dries the fuel l
bundle. The upper end fitting is installed and the finished fuel assembly is ready for the final quality checks made before being shipped.
34
= _______w
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ORGANIZATION: COMBUSTION ENGINEERING INCORPORATED WINDSOR, CONNECTICUT q l
REPORT INSPECTION RESULTS: PAGE 11 of 15 J NO.: 99900002/86-01 There are two basic quality checks performed on'the final fuel assembly. The first is a free path dimensional check. Four rollers i J
forming a square are slid up the assembly. The rollers are set at the dimensional tolerances specified in the P0 and specifications and if any of the dimensions are-exceeded, the operator can read the amount by which it is out of tolerance on the dial indicators. This check is performed to assure that the fuel assembly is straight and no bowing has' occurred during manufacturing.
The second test is an overall visual examination of the fuel assembly by a quality control inspector. This inspection consists of examining the exposed surfaces of the fuel rods, grids, and end fittings for any unacceptable conditions such as scratches, loose material on grids, rust, and grease, and examining all spacer grid corner velds to assure that they are free of cracks. If no problems are identified, the fuel assembly is either wrapped in plastic and held in storage or placed into a shipping container which holds two fuel assemblies.
The inspectors observed the personnel following the procedure and no items of nonconformance or unresolved items were identified in this area.
- 8. Fluoroscope Inspection CE performs fluoroscope inspection on all the fuel rods. 0.S. Number 990, " Fluoroscope Inspection," Revision 19, dated February 20, 1986, delineates the process. The 0.S. describes how to perform equipment start-up and how to perform the inspection.
The fluoroscope inspection consists of passing the fuel rod through a chamber. As the fuel rod is passed through the chamber, the operator is able to see an image of the fuel rod. The 0.S. outlines what the operator should see and also has a chart of conditions which would result in rejection of the rod, the permitted corrective action and the required action when a deviation is found. In most cases When a deviation is identified, the rod is rejected and a DN written.
l The inspectors determined that the operator was performing the fluorscope inspection as described in the 0.S. and no items of nonconformance or unresolved items were identified.
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i ORGANIZATION: COMBUSTION ENGINEERING INCORPORATED WINDSOR, CONNECTICUT REPORT INSPECTION NO.: 99900002/86-01 RESULTS: PAGE 12 of 15
- 9. C_hemical Analysis of Fuel a) Measuring and Test Instruments The NRC inspectors reviewed procedures and practices related to the control of measuring and test instruments used in the analytical laboratory to determine elements and impurities in oxides of uranium. The procedures discussed in (b) below specify the steps for using the measuring and test equipment. i Standards are used to verify the accuracy of the instrument i readings prior to performing an analysis of a production sample.
b) Procedure Review The NRC inspectors reviewed five proc 6dures which covered:
spectrographic analysis of low boron in uranium dioxide, rare earth and thorium in uranium oxides, and comon impurities in uranium dioxide; chromatographic analysis of carbon in uranium oxides, and gravimetric determination of oxygen to uranium ratio and percent uranium in uranium dioxide. The procedures had been properly reviewed and contained the required signature approval. The procedures were complete but contained the incorrect information discussed in (c) below. These procedures were revised during the inspection, c) Observation of Analysis The NRC inspectors observed technicians performing sample analyses utilizing the procedures discussed above. Observa-tions are itemized below:
(1) A technician used an ARL spectrometer to analyze uranium dioxide for impurities in accordance with Procedure Number 10, " Standard Operating Procedure for the Emission Spectro- l graphic Detenninations of Common Impurities in Uranium !
Dioxide," Revision 2, dated February 18, 1982. The j procedure specifies that the sample being converted from U0, to U3g0 be heated to 450 1 25 C. The technician was observed to heat the sample in a muffle furnace at a 1 maximum temperature of 540 C. While the higher temperature does not adversely effect the test results, CE issued Revision 3 of the procedure during the inspection to permit '
use of the higher temperature.
36
ORGANIZATION: COMBUSTION ENGINEERING INCORp0 RATED WINDSOR, CONNECTICUT REPORT INSPECTION NO.: 99900002/86-01 RESULTS: PAGE 13 of 15 (2) Section 8 of Procedure 10, Revision 2 states that the overall precision of the method is approximately plus or minus 25 percent of the amount present. An examination of ;
the calibration data by the NRC inspectors showed that the precision (numerical magnitude) of the data was significantly larger than plus or minus 25 percent for some of the elements such as iron. The spectrographic data-for a standard deter-mined on June 27, 1986, shows a difference of as much as 100 percent from that specified by the supplier of the' standard ,
sample. Revision 3 of Procedure 10 was revised during the inspection and notes that the precision of the method may vary depending upon the impurity element being analyzed and the working range of the standard curve for each element.
(3) A spectrograph is used to analyze uranium oxide for rare ;
earths and thorium in accordance with Procedure Number 24, ;
! " Standard Operating Procedure for Spectrographic Determina- l tion of Rare Earths and Thorium in Uranium 0xides," Revision 0, dated September 2, 1983. The analysis uses calibration curves drawn from measurements on standard samples. The curves for the rare earth elements are drawn through five points at 0.5 to 5 micrograms while that for thorium is-drawn through five points at 0.1 to 4 microgram. Examination i of the calibration curves indicated that the curves for most cf the elements increased abruptly at the highest content of the element. Consequently, if the quantity of these elements in a uranium oxide sample were greater than 5 micrograms for the rare earth, and greater than 4 micrograms i
for thorium, the extrapolation of the calibration curve l to these high concentrations becomes highly questionable.
Discussions on this matter with CE personnel indicated that additional determinations are made under these circumstances ,
with reduced quantities of UG,, samples, so that the spectro- '
graphic readings fall within the valid region of the calibration curves. Procedure 24 Revision 0, did not clearly define the method for the redetermination with l the reduced quantity of uranium oxide, so that accurate determinations may be made for these elements. CE issued Revision 1 of the procedure during the inspection to ensure that the correct curves are used for the quantity of sample chosen.
37
ORGANIZATION: COMBUSTION ENGINEERING INCORPORATED WINDSOR, CONNECTICUT REPORT INSPECTION N0.: 99900002/86-01 ! RESULTS: PAGE 14 of 15
- 10. Moisture and Hydrogen Control a) Measuring and Test Instruments The NRC inspectors reviewed procedures and practices for the control of measuring and test instruments used to determine hydrogen and moisture content in uranium dioxide fuel pellets.
The procedures discussed in (b) below specify the steps for using the measuring and test equipment. Standards are used to verify the accuracy of the instrument readings prior to performing an analysis of a production sample, b) Procedure Review The NRC inspectors reviewed four procedures which covered:
determination of hydrogen in uranium dioxide pellets, moisture in uranium dioxide, hydrogen analysis of special fuel, and moisture analysis of uranium dioxide powder. The procedures had been properly reviewed and contained the required signature approval. The procedures were technically correct and complete with the exception of the wrong units being used on a reporting form discussed in (c) below. The reporting form units were corrected during the inspection.
c) Observation of Analysis The NRC inspectors observed technicians performing sample analyses utilizing two of the procedures discussed above. Observations are itemized below:
(1) A technician used a Leco hydrogen analyzer to analyze uranium dioxide pellets for hydrogen content in accordance with Procedure Number 12. " Determination of Hydrogen in Uranium Dioxide Pellets," Revision 1, dated February 27, 1977. The steps of the procedure were properly followed without deviation and standards were analyzed to verify the accuracy of the method.
(2) A technician used a hydrogen analyzer to analyze a special fuel for hydrogen content in accordance with 0.S. Number 2144, " Hydrogen Analysis-SCE Fuel," Revision 2, dated October 2, 1985. The procedure was technically correct and complete with the exception of the reporting form entitled " Hydrogen Analysis Sheet" Figure 11. Figure 11 incorrectly indicated millimeters of mercury for 38
ORGANIZATION: COMBUSTION ENGINEERING INCORPORATED WINDSOR, CONNECTICUT ,
i REPORT INSPECTION NO.: 99900002/8j-01 RESULTS: PAGE 15 of 15 barometric pressure instead of the centimeters of mercury being measured by the technician. The reporting form was revised to specify the correct units.
- 11. Sampling and Density Measurement The NRC inspectors reviewed 0.S. Number 1487, " Chemistry Sampling,"
Revision 14, dated November 30, 1984, and 0.S. Number 1475, " Chamfered Fuel Pellet Inspection," Revision 29, dated June 18, 1986. These procedures were found to_be technically correct and complete.
Observations of technicians using these procedures found that the l steps of the procedures were properly followed. j i
No items of nonconformance or unresolved items were identified in j this area. i
- 12. Qualification of Chemistry Technicians 1
The NRC inspectors reviewed Procedure Number QC-14-20, " General !
Procedure for Analytical Chemistry Technicians," Revision 2, dated June 24, 1986, and observed the performance of personnel who had been qualified to this procedure. Technicians at CE are selected on the basis of passing examinations utilizing the skills required for effectively performing the operations affecting accurate chemical enalyses of samples submitted to the analytical laboratory. The i NRC inspectors determined that the Technicians were well qualified to perform the analyses that were observed. The Technicians adhered ,
rigidly to the steps of procedures thac were utilized. '
No items of norconformance or unresolved items were identified in this area.
F. PERSONS CONTACTED:
- G. Chalder E. Chan
- P. Ferwerda B. Sharp
- W. Graves A. Mistos H. Nelson R. Davis I. Rickard D. Byerly S. Hanson I. Corser P. Nelson F. Magnan
- Attended exit meeting.
39/40
ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE, CALIFCRNIA REPORT INSPECTION INSPECTION NO.: 99900403/87-01 DATES: 04/06-9/87 ON-SITE HOURS- 10P CORRESPONDENCE ADDRESS: General Electric Company Nuclear Energy Business Operations ATTN: Mr. N. L. Felmus, Vice President and General Manager 175 Curtner Avenue San Jose, California 45125 ORGANIZATIONAL CONTACT: Mr. J. J. Fox, Senior Program Manager TELEPHONE NUMBER: (408) 925-6195 -
NUCLEAR INDUSTRY ACTIVITY: General Electric Company's, Nuclear Energy Business Operations (GE NEB 0) is engaged in furnishing engineering services for domestic h
and foreign nuclear power plants.
ASSIGNED INSPECTOR: b. , 6* b'k R. L. Pettis, Special Projects Inspection Section Date (SPIS)
OTHER INSPECTOR (S): R. P. McIntyre, SPIS P. .
3 SMC O'Donnell n..
APPROVED BY: bAM (.i> Jn_ hr> fo-to-3 7 Date U. Potapovs, Chief. SP(S Vendor Inspecticn Branch INSPECTION BASES AND SCOPE:
A. BASES: 10 CFR Part 21 and 10 CFR Part 50.
B. SCOPE: The purpose of'this follow-up inspection was to review allegations involving potential deficiencies in design control activities within the ,
Qualit." Assurance (QA) program at GE San Jose, during the period March '
1978 to April 1982. In addition, the status of previous inspection findings was reviewed.
PLANT SITE APPLICABILITY: Potentially multiple plant sites, including River Bend, TVA Units 17-22 (identified by GE as cancelled), Perry 1/2, Nine Mile Point 2. Hope Creek 1/2, Grand Gulf 1/2, Linerick, Clinton, and Susquehanna 1/2.
41
ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY 8USINESS OPERATIONS SAN JOSE. CALI70RNIA REPORT INSPECTION
_ Hn - I RFSUITS- PAGF P of 17 A. VIOLATIONS:
None.
B. NONCONFORMANCES:
Contrary to Section 4.1.2.(a) of GE Engineering Operating Procedure (EOP) 55-2.00, " Engineering Change Control " Engineering Change Control (ECN)
NJ-17436, dated June 2, 1980, was initiated without a technical evaluation to: (1) delete the requirement for glyptal coating of GE Electrical Metallic Tubing (EMT) because of unavailability; and (2) delete the notation " Approved Vendor: GE Company, Requires Engineering Approval."
(87-01-01).
This item of nonconformance was previously identified in NRC Inspection Report No. 99900403/86-01 as unresolved item (86-01-10).
C. UNRESOLVED ITEMS:
None.
D. STATUS OF PREVIOUS INSPECTION FINDINGS:
- 1. (Closed) Nonconformance (86-01-01)
Contrary to Section 4.1.2.e of GE E0P 42-6.00, " Independent Design Verification," documentation was unavailable to support the processing of a Deferred Verification Status Notice (DVSCN), on ERM AML 973, identifying to Engineering Services 1 that Panel Module 0710 (assembly Drawing PL914E989), for the River 8end project, was cleared of deferred verification.
A further review of documentation conducted by GE and reviewed by the inspector indicated that independent design verification of Panel Nodule V710 had been deferred initially on Engineering Review Memorandum (ERM) AML 973, dated December 2, 1980. On May 13, 1982, ECN NJ 34893 was issued, which applied only to design changes made on that ECN and did not clear the deferred verification, as evidenced by the verification statement on the ECN. GE performed a search in February,1985 to identify all unverified documents on River Bend 1, which identified assembly drawing PL914E989 as unverified. As a '
result, DVSCN No. 02202, dated March 26, 1985, was written to clear the deferred verification on all these design documents including Panel Module Assembly U710, drawing PL514E989, Revision 4, as required per GE E0P-42-6.00. This item is closed.
42
4 1
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ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE. CALI 0RNIA REPORT INSPECTION ,
Nn . RFSUITS- PAGE 3 of 17
- 2. (Closed) Nonconformance (86-01-02)
Contrary to GE E0P 25-4.00, Sections 2 and 4.3, GE was unable to retrieve Engineering Work Authorization (EWA) No. FE-8446. This document defined the scope of work requested, per DVSCN No. C1706, for the qualification of electrical components to be used in nuclear safety-related Category 1 electrical circuits, for panel No. H13-P861.
EWA FE-8446 referenced on DVSCN 01706 was verified by the inspector to be an erroneous number. The actual document which authorized the qualification work for components associated with the panel is Project Work Authorization (PWA) 5447KG, dated April 26, 1985, which authorized the qualification of the same relays listed on DVSCN 01706 by part number. DVSCN 02409, dated April 25, 1985, cleared the deferred verification from DVSCN 01706. The inspector reviewed both DVSCNs which now include the reference to PWA 5447KG, as recorded on DVSCNs 01706 and 02409, dated January 21, 1987. This item is closed.
- 3. (Closed) Nonconformance (86-01-03) !
Contrary to 10 CFR Part 50, Appendix B, Criterion XV, and GE Corrective Action Request (CAR) Procedure MP 5.01 Section 4.2.3.2, a 4 GE manufacturing engineer dispositioned CAR No. SJ-5771125 days after the required reply date specified by the initiator, and a revised reply date was not negotiated.
A review of additional documentation for CAR No, SJ-57711 confirmed !
that the noted reply date was not in accordance with GE procedure i MP 5.01. A review of Rework Records RR004, RR011, RR022, and RR029 associated with this CAR indicated correction to the specific condition noted. This correction was accomplished on June 7, 1980.
The affected panel was accepted for shipment to TVA-X21 (cancelled ,
plant) on June 17, 1980. CAR SJ-57711 was signed off by the !
responsible design engineer on June 25, 1980. This delayed sign-off l had no effect on the quality of the product since the problem identified on the CAR was already resolved on Rework Record RR011.
A drawing change to the connection diagram was noted for other GE l plants utilizing the same nanel. GE's nuclear panel manufacturing l operations at San Jose have since been discontinued. This item is considered closed.
l l
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ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE. CALI 0RNIA REPORT IhSPECTION rn - RFSiff TS- par;F 4 nf 17
- 4. (Closed) Nonconformance (86-01-04)
Contrary to Section 4.1.1.e of GE Engineering Operating Procedure (E0P) 42-6.00, " Independent Design Verification," the required DVSCN, which would have documented the opening of a deferment per ECN NJ-43120, was not generated through GE's Engineering Services 1 department as required.
The NRC inspector reviewed additional information as provided in GE's response to this nonconformance. Although the required DVSCN necessary to close the deferred verification was not available during NRC inspection No. 99900403/86-01, it was produced by GE during this inspection. DVSCN 01010 was issued on January 25, 1984, as a result of an October 1983 internal GE Quality Assurance Audit.
This audit of deferred verification identified that DVSCNs had not been issued for some deferred verifications that were already documented on approved ECNs, ERMs, and conditional releases.
The GE documentation reviewed also identified that when items were released for manufacture or shipment prior to completion of equipment qualification, a conditional release was required in accordance with GE E0P 42-5.00, dated February 12, 1982. The approved conditional shipment release for Transmitter 18807360P017 was noted as being recorded on a Product Quality Certification (PQC) dated June 9,1983.
The inspector verified that part of the corrective action resulting from this audit was to revise E0P 42-6.00 (December 5,1983) to eliminate the redundant requirement for DVSCNs for items with incomplete equipment qualification. As a result of the 1983 audit, the necessary corrective and preventive action related to this nonconformance was accomplished. This item is closed.
- 5. (Closed) Nonconformance (86-01-05)
Contrary to Criterion V of 10 CFR Part 50, Appendix B and GE Nuclear Energy business Group (NEBG) Procedure No. 70-30, Revision 1, Paragraph 2, training records reviewed indicated Mr. Sam A. Milam, III attended a Quality Assurance (QA) course 15 months after assuming his new job assignment, rather than within the required 3 months.
A review of documentation supporting GE's response to this item indicated that attendance at a formal QA training course within three months of assuming a new job assignment is not required by GE
\
44
ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SA*1 JOSE. CALI :0RNI A ,
REPORT - INSPECTION tin - _ RESULTS: PAGE 5 of 17 Procedure 70-30. Rather, the procedure assigns managers the ~
responsibility to ensure that each employee is instructed ard indoctrinated in the procedural requirements applicable to the wo. -
assigned. Formal QA training courses were developed by GE to assis, managers in executing their responsibility for training or indoctrination of assigned employees; however, participation in these courses was not and is not mandatory.
According to Mr. Milam's manager at the time he assumed his new job assignment in 1978, Mr. Milam was initially assigned to a lead engineer who was responsible to instruct and indoctrinate him in the use of procedures and forms related to his assigned work. Design documents assigned to Mr. Milam had to be reviewed, signed, and dated by both the responsible design engineer and the independent I
design verifier, in addition to being reviewed by the responsible design engineer's manager. Following these reviews, the documents were submitted to Engineering Support Services (ESS) for processing and issuance in accordance with established procedures. Engineering Services' responsibility was to perform various checks on the quality of the documents it reviewed. As stated by GE, when any of these levels of review indicated errors or 6eviations from procedural requirements, this information would be fed back to Mr. Milam, either directly or through his manager. This iterative process of continuing on-the-job reviews and feedback ensured that Mr. Milam and other design engineers were adequately trained in the use of procedures and forms related to their assigned work. As a result, GE recognized this on-the-job training to be sufficient to instruct and indoctrinate design engineers in design documents such as drawings, ERMs, and ECNs. This item is closed.
- 6. (Closed) Nonconformance (86-01-06)
Contrary to Criterion III of 10 CFR Part 50, Appendix B and GE E0P 42-6.00, GE issued Engineering Review Memorandums (ERMs),
AMD-1302 and AMC-3035 in which: (1) the desi to perform the document review (ERM-AMD-1302)gner was identified; and (2) where the assigned reviewer also provided document approval as the responsible engineer (ERM-AMC-3035).
(1) A review of the QA procedures of GE and C.F. Braun (subcontractor to GE) and interviews with GE staff members determined that Braun was given full responsibility for the design of this panel utilizing Braun QA procedures. These procedures were (1) based ,
upon GE's QA program and (2) monitored by GE for conformance and implementation. The designer who also performed the 45
ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE. CALIFORNIA REPORT INSPECTION Nn - RFSULTS: PAGE 6 of 17 document review for ERM-AMD-1302 appeared initially to have reviewed his own work efforts, but, in fact, he reviewed comments that resulted from GE " spot check" and an " application review" of the ERM and associated drawing. The subject drawing was later approved for issue by Braun personnel on November 11, 1977, with all signature blocks signed and dated by Braun personnel. Subsequent to the completion of the design, independent verification, and checking by Braun personnel, the drawing was transmitted to GE for approval and issuance, which occurred on November 29 and December 30, 1977 respectively.
(2) On ERM AMC 3035, Mr. Milam provided review and comments and signed as the independent design verifier on July 5,1978.
Mr. Milam's comments were resolved by Mr. R. D. Thompson who signed as the responsible desiga engineer the same day. ERM AMC 3035 contains a comment made by a lead systems engineer, dated September 27, 1976; however, this comment had no effect on the previously completed design verification. This comment was resolved on October 4,1978, by Mr. Milam, who at the time was the responsible design engineer. This ERM was approved by Mr. Milam's manager on October 5,1970. GE explained that the 3 months between Mr. Milam's signing as the independent design verifier and the signing by the responsible design engineer was the result of a reassignment of Mr. Milam's job responsibility, which was not noted on the ERM itself.
Based on the inspectors review of actual signatures and dates recorded on the ERMs and on the drawings for items (1) and (2) above, the designers, verifiers, and the checkers were independent. This item is closed.
- 7. (0 pen) Unresolved Item (86-01-07)
GE Engineering Practices and Procedure (EP&P) 5.38 Addendum 4, dated December 1975, required that a tracking system and status log of deferred verifications be maintainea. The inspectors verified during the NRC 86-01 inspection that the first entry was made in the status log for deferred verifications in May 1977. At that time, it could not be determined whether verifications had been deferred before May 1977 since the status log did not contain of any deferred verifications prior to that date.
During the April 1987 inspection, GE provided additional documentation (several ERMs deferring verification from three separate work units within GE) that indicated that deferred verification activities were 46
ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS S_AN JOSE CALIFORNI A REPORT INSPECTION NO.- RESULTS: PAGE 7 of 17 initiated as early as November, 1975. Although documentation did not exist to support the inclusion of these documents into the status log, GE stated that they were incorporated but subsequently transferred to the Work Planning and Scheduling System (WPSS) for scheduling and tracking to completion. The only documentation produced by GE during the inspection, to verify completion status of these documents, was a computer run from the Engineering Informa-tion System (EIS), dated April 4,1987, which indicated that 272 documents presently exist in the system as "U" (unverified).
GE's position was that since the drawings referenced on the ERMs produced during the inspection were absent from this list, they must have been verified. It was also pointed out by GE that the WPSS scheduling records prior to 1980, which would have demonstrated the documents tracking status and eventual closure, have been eliminated from the system data base. As a result, documentation does not exist at this time to support the overall tracking status of these documents including the clearing of the referenced deferred verification.
GE committed to perform an extensive review of deferred verifications from inception through May 1977 to positively demonstrate closure of all design documents produced during this period which contain deferred design verifications. This unresolved item will remain open pending completion of this review and will be reviewed during the next inspection.
- 8. (Closed) Unresolved Item (86-01-08)
A determination as to the correct elementary drawing used as a basis to test Panel G41-P003 for TVA16 cocid not be made during the previous NRC inspection.
Additional information produced by GE was reviewed during this inspection to identify documentation used by GE to test Panel G41-P003 for TVA18. The Quality Assurance Test Data Sheet for this panel, dated December 30, 1978, identified connection diagram No.
865E859 and elementary diagram No. 386X994-034, to be used with test instruction TI 1200, as applicable. Specific documentation was unavailable to indicate which elementary diagram had been used to test the panel, because the elementary drawing was not a specifically referenced item on the test data sheet. The connection diagram contained information for both TVA and Grand Gulf panels; therefore, the inspector performing the test had to select the appropriate information. A review of the rework records for this panel did not 47
ORGANIZATION: GEhERAL ELECTRIC COMPANY MUCLEAR ENERGY BUSINESS OPERATIONS SAh JOSE, CALIFORNIA REPORT INSPECTION Nn - RESULTS: pAGE 8 of 17 identify any misapplication of the elementary diagrams. This item is closed.
- 9. (0 pen) Unresolved Item (86-01-09)
GE's Problem Review Board (PRB) stated in a letter Odted June 13, 1980, that a new reactor mode switch should replace those previously shipped and that the design changes required were to be documented via a Field Disposition Instruction (FDI) document. The FDI referenced by the PRB was not available for review during the previous NRC inspection.
GE's response to this unresolved item stated that the correct reactor mode switch is installed in all boiling water reactor (BWR) plants.
This installation was initiated as a replacement in BWRs in late 1983 and 1984, based on a GE product improvement redesign and the issuance of NRC Information Notice 83-42. Because this product improvement redesign is not the original issue that unresolved item (86-01-09) addressed, this item will remain open pending a further review of the 1980 upgrade data and the 1983 product improvement redesign. This unresolved item will remain open and be reviewed '
during the next inspection.
- 10. (Closed) Unresolved Item (86-01-10)
The resolution as to whether Electrical Metallic Tubing (EMT) incorporated glyptal coating on the inside diameter, as specified by GE, could not be determined during the previous inspection. This tubing is primarily used by GE in the construction of nuclear power plant control room panels to house related cables.
The inspectors reviewed GE's response to this unresolved item in a letter dated January 27, 1987. The response indicated that the 4 glyptal coating, as specified on GE purchase part drawing 175A9666 issued May 24, 1967, was not incorporated in EMT supplied to GE, nor does the absence of the coating present any safety problem since all conduit supplied and used by GE was of the seamless type with very smooth surfaces on both the inside and outside.
Peceipt inspection records reviewed during the last NRC inspection revealed that only the electro-galvanized external finish had been inspected and not the presence of glyptal. During these discussions, GE cculd not provide an explanation, nor could GE clearly explain the purpose for specifying such coating on the purchase part drawing.
However, discussions narrowed down the possibilities that glyptal 48
ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE. CALIFORNIA REPORT INSPECTION Nn - RESULTS: PAGE 9 of 17 had been specified to either aid the cable pulling operation through the EMT or to prevent corrosion of the inside of the tubing. As a result of the many questions surrounding the use of the glyptal coating and absent a complete and comprehensive technical justifica-tion for its removal in 1980, as required by GE E0P 55-2.00, this unresolved item is elevated to an item of nonconformance. (87-01-01)
- 11. (Closed) Unresolved Item (86-01-11)
Specific review requirements provided to the responsible engineer at the start of the project for panel compatibility with other system elements could not be determined during the previous NRC inspection for ERM-AMC 3035, dated October 11, 1987 for CNV Panel H22-P028.
When this ERM was issued in October 1978, no means was provided to indicate what information was available to the responsible engineer as to the design basis. Mr. C. W. Hart, engineering manager at the time, approved this document without a system interface review, as evidenced by " Review Refused" in the signature block. Mr. Hart i indicated that his review was inappropriate, because the responsible engineer, Mr. R. D. Thompson /Mr. Sam A. Milam, had the required information before the panel design. This particular ERM has misdated entries and out of sequence comments; however, the support documentation does not indicate hardware problems for this panel.
During this period other ERMs were examined and the interface reviews were performed with "no comment" entries. This item is considered closed.
E. OTHER FINDINGS AhD OBSERVATIONS:
Background Information As stated previously, NRC Inspection Report No. 99900403/86-01 did not attempt to address all of the items raised by the alleger and the consultant for GAP, Mr. Charles E. Stokes, but, rather, a representative sample of potentially more significant allegations was selected for review. However, all allegations received by the NRC are in the process of being addressed and will be documented in future inspection reports.
Previously, the area of deferred design verification was addressed, which represented the allegers major concerns (as noted during an NRC interview with the alleger in April 1986). This inspection report primarily focuses on the follow-up of items of nonconformance and unresolved items identified in NRC Inspection Report No. 99900403/86-01. This was 49
ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE. CALIFORNIA REPORT INSPECTION Nn -
RESULTS: PAGE 10 of 17 accomplished in part by a review of GE's responses to such items (GE's letters to the NRC dated February 5 and March 5, 1987) in addition to a formal review of all documentation supportive of the GE response.
In addition, several other allegations from Mr. Stokes' summary of Mr. Milam's work record were reviewed during this inspection.
The representative sample of allegations inspected is summarized below, along with the results of the NRC review of each item. The inspection was comprised of personnel interviews, examination of applicable files, records, and procedures. References to nuclear power plants identified as TVA (17-22) represent cancelled plants. The section references below have been restated verbatim and correlate directly to those listed in Mr. Stokes' repnrt; they do not constitute an NRC interpretation of such allegations.
- 1. Engineering Review Memorandums (ERMs)
Stokes Report Section 1.6 "In the first week of November 1978, the following line was part of an entry: Bill Millard said either he would sign the ERMs or I (Sam) could forge his signature to them."
(Clarification added by Mr. Stokes.)
Insaection Findings - During the inspection, a discussion was held wita Mr. Millaro, in the presence of Mr. Barton Smith, GE Counsel, at which time Mr. Millard denied any such statement about " forging" his signature. Mr. Millaro stated that it was common practice for Project Manegers (PMs) to authorize other individuals to sign for them in instances where, for example, logistics did not permit the PM easy access to sign such documents. The NRC inspector reviewed the responsibility given to the PN as outlined in GE E0P 42-6.10, Section 4.8a, to determine the significance placed on the PMs dpproVal of such documents. The PM has the responsibility to approve ERMs for the purpose of authorizing the application of the specific document to the assigned project and to supply project information to the initiating responsible engineer as requested.
The PM is not responsible for verifying or checking the technical adequacy of the document, because this has already been done by the design engineer and the verifier. Because specific detail was not available as to the ERM referenced by Mr. Milam's work record entry in November 1978, the inspector was unable to verify whether Mr. Milam signed his own name in lieu of Mr. Millard, or whether Mr. Milam " signed for" Mr. Millard. As a result of this review, this item will remain open.
50
ORGANIZATION: GENERAL ELECTRIC COMPANY NL' CLEAR ENERGY BUSINESS OPERATIONS SAN JOSE. CALIFORNIA REPORT INSPECTION tJn - RFStil TS - PAGF 11 nf 17
- 2. Elementary Diagram Drafting Effort Stokes Report Section 1.7 " Continuing with a problem of a similar nature on November 14, 1978, a letter to C. W. Hart on the subject of the CNV connection had an interesting paragraph. It seems that the CNV elementary diagram drafting effort was subcontracted to an outside firm, the Power Division of C.F. Braun & Company, in Alhambra, California. When completed, the diagrams were provided to the General Electric System Engineers for signature. The System Engineers felt that they were not being given sufficient time for review and refused to sign the documents. The documents were later signed by the C&EE CNV Engineer, without review."
Ins aection Findings - During the inspection, discussions were held witi Mr. C. W. Hart, Mr. Milam's supervisor during this period, who stated he had never received the November 14, 1978 letter. In addition, specific examples of insufficient review times could not be identified from the comments contained in the ERMs or the discussion with Mr. Hart. As a result, this item will remain open.
- 3. Clinton Interlocks Stokes Report Section 4.3 "On 6-10-81, Mr. Milam, when investigating an open item concerning the safety related status of Clinton recirculation flow control valve interlocks, determined that interlocks exist for feedwater pump trip with low reactor vessel water level and for high drywell pressure in the design specifications for the following referenced projects: Clinton, Zimmer, and LaSalle.
The interlocks definitely exist in elementary diagrams for Zimmer and LaSalle, but he could not find out positively that they exist for Clinton. Mr. Milam passed the information on to George Strambach, who was with the NRC, at the Pepper Tree Lounge." (Emphasis added by Mr. Stokes.)
Inspection Findings - The NRC inspector reviewed Clinton Reactor Recirculation System elementary Drawing No. 851E700 AC, Sheet 12, Revision 0, dated August 11, 1980, and Sheet 14, Revision 0, dated September 21, 1979, and verified the existence of the referenced interlocks. This item is closed. A discussion was held with Mr. George Strambach of GE concerning the statement regarding the Pepper Tree Lounge. Mr. Strambach stated that this referred to a technical meeting held between members of the NRC and GE staff for the purpose of discussing NRC Chapter 7 Safety Evaluation Reports 51
ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE. CALIFORNIA REPORT INSPECTION Nn - RESULTS: PAGF 19 of 17 for future GE plants. The meeting was held at the Peppertres Inn meeting room, with approximately 50 attendees, during the period June 8-10, 1980.
- 4. Grand Gulf Control System Issue Stokes Report Section 4.21 "In Mr. Milam's work record for FW8141, (week of October 5, 1981), it was stated that there had been another Grand Gulf Control system failure question." (Clarificationadded)
Insaection Findings - A review of the allegers work record associated wit 1 this item identified the following entries:
a/wk 8136 - Grand Gulf Control System Failure Analysis b/wk 8137 - Discussion of Failure Analysis for Grand Gulf with sign-off problems c/wk 8138 - Received more work sheets from Bechtel Engineer (Grand Gulf) d/wk 8141 - Responded to another Grand Gulf failure question The above references related to GE responses to control system failure questions in an NRC Letter dated April 16, 1981 (from l R. L. Tedesco, Assistant Director for Licensing (NRC) to Mr. J. P. McGaughy of Mississippi Power and Light). Mr. Milam's entries give a summary of his day-to-day work activities, which involved problem resolution, as assigned him by his supervisor.
This item is closed.
- 5. River Bend Excluded Equipment List Stokes Report Section 5.13 "Mr. Milam's work record includes a nonapproved form titled PWA No.1229LD, Revision IJ for River Bend.
This document, which is dated February 5,1982, was caused by an excluded equipment list which was sent to the utility, Gulf States Utilities Company, by the NRC. The second page of this document states that there is no controlled tracking system for vendor identification of these devices and that a complete item by item search of the entire River Bend database would be necessary. GE felt that the scope of such a search was prohibitive and furthermore was not considered to be necessary. Excluded equipment as referred to in this list is equipment which has been found at other facilities 52
ORGANIZATION: GENEPPL ELECTRIC COMPANY NUCLE R ENERGY BUSINESS OPERATIONS SAN JOSE, CALIFORNIA REPORT INSPECTION NO.: 99900403/87-01 RESULTS: PAGE 13 of 17 to be so deficient that plant safety is seriously in question. GE .
neither admitted nor denied that this equipment was installed at River Bend."
Inspection Findings - The NRC inspector reviewed a letter (RBV-1643) dated October 30, 1981, from the Stone & Webster Engineering Corporation (SWEC), agent for Gulf States Utilities, requesting that GE confirm that none of the listed equipment is being used by GE in any safety-related component on River Bend. This list of excluded equipment was generated by SWEC based upon a review of defective equipment identified by the NRC as a result of inspections of nuclear power plants, vendors, and published reports. A review of GE's February 22, 1982, response indicated that GE does not exclude equipment from use based upon lists of equipment found defective by the NRC. GE's response also indicated that the SWEC list did not contain information concerning the nature of the defects and environmental conditions necessary to evaluate the equipment's exclusion from use. Also, as individual problems are encountered, investigations are undertaken by the manufacturer and if necessary, the affected hardware is modified or replaced.
The GE reviewer, Mr. Sam Milam, had stated that no controlled tracking system exists within GE for vendor identification of these devices. This statement refers to the fact that to confirm the use of this equipment, a complete item-by'-item search of the River Bend data base would be necessary and would only apply to the control and instrumentation scope of equipment supply. As a result, GE stated in their February 1982 response to SWEC that at the present time a research program of this magnitude was not necessary. GE also indicated to the NRC inspector that SWEC did not provide additional correspondence after it received GE's response. This item is considered open pending NRC's review of a sample of GE's actions on affected hardware.
- 6. Unresolved NRC Items Stokes Report Section 5.14 "A letter dated January 29, 1962, and received on February 8,1982, informed everyone of the forthcoming .
NRC inspection on February 22 through February 26 and indicated the o NRC team assignments would be as follows: F 53 ,
ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE. CALI 0RNIA REPORT INSPECTION NO -
RFRtH TS- PAGF 14 of 17
" Chamberlain
- 1. Review and close out, if possible, the nonconformance reported in the previous inspection 81-03. This nonconformance relates to the material requests for safety-related components not being controlled by established procedures and instructions.
- 2. Follow-up regional request for investigations:
- a. Check allegations received by Region V that GE has furnished technical publications with technical, as well as typographical, errors. Subjects are control rod drives, control systems, and fuel handling platforms.
b.
Check allegations by Isa Yin of Region III, who accompanied LaSalle on LaSalle's 1981 audit, that there are problems with design report specifications. Yin felt that wrong revision numbers were being used.
- c. Region II, involving Hartsville and Phipps Bend, a TVA audit finding indicated that GE was not performing and reporting reportable conditions to TVA per 10 CFR 50.55(e).
- d. Inspect the situation report in which, on the Grand Gulf I project, GE furnished RPV level transmitters which cannot be adjusted to the specified setting.
" Foster
- 1. Inspection of series 20, type PR-20, electro switches (reported atGrandGulf).
- 2. Termination of wiring in termination cabinet jpnction boxes (PGCC reported by Grand Gulf).
- 3. Incompatible design documents with actual hardware ("as shipped" panels did not match "as shipped" drawing reported by Grand Gulf).
1
- 4. Inadecuate circuit separation (separation of Class IE and '
non-1E in accordance with Regulatory Guide 1.75. Reported by Grand Gulf).
- 5. Incorrect assembly of cable connectors.
u.Wim 54
ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE. CALIFORNIA REPORT INSPECTION Nn - RESULTS: PAGE 15 of 17
" Note: These five items were left unresolved since the 81-03 NRC audit, one year ago. The NRC apparently has its own deferred verification system. It is easy to see why GE started the Deferred Verification program when the NRC defers verification of noncon-l formances from inspection to inspection or longer."
l Inspection Findings - A review of NRC inspection reports indicated that the items raised by Mr. Stokes had been closed. Mr. Stokes' statement on the status of items assigned to Mr. Foster is incorrect.
Mr. Foster formally resolved items 2, 3, 4, and 5 during NRC Inspection No. 99900403/82-01 in February 1982, 5 months after the 81-03 inspection, although it appears from the report numbering sequence (81-03 vs. 82-01) that 1 year had elapsed between inspections.
Mr. Foster's remaining item (No.1) was the subject of a GE Quality Assurance action item, as stated in paragraph E.5.2 of NRC report 82-01. GE was to (1) verify that Series 20, type PR-20 electro-switches had not been shipped to other plants and (2) determine how the configured switch had been shipped initially to Grand Gulf, Unit 1. During this review, GE opened Potentially Reportable Condition (PRC) File No. 80-46 to examine switches for Clinton, Unit I and Grand Gulf, Units 1 and 2. NRC inspection 83-02 (May 1983) determined that another safety-related Series 20 electroswitch, GE Part No. 272A8005, may also be affected, however, a review of the PRC file at that time did not indicate that this part had been included in the evaluation.
NRC Report 83-02 also stated that Illinois Power Company filed a 10 CFR Part 21 report on April 17, 1983, with NRC Region III. The report indicated that deficient Series 20K electroswitches had been supplied as basic components in various panels furnished by GE for use at Clinton, Unit 1. The report also stated that GE will perform the revisions of any design documents affected. The NRC inspector reviewed the PRC file again during NRC inspection 84-01 (March 1984) and determined that the defective switches were checked and, replaced and the GE NEB 0 inventory cleared of all defective switches. This subsequent review of PRC 80-46 also considered the previously omitted switches in the evaluation.
With respect to Mr. Chamberlain, items 1, 2a, Eb, and 2c also were closed during NRC inspection 82-01 (February 1982), 5 months after they were identified in NRC inspection 81-03. Item Ed, which 55
ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE. CALI:0RNIA REPORT INSPECTION Hn - RFSUITS. PAGF 16 nf 17 resulted in an Item of Nonconformance during NRC inspection 81-03, was closed during NRC inspection 82-02 (June 1982), 4 months after it was identified. This item is closed.
- 7. Unauthorized Signature Changes Stokes Report Section 6.2 "Mr. Milam wrote a letter to W. M. Barrentine en April 14, 1982 about unauthorized, post signature changes. In this letter, Mr. Milam states that R. L. Reghitto made an authorized change to ERM AML-2997 without Mr. Milam's knowledge and in direct conflict with specific instructions."
Inspection Findings - A discussion during the inspection with Mr. Barrentine, in the presence of Mr. Barton Smith, GE counsel, inquired as to what actions were taken concerning this subject.
3 Mr. Barrentine stated he had not received Mr. Milam's letter of April 14,1982. He also stated that he was not aware of anyone else who might have known about the letter and also might have acted on it in his (Mr. Barrentine's) place while he was on business travel.
As a result, this issue will remain open.
- 8. Letter to Management Stokes Report Section 6.3 "On May 22, 1982, Mr. Milam wrote Mr. Barrentine a letter and included a copy of his work record while working for Mr. C. L. Cobler. In this letter, Mr. Milam requested Mr. Barrentine to read about the on-going underworld of C&ID and says he tried to communicate some of these things to Mr. Barrentine on severri occasions but was discouraged by Mr. Barrentine's managers and attitude. Mr. Milam says:
Since you no longer hold my form 38 (a standard threat), I have nothing further to fear from either you or your conspiratorial managers. I hope, by sending you this Rece-d. to give you a glimpse into that hidden world of uncontrolled bootleg activity we all know so well.
"Mr. Barrentine was the manager of the Nuclear Control & Instruments-tion Product Design Operation (NC&ID) or (C&ID). He was Mr. Hart's, Mr. Cobler's, Mr. Reghitto's, Mr. Strambach's, Mr. Koslow's, and Mr. Wortham's supervisor. Mr. Milam had been notified of his layoff when this last letter was written and his reference to form 38 had to do with the constant threat of layoff if you did not go along with the system. He did not." (Emphasis added by Mr. Stokes.)
1 56
ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE. CALIFORNIA REPORT INSPECTION NO +
RESULTS: PAGE 17 of 17 Inspection Findings - A discussion during the inspection with Mr. Barrentine, in the presence of Mr. Barton Smith, GE counsel, inquired as to what action was taken by Mr. Barrentine when he received Mr. Milam's letter and work record, which document problems Mr. Milam felt existed within the Control and Instrumentation Department (C&lD). Mr. Barrentine stated he never received the letter nor the portion of Mr. Milam's work record while he was assigned to Mr. C. L. Cobler. He did state that if he had received information concerning problems within C&ID, he would have met with the managers and throughly researched the issues. He. Barrentine also stated that a Potential Reportable Condition (PRCl evaluation of Mr. Milam's concerns would have been initiated if warranted. As
- a result, this item will remain open.
stED- s 57/58
ORGANIZATION: INLAND STEEL CDMPANY-EAST CHICAG0, INDIANA REPORT INSPECTION INSPECTION NO.: 99901081/87-01 DATES: 02/17-20/87 OH-SITE HntlR9- ??
CORRESPONDENCE ADDRESS: Inland Steel Company ]
3210 Watling Street ;
East Chicago, Indiana 46312' 0 ORGANIZATIONAL CONTACT: Richard L. Yancey TELEPHONE NUMBER: 219-392-4975 NUCLEAR INDUSTRY ACTIVITY: Steel producer, i
'l ASSIGNED INSPECTOR: w_, _
//z/87 J. C. Sirper, Progra'm Development and Reactive 'Date l InspectionSection(PDRIS) j 4
OTHERINSPECTOR(S)-
APPROVED BY: # .
J. $/. Stone, Chief, PDRIS, Vendor Inspection Branch IhSPECTION BASES AND SCOPE:
A. BASES: 10 CFR 50, Appendix B and 10 CFR Part 21, B. . SCOPE: The inspection was performed to follow-up on a November 1986 Virginia Electric and Power Company (VEPCO) Part 21 report concerning laminations in steel columns produced by Inland Steel Company (ISC) and supplied by Rockwell Engineering (RE) to North Anna for a-new service.
water spray array system.
PLANT SITE APPLICABILITY: North Anna Power Station. Units 1 and 2.
59
ORGANIZATION: INLAND STEEL COMPANY EAST CHJCAG0, INDIANA REPORT INSPECTION N0.: 99901081/87-01 RESULTS: PAGE-2 of 6 A. VIOLATIONS:
- 1. Contrary to Section 21.6 of 10 CFR Part 21, ISC failed to post copies 7 of Section 206 of the Atomic Energy Act. (87-01-01) I B. NONCONFORMANCES: j l 1. Contrary to Crii.erion V of Appendix B to 10 CFR Part 50, at the time I of the inspection, ISC did not have a procedure for performing sulfur rinting at the #2 Basic 0xygen Furnace (B0F) Bloom Caster.
87-01-02)
- 2. Contrary to Criterion II of Appendix B to 10 CFR Part 50, at the time of the inspection, ISC did not have a procedure or instructions pertaining to indoctrination and training of personne1' performing activities affecting quality. (87-01-03)
C. OTHER FINDINGS OR COMMENTS:
On November 7, 1986, VEPC0 notified the NRC that RE had supplied defective ASTM A36-84A wide flange structeral members for the new SWSAS supports.
On September 15, 1986, a ruutine post weld QC inspection initially revealed laminations in the web section of a steel column. RE was the seller of the steel columns to VEPC0 and ISC was the supplier of 10" x 49" wide steel flange shapes to RE. The original purchase order 29191 from RE was place with ISC on September 12, 1985 for Inland Wide Flange Shapes, ASTM A36-84A, W 10 in. x 49 in. x 35 ft. 3 in. On October 23, 1986, pre-installation ultrasonic testing (UT) of the steel from heat #17802 4 revealed lamination type indicatfans on several columns. RE was notified !
and an ISC UT-NDE Level II technician visited North Anna on October 28, '
1986 and identified the defect as " pipe type indications." According to the ISC Level II technician,18 out of the 25 beams ultrasonically tested .
were rejected for pipe type indications. All beams were tested end to i end. According to the ISC " Nondestructive Testing Personnel Qualification
& Certification Record," the ISC UT-NDE Level II technician was qualified according to the ISC " Training Qualification and Certification Procedure,"06-101, Revision 4, dated June 1981. ;
After review of the production records for the defective steel, ISC isolated the pipe condition to heat #17802 specification 52-05, AL killed steel, ingots #11 and #12. ISC attributes the pipe problem to abnormalities occurring during the pouring of the ingot. The '
abnormality was a full-running stopper which essentially affects the i control over ingot pour height (in this case, it involved a short pour of 85" in a 90" mold).
60
ORGANIZATION: INLAND STEEL COMPANY EAST CHICAGO, INDIANA REPORT INSPECTION ,
NO.: 99901081/87-01 RESULTS: PAGE 3 of 6 D. CONCLUSION:
Inland successfully isolated the pipe condition to heat #17802 (261 ingots were produced in that heat), open top ingots #11 and #12. Isolation of the problem was possible by narrowing down the ingot production size, steel making history, times and weight. The heat was produced at the 25*
structural mill and tapped on October 6, 1985. The pour time was normal and there were no stickers / scrappy ingots. Ingots #1-13 were charged in pit #6 at 6:05 pm on October 6, 1985 and drown on 3:37 am on October 7, 1985. There was apparently no recognizable causes for excessive pipe in the ingots (i.e., dished top contour, ingot eruptions, boiling, metal loss). However, ingots #11 and #12 did experience an irregularity since they were poured with full running stoppers (a short poured ingot of 85" in a 90" mold). It is probable that pour height control and/or top end turbulence as a result of the full running stopper condition led to excessive pipe. According to the "#2 B0F Metallurgical Report of Steel Making Inspection," the pour stream condition was solid, the metal action was quiet and the top condition was foam. Yields are calculated to be 81.5% and 82.8% for ingots #11 and #12 respectively.
Procedure #404-P-024, " Pouring Stand Set-Up Preparation Prior to Teem i Killed Open Tap Heat," #404-P-025 " Pouring Stand Mold Pouring Height Marking," #404-P-013. " Emergency Pouring," and #661-P-101., " Charging i and Drawing Ingots," were reviewed and found adequate. All applicable l procedures were followed during the production of heat #17802. ,
i In as much as pipe is an inherent quality of ingot solidification, the critical operation for good ingot quality is the cropping operation.
Therefore, in this particular case, the origin of the pipe defect is academic since it should have been spotted and eliminated during the cropping operation. Under the Metallurgical comments in the " Steel Application Memo" the instructions to the shearman were ingots "#11-12 =
cut out pipe." However, after the ingot was rolled and sent to the 28" shear mill, #2 bloomer for cropping, the shearman was required to
" shear to obtain the ordered Steel Application Memorandum (SAM) weight" according to the procedure #661-1-107, "Shearman/ Operating Shear." The ordered SAM weight for ingots #11 and #12 were one piece at 7260#, one piece at 5550# and one piece at 7460#, one piece at 5700#, respectively.
The order width and height for ingots #11 %d #12 were both 13.8750" and 12.00".
The shearman apparently followed the SAM and the procedure JQA No.
661-J-107 withcut either looking for or noting excessive pipe. Upon reviewing the #2MPR transaction log for the #2 blooming mill and the SAM, 61
ORGANIZATION: INLAND STEEL COMPANY EAST CHICAG0, INDIANA REPORT INSPECTION NO.: 99901081/87-01 RESULTS: PAGE 4 of 6 it was evident for heat #17802 that two groups of ingots, #2-5 and #6-11, the same amount of steel was cut for each group (each group has pieces with the same length). Ingots #2-5 all had 5" of steel cut from rolled ingots with an original 12' 4" length and 6" of steel cut from rolled ingots with an original 11' 11" length. Ingots #6-11 had 7" of steel cut from the rolled ingots with an original 13' 9" length and 6" from the rolled ingots with an original 10' 7" length. Ingot #12 had 6" of steel cut from the rolled ingots with an original 14' 2" length and 7" of steel from the rolled ingots with an original 10' 11" length. With this consistant pattern of the amount of steel sheared, it is probable that the shearman failed to note and eliminate some secondary pipe in the ingots. It is also possible that some pipe was interpreted as segregation and subsequently disregarded by the shearman. In many cases, pipe is not visible on the surface but can only be detected by ultrasonics or radiography.
Heat #17802 was not originally produced as a nuclear order for RE. The inspection revealed that RE did not specify the requirements of 10 CFR 50, Appendix B on the original purchase order (P0 29191) to ISC or the replacement order (R4-88962) but they did impose the requirements of 10 CFR Part 21. As a result of not imposing Appendix B on ISC, RE was not effective in communicating the quality requirements of the order.
At the time of the order (prior to July 1, 1984) ISC had a two tier !
quality system one for Nuclear Quality orders and one for commercial grade orders. l I
Presently, ISC treats all prime steel orders as quality system orders.
Therefore, at that time ICS justified filling the RE order from their commercial grade stock using their normal and routine (quality require-ments) procedures, which at the time were different from their equivalent Appendix B quality program (from a procedural standpoint). Nevertheless, since Part 21 was imposed on the order, ISC should have been alerted to the possibility that the structurals were to be put into nuclear related service.
As part of their corrective action, Inland has committed to inquire about end uses on all orders specifying 10 CFR Part 21 (in cases where 10 CFR 50, Appendix B is not specified as in this case). ISC plans to supply all orders specifying NCA 3800, N45.2-77,10 CFR 50 or 10 CFR 21 as bloom-cast (as a primary steel specification) and ingot-cast bottom-cut-only (as an alternative steel specification). UT inspection will be mandatory only if' ordered by the customer or in the specification. Bloom-cast and ingot-cast bottom-cut-only will reduce the probability of pipe or secondary pipe from being missed during ingot shearing since these production methods are inherently less prone to the pipe defect.
62
ORGANIZATION: INLAND STEEL COMPANY FAST CHICAGO, INDIANA REPORT INSPECTION l Nn - 44QninA1/R7-01 RESULTS: PAGE 5 of 6 l
1 10 CFR Part 21 Issues ;
As part of a plant tour of ISC, the NRC inspector verified that the i l
Part 206 was 21 notice not posted and see p(rocedures were adequately Violation 87-01-01). posted, however, Upon interviewing two Section supervisors in the finishing and shipping department it was evident that .
they had no knowledge of 10 CFR Part 21 or of a Part 21 procedure. There i was no objective evidence that employees involved with tasks affecting quality at ISC have knowledge of the requirements of 10 CFR Part 21.
Additionally, there is no objective evidence that 10 CFR Part 21 is addressed in any indoctrination and training of personnel performing '
activities affecting quality.
Training and Qualifications From the information provided by ISC, the NRC inspector concluded that a procedure for guidance on training and indoctrination of personnel involved in quality related activities did not exist. (See Nonconformance 87-01-03.) ISC submitted information to the NRC inspector which indicated that the quality training and indoctrination program consisted of crew !
meetings, audits on job performance and departmental training classes ,
(statistical process control (SPC) and bander marker training classes).
These types of training methods (except the audits) generally covered )
ISC's quality philosophy, future quality goals and covered in depth I training on applicable job procedures. No objective evidence existed I to support that these classes or any other classes formally incorporated and covered the requirements of 10 CFR Part 21, 10 CFR 50 Appendix B, or i hCA 3800 of N45.2-77 or the ISC quality systems manual.
A lack of training in these areas was evident from interviews conducted by the NRC inspector with two ISC supervisors. Neither supervisor was aware of 10 CFR Part 21. In addition, non-salary employees directly involved in quality activities (i.e., shearman) are not formally trained and indoctrinated on the ISC Quality Systems Manual or on 10 CFR Part 21.
The shearman is responsible for cropping out pipe from rolled ingots, therefore, eliminating laminations in the final product. Since the shearman is the last check point for pipe in the 28" mill, he is an essential part of the ISC ingot steel quality production program.
According to a January 28, 1985 Inter-Communication to all Department Heads from the Superintendent, Qualification Planning and Development,
.. 0ur intent is to train employees who would immediately use SPC."
63
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l l
l ORGANIZATION: INLAND STEEL COMPANY I EAST CHICAG0, INDIANA i REPORT INSPECTION !
I
_ Mn . ooon1nni/n7 n1 RFSul TS - PAGF 6 nf 6 l l
i The " Training Qualifications and Certification Procedure for Nondestructive I Testing Personnel,"06-101, Revision 4, June 1981 was reviewed and found to be adequate. The Nondestructive Testing Personnel Qualification and Certification record was also reviewed. Qualifications for an ultrasonics Level II inspector (the inspector that tested the structurals at VEPC0) and a Level III inspector ultrasonics were checked. Certifications, <
qualification tests, eye examinations and performance summaries for these '
inspectors were all up to date.
Plant Tour The No. 2 Blooming Mill and the No. 2 B0F/ Continuous Casting facility was toured. This tour included visits to the Basic 0xygen Furnace, the Ladle Metallurgy Station, the Strand Slab Caster, and the Strand Bloom / Slab combination caster. During the tour it was revealed that there was no procedure for the sulfur printing technique being performed at the #2 B0F Bloom caster. (See Nonconformance 87-01-02.) l l
Considerable time was spent in the shearman's pulpit observing shearing practices on shear slabs finished at #2 bloomer. Procedure 661-J-107 4 "Shearman/0perating Shear" was available at the work station and was i correctly implemented.
E. PERSONS CONTACTED:
Jeff 0'Barske, Quality Coordinator, Inland / East Chicago
- Frank A. Surgot, Staff Engineer, Inland / East Chicago
- Phillip A. Speer, Section Manager, Inland / Chicago
- James L. Federoff, Audit Coordinator, Q.S., Inland / East Chicago
- Richard L. Yancey, Audit Coordinator, Inland / East Chicago
- Cedric Oliver, Sr. SPC Engineer, Inland / East Chicago
- Rich Pruitt, Section Manager, Inland / Chicago A. J. Rudis, Jr., President, Rockwell Engineering Company R. A. Ruter, QA Administrator, Rockwell Engineering Company
- Attended the exit meeting.
64
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i ORGANIZATION: ITT BART0k INSTRUMENTS C0hPANY CITY OF Ih0VSTRY, CALIFORNIA INSPECTION INSPECTION 4 REPORT ON.. SITE H0tlRS. ?? l NO.: 99900113/87-01 DATES: OE/P4-27/87 I
CORRESPONDENCE ADDRESS:
ITT Barton Instruments Company ATTN: Mr. G. R. Welt, Director !
-Quality Assurance 900 South Turnbull Canyon Road City of Industry, California 91749-1882' ORGANIZATIONAL CONTACT: J. Dwyer ;
TELEPHONE NUMBER: 818-961-2547 NUCLEAR IhDUSTRY ACTIVITY: Manufactures pressure switches, pressure transmitters, and analog trip systems.
J T l
'l ASSIGNED INSPECTOR: C- '
/ Date A)
K. R. Naidu, Program Development and Reactive ,
Inspection Section (PDRIS)
OTHERINSPECTOR(S):
o/py APPROVED BY: [
J/ C. Stone, Ch'ief, PDRIS, Vendor Inspection Branch Dat6 INSPECTION LASES AND SCOPE:
A. BASES: 10 CFR Part 21 and Appendix B to 10 CFR 50. .
1 B. SCOPE: Review implementation of corrective action taken on noticon-formances identified in Inspection Report No. 99900113/02-04. Review actions taken to evaluate set point drift experienced at selected plants and review the design and manufacture of transmitters to ascertain similarities and differences when compared to the product of other manufacturers which have experienced set noint drift.
PLANT SITE APPLICABILITY: All plants using ITT Barton equipment.
65
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'0 ORGANIZATION: ITT BARTON INSTRUMENTS COMPANY CITY OF INDUSTRY, CALIFORNIA REPORT INSPECTION NO + QQQnn113/A7-01 RESULTS: PAGE 2 of 11 A. VIOLATIONS:
No violations were identified during this inspection.
B. NONCONFORMANCES:
No nonconformances were identified during this inspection. j C. UNRESOLVED ITEMS:
No unresolved items were identified during this inspection.
D.
ACTION TAKEN ON PREVIOUSLY IDENTIFIED ITEMS OF NONCONFORMANCE:
The inspector reviewed the corrective action taken by ITT Barton (Barton) on items of nonconformance identified during an inspection conducted during November 15-18, 1982 and documented in Inspection Report No.
99900113/82-04. Barton outlined corrective actions taken-to adoress the noncompliance in their letter dated March 4,1983 and provided additional information in their letters dated March 31 and April 11, 1983.
- 1. (Closed) Nonconformance 99900113/02-04-1 The nonconformance identified that the following were contrary to Criterion V of Appendix 8 to 10 CFR 50 and paragraphs 3.5.6, 4.1, and 5.7 of Barton Engineering Instruction (EI) No. 0764.1172.2, Revision 004, dated May 19, 1980:
- a. The noise level of the transmitter output had not beer, logged on the initial calibration sheet' .
- b. The zero output (initial or final) had not been recorded on the initial calibration sheet. ,
- c. Thermal effects had not been plotted on the data sheet.
EI No. 0764.1172.2 has since been revised several times. Current EI 0764-1300.2, dated July 23, 1966, has provisions to record the above.
The inspector reviewed Els 0752-1040.2, 0763-1300.2, and 0763A1149.2 for 752, 763, and 763A type transmitters and determined that they also have provisions to record the above data.
66 E ..
ORGANIZATION: ITT BARTCN INSTRUMENTS COMPANY CITY OF INDUSTRY, CALIFORNIA REPORT INSPECTION Mn - coonn11Un7_ni RESUITS: PAGE 3 of 11 ,
- 2. (Closed) Nonconformance 99900113/82-04-2 The nonconformance identified that contrary to Criterion V of Appendix B to 10 CFR 50, and paragraph 5.0 of procedure QU-11, Revision 02, dated June 1, 1982, the documented test results of the "ohmicity tests" identified in the Data Sheets of Test Procedure 0331.1027.2 had not been evaluated to assure that test requirements had been satisfied for strain gage beam assemblies identified in specific data sheets. The nonconformance also identified that QA had not reviewed the temperature compensation data sheets for !
j conformance to requirements prior to final acceptance for inprocess '
activity related to specific assemblies. The lack of review was evidenced by no indication of QC/QA stamp or initial in the appropriate block of the data sheets. The current data sheets have provisions to record values of the ohmicity test. The circuit board -i test data sheet is completed when the assembly is initially subjected {
to temperature cycles. QC is required to sign off on this sheet. j During final temperature compensation test, the entire unit (circuit i board and strain gege) is taken through the temperature cycle and ]
final adjustments to the zero and span are made. QC is required to -
sign off on the final calibration data sheet after verifying that the final temperature compensation test was performed.
- 3. [ Closed)Nanconformance 99900113/82-04-03 The nonconformance identified that contrary to Criterion V of Appendix B to 10-CFR 50, and the requirements contained in paragraphs 2.0 and 4.0 of Procedure QU-3, Revision 02, dated June 1, 1982, a review was not performed to assure that Barton drawings or specifications met the customer requirements as evidenced by the j supply of incorrect pressure transmitters for use at Palo Verde Nuclear Generating Station. Corrective action taken by Barton, is to utilize a design control checklist to route incoming orders. The !
contracts department reviews the order to ascertain whether the instrument selected meets the customer requirements and whether there are any special requirements. If there are any special requirements, the order is reviewed by application design engineering personnel. Subsequently the package is independently verified by perscnnel in marketing and cuclity engineering and finally sent to the contract administrator. The inspector reviewed the data packages for orders issued by Fitzpatrick and South Texas Project nuclear power stations and determined that the design control check 1'st was implemented.
67 s 1
ORGANIZATION: ITT BARTON INSTRUMENTS COMPANY CITY OF INDUSTRY, CALIFORNIA REPORT INSPECTION No - oconn119/n7-01 RESULTS: PAGE 4 of 11
- 4. (Closed) Honconformance 99900113/82-04-04 The nonconformance identified that contrary.to Criterion V of _
Appendix B to 10 CFR 50, Procedure QU-12 and QA Instruction No. QM ,
12-07, the. hydrostatic pressu~ gage mounted on the, test stand in Department' 019 had not been calibrated by its due cate of October 22, 1982. The inspector reviewed the calibration records and determined that'the hydrostatic test pressure gage was calibrated. To prevent l recurrence, QC generated a computer list of 4947 instruments with provisions to print out each measuring instrument and it's location (by area) which requires calibration during the following 30 calendar days. The cognizant area supervisor, in whose area the instrument is located, is vested with responsibility to get the' j
l instrument calibrated. QCperiodicallyfam.'itsthecalibration status.
~
- 5. (Closed) Noncompliance 99900113/82-04 '
)
The noncompliance identified that contrary to Criterion V of Appendix B to 10 CFR 50, and Procedure QU-5, the QA program did not i contain documented instructions and procedures to cover the use of the notice of deviation form that was used to document test equipment malf unctions or deviaticos noted<.durfag "inhouse" ,
qualification testing of model Noc. 763 an#764 pressure transmitters. p
Barton developed QA Instruction 1%03 titied Notice of Deviation (
Qualification Testing, to document deviations during tests to qualify instruments. Barton sta5ed in their letter dated March 4, 1983, that the QAI was 9-02. During a subsequent review, QC determined that it was appropriate for this procedure to be in Section 11 instead of Section 9.
- 6. (Closed) Nonconformance 99900113/82-04-06 , y The nonconformance identified that contrsry to Criterion-V of Appendix B to 10 CFR 50 and paragraph 4.2.2 of M. Instruction No. QAI 06-01, purchase orders ~(PO) issued for tei' ting services were not entered into the purchase order revietlog. Bartor/
dispensed with the log book. QA Instruction QAl 04-01, Revision F, dated February 26, 1986 requires all the originals of POV for nuclear power plant applications including calibration services to be stamped - Nuclear Contract "QA Engineering Review Requi ed." A PO
- is coded with a unique number to identify that the order is for ,?
nuclear application. When the respective material is received, " =
w
, }ye p9 68
-y
ORGANIZATION: ITT BARTON INSTRUMENTS COMPANY CITY OF INDUSTRY, CALIFORNIA REPORT INSPECTION NO - 44900111/A7-01 RESULTS: a PAGE 5 of 11 Receiving Inspection retrieves the P0, verifies from the inspection planner whether the referenced code relates to nuclear application.
If the P0 is for nuclear application, Receipt Inspection verWies whether the document was reviewed by QA Engineering. If the necessary reviews were not performed, the received material is required to be quarantined in the designated " Holding Area," pending verification from QA Engineering.
- 7. (Closed) Unresolved Item 99900113/82-04-01 The unresolved item identified that eve.though Engineering Instruction No. 0764.1172.2, Revision 004, dated May 19, 1980, required oata recording after 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> at various temperatures, there were no requirements to document the initiation and completion of temperature e/posure.
l The current procedure 0764-1253.2 dated July 23, 1986, contains I
provisions to record the initiation and completion of the respective temperature exposure during the temperature compensation tests. The temperature cycle is 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> duration at temperatures of 80 2, 130 2, and 90 2 degrees Farenheit. This provision to document the initiation and completion of temperature exposures have also been implemented for the model 763 and 752 transmitters.
E. INSPECTION FINDINGS AND OTHER COMMENTS:
- 1. Review of 10 CFR Part 21 Report on Model 580 Pressure Switches Barton notified the NRC on April 14, 1986 of two potential defects in their Model SE'l series pressure indicating switches.
a., Specifically, under loss-of-coolant accident conditions, switch malfunctions occurred when the test chamber temperature was raised to 340 F. The symptoms of the malfunctions were:
- 1. In three of' the five instruments tested, one or both of the switches failed to operate (no change in switch state) when input pressure was varied.
- 2. Switch setpoint shifted in excess of the allowable 10 percent on two of the instruments.
69
ORGANIZATION: ITT BARTON INSTRUMENTS COMPANY CITY OF INDUSTRY, CALIFORNIA REPORT INSPECTION Nn_- 99900113/87-01 RESULTS; PAGE 6 of 11 When the chamber temperature was lowered to room ambient, all of the switches changed state when the input pressure was varied. However, the switch setpoint shifts which occurred at the elevated temperature was permanent. Barton identified that the specific component used in these applications was a Honeywell snap-acting switch, part number 11SM 403. Barton considered this defect generic based on the premise of Honeywell's certificate of conformance that parts, materials i
and processes used to manufacture these switches were similar to those previously supplied to Barton.
NRC issued Information Notice IN 86-65 alerting owners of nuclear power plants of this problem.
- b. On June 16, 1986, Barton informed the NRC that in addition to the above mentioned problems, Barton determined that there is a deflection of the instrument case due to an increase in test chamber pressure. The deflection affected the position of the switch actuating mechanism and may result in either switch set point drif t in excess of 15 percent or no operation of the switch. In this notification, Barton informed the NRC that the problems would be resolved by April 15, 1987.
The inspector discussed the progress of the resolution of the above problems with cognizant engineering personnel and determined that Barton procured a different type of microswitch and subjected the switches to stabilizing . temperature exposure prior to use.
This thermal stabilization process provided stress relief to the switch and it's internal components. The redesigned case, with components, is scheduled for qualification testing from mid June through December, 1987. Aging tests are performed in-house. Radiation, seismic and LOCA tests are done by outside testing laboratories.
2.
Review of 10 CFR Part 21 Report on Model 580A Pressure Switches On October 11, 1984, Barton notified the NRC of a potential defect in their model 580A, 581A and 583A differential pressure switches.
Specifically, in these types of switches, the setpoint shift exceeded the specification requirements when the instruments were operated at temperatures above 225 F. Barton determined l
l 70
ORGANIZATION: ITT BARTON INSTRUMENTS COMPANY CITY OF INDUSTRY, CALIFORNIA REPORT INSPECTION Nn - 494nn119 /n7-01 RESULTS: PAGE 7 of 11 that the setpoing drift was the result of relieving of stresses in a switch plate assembly which is a subassembly of the complete instrument. Use of stress relieved switchplate assemblies eliminated the setpoint drift. Barton notified the owners of nuclear power' plants where the switches were required to perform satisfactorily when the temperature exceeded 225 F that Barton would provide stress relieved switchplate assemblies with installation instructions.
- 3. Review of 10 CFR Part 21 Report on Model 763 Pressure Transmitters On June 30, 1983, Barton notified the NRC of a potential defect in suppressed zero model 763 static pressure transmitters. Specifically, the transmitters exhibited a negative shift in the output during initial exposure to operating pressure. The amount of shift was a function of process pressure and calibrated span of the instrument.
In interim reports dated Au;ust 16, 1983; October 14, 1983; November 16, 1983; January 16, 1984; February 7, 1984; and March 12, 1984, Barton informed the NRC of progress in the identification of the defect, and corrective action taken to resolve the problem.
Initially, Barton suspected that the cause of setpoint drift in zero suppressed model 763 transmitters was the attachment material of the link wire connecting the pressure sensing Bourdon tube with the strain gage beam. Subsequently, Barton determined that the drift was caused by the creep of the link wire connecting the pressure sensor to the strain gage beam. A design change was mad.e to the method of mounting the strain gage beam to control the deflection. The location of the insulating pad, where the strain gage lead wires were terminated, was also changed. The modified design eliminated the setpoint drif t. Test results of the modified sensor assemblies and completed transmitters indicate that the instruments perform to the published specifications. Barton provided the above information to their customers.
- 4. Review of 10 CFR Part 21 Report on Model 763 and 764 1 transmitters On October 29, 1982, Barton informed the NRC of a potential defect in model 763 and 764 transmitters. The potential defect e).hibited itself in the form of nonrepeatability and resulted in performance outside Barton's published specification. Specifically, at elevated temperatures above 130 F the error was measured to be more that the guaranteed 1.3%. At 420'F, the error exceeded 10%. The defect affected both zero based and suppressed zero based instruments.
71/72
f ORGANIZATION: ITT BARTON INSTRUMENTS COMPANY CITY OF INDUSTRY, CALIFORNIA REPORT INSPECTION wn . 000nn119/p7.n1 RESULTS- PACF P nf 11 Barton determined that the thermal non-repeatability was caused by the following:
- a. There was electrical leakage through the wiper arms and the shafts of the potentiometer (zero and spa'n) to the instrument-case.
- b. Incorrect temperature compensation calibration methodology was used in-house to calibrate the transmitters.
Barton corrected the above problems by 1 implementing the following: ]
- a. Installing isolation washers between the potentiometer mounting bracket and the potentiometer to isolate the potentiometer from case ground.
- b. The temperature compensation calibration procedure was revised.
- 5. Evaluation of Problems Identified in Licensee Reports (LER) !
The sequence coding and search system (SCSS) data base contains events reported by various nuclear power plants which are documented in LERs. A search of the SCSS yielded 497 LERs related to Barton pressure switches and transmitters during the period 1980 to mid 1986. The abstracts of the LERs were reviewed and the problems identified were discussed with Barton engineering personnel to ascertain whether there were any generic deficiencies with certain instrument models. Results of the discussions relative to eacn instrument model are documented in suceeding paragraphs.
- a. Model 368 Differential Pressure Transmitter (DPT) q l
I This DPT model was primarily supplied to General Electric Company (GE) for installation in Boiling Water Reactors (BWR). One instance of setpoint dr.ift was identified in 1980 in a BWR which has been operating since the early sixties.
Barton engineers were unaware of any major design problems with this transmitter model. GE routinely purchased a variety of instruments without stating the specific application. This problem appears to be isolated. The manufacture of this DPT has been discontinued and the DPT has been deleted from current I catalogues.
1 I
73 { l 1
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0FiGANIZATION: ITT BARTON INSTRUMENTS COMPANY CITY OF INDUSTRY, CALIFORNIA REPORT INSPECTION NO - QQQn011 '1/ A */-01 RESUITS: PAGE 9 of 11
- b. Model 386 DPT This DFT model was primarily supplied to Westinghouse Electric Corporation (Westinghouse) for installation in Pressurized Water Reactors (PWR). In this model, the electronic components are designed to withstand higher temperatures. Occasional setpoint drift problems have been reported. Barton engineers were unaware of any major design problems. Westinghouse routinely purchased a variety of instruments without stating the specific application.
Earton discontinued the manufacture of this DPT and deleted it from the current catalogues.
i
- c. Vcdel 384 DPT The model 384 DPT is similer in construction to the 368 and 386 models. Occassional setpoint drift problems were identified.
Barton engineers stated that it is possible that epoxy bonding of the piezo strain gage to the Kovar beam would yield under stress with elapse of time. Currently, Barton utilizes molecular bonding for attaching strain gages to the Kovar beam.
One LER stated that the license replaced the strain gage, and i recalibrates the transmitter. It is Barton's position that j instruments with repetitive setpoint drift are to be replaced.
The same corrective action is also applicable for models 368 l
and 386. Barton discontinued the model 384 and replaced it !
with model 752. Westinghouse is the primary user of model 752.
l d. Liquid Level Measuring Sensors 1 1
Model 352 and 353 liquid level sensors are equipped with pressure sensing devices at one end which transmit the sensed pressure >
through liquid filled capillary tubes to a remotely located pressure transmitter. The models 352 and 353 are qualified to IEEE-323-74 and IEEE-344-75 standards for use inside containment.
Some drift problems were identified in PWRs in applications to measure the reactor cavity sump level. Barton engineers stated '
that problems will arise if the recommended instructions are not followed during the initial installation to hard vacuum fill the capillary tubes. Since these instruments were '
purchased by the NSSS manufacturer and installed under their supervision, they were not informed of the problems.
74
ORGANIZATION: ITT BARTON INSTRUMENTS COMPANY CITY OF INDUSTRY, CALIFORNIA REPORT INSPECTION Nn . coQnn111/A7-01 RESULTS: PAGE 10 of 11
- e. Differential Pres.sure Ind1,ca_tingg, Switches Models 278,_2,88,_288A,
~
289, 289A, and 290A Numerous LERs from different NPPs reported various problems relative to inaccurate alarms, setpoint drift, and switch malfunctions. Barton engineers were unaware of any of the problems because the instruments were purchased by NSSS manufacturers. As such, when problems arise, the hPP personnel contact the NSSS manufacturers for service and spare parts.
The identified prcblems were discussed with Barton engineers and the causes evaluated. Evaluations indicate three types of reoccuring problems; namely, inaccurate alarm indication, switch failures, and setpoint drifts. Barcon suspects that the j contacts of the snap action switches were damaged by either high "
inrush currents when used to energize AC relay coils or by high inductive currents when used to energize DC relay coils. During the replacement of the snap action switches, the actuator arms have to be removed. The actuator arms are reconnected after the replacement and realigned for proper operation. Improper tools and techniques used during this process may result in recurring problems. Barton compiled a switch training course manual which discusses the probable causes, the recommended solution and illustrations of circuit diagrams which will not damage the switch contacts.
l Barton offers a ccmplete switch plate assembly to replace the ones used in the 228, 288A, 28; and 289A models. The new switch plate assembly permits the replacement of the snap action switch without removal of the actuator arms. In the original switch plate assemblies, the actuator arms had to be removed and realigned during the replacement of the snap action switches. 1 Barton provides the ITT Barton switch training course manual !
along with the replacement kit.
F. EXIT INTERVIEW:
The inspector met with Barton representatives mentioned in Section G .
at the conclusion of the inspection and discussed the scope of the '
inspection and findings.
l 75
. . . ......j
ORGANIZATION: ITT BART0f INSTRUMENTS COMPANY j CITY OF INDUSTRY, CALIFORNIA REPORT INSPECTION Nn - 00 Grin 113/R7 01 RESULTS: PAGE 11 of 11 l
G. PERSONS CONTACTED: I
- G. R. Welt, Director, Quality Assurance C. Watson, Manager,. Development Engineering
- V. Lawford, Senior Mechnaical Engineer A. Preiser, Senior Mechnaical Engineer K. J. McLean, Contractor Administrator D. Bell, Supervisor, Manufacturing )
J. Dwyer, Manager, Quality Control D. A. . Price, Director of Marketing J. P. Doyon, Manager, Sales and Service P. Elderton, Service Manager l
- Denotes persons who attended the exit meeting on February 27, 1987.
7b
ORGANIZATION: PAUL MONR0E HYDRAULICS, INCORPORATED ORANGE, CALIFORNIA REPORT. INSPECTION INSPECTION NO.: 99900337/87-01 DATES: 3/31-4/1/87 ON-SITE HOURS: 16 CORRESPONDENCE ADDRESS: Paul Monroe Hydraulics, Incorporated ATTN: F. Erlach President 1701 West Sequoia Avenue Orange, California 92668 ORGAt:IZATIONAL CONTACT: R. C. Fisher TELEPHONE NUMBER: 714-978-9600 NUCLEAR INDUSTRY ACTIVITY: Hydraulic snubbers and hydraulic actuators for valves. Refurbishment of actuators manufactured by others.
A ASSIGNED INSPECTOR: -/ A f[~//#
K. R. Na'du', Pro 9 ram Development and Reactive ~0 ate Inspection Section (PDRIS)
OTHERINSPECTOR(S):
APPROVED BY: M ~ 6ae7f7 J. C. f tone, Chief, P0lil~S','Ve'ndor InspecTio'n7r'anc~h INSPECTION BASES AND SCOPE:
A. BASES: 10 CFR Part El and Appendix B to IC CFR 50.
B. SCOPE: Review corrective action taken relative to a 10 CFR Part 21 report dated August 30, 1984, review selected records, and witness I tests on pressure switches installed in hydraulic valve actuators '
l which leaked oil at the South Texas Project.
]
FLANT SITE APPLICABILITY: South Texas Project Units 1 & 2 (50-498; 499).
77
ORGANIZATION: PAUL h0NR0E HYDRAULICS, INCORPORATED ORANGE,' CALIFORNIA REPORT INSPECTION NA -
QQQnn197/R7-01 RFSULTS; PArJ 9 nf 6 A. VIOLA _TIONS:
No violations were identified during this inspection.
B. NONCONFORMANCES:
Ne nonconformances were identified during this inspection.
C. UNRESOLVED ITEMS:
No unresolved items were identified during this inspection.
D. ,1N,SPECTION N FINDINGS AND OTHER COMMENTS: l
- 1. Background Informa_t_ ion Paul-Monroe Hydraulics (PMH) manufactures hydraulic snubbers and hydraulic valve actuators to operate valves manufactured by others.
PMH also services hydraulic valve actuators manufactured by others who have since discontinued production. PMH provided hydraulic actuators to operate valves manufactured by Control Components, Incorporated. These valves are used as power opwated relief valves at the South Texas Project Nuclear Pcwer Plant.
- 2. Review of 10 CFR Part 21 Report On August 30, 1984 PMH informed the NRC Region.I office of potential problems in hydraulic valve actuators installed in the Limerick Nuclear Power Plant. PMH supplied similar actuators to Byron 1 & 2, Braidwood 1 & 2 Limerick 2, and WPPS 1 & 4.
Specifically, the hydraulic pump, that maintains hydraulic pressure in an accumulator that supplies fluid to operate the valve actuator, cycled more frequently than specified in the PMH operation manual.
PMH stated that frequent cycling to recharge the accumulator does not prevent the system from performing its intended function.
However,'it reduced compenent life and caused the system to operate at elevated temperatures which could, at some point, cause automatic shutdown of the unit. PMH determined that the frequent cyclin ,
occurred because of pressure leakage from three components a) gthe solenoid operated directional control valve, b) the piston seals in '
the valve actuator and c) the gas precharge in the accumulator.. PMH established corrective action to prevent recurrence as follows:
78
ORGANIZATION: PAUL MONROE HYDRAULICS, INCORPORATED ORANGE, CALIFORNIA REPORT INSPECTION Nn . QQQnn197/n7_n1 RFSU!TS- PAGF 3 nf 6
- a. The solenoid operated directional control valve, manufactured by Kean Controls Corporation, leaked because the solenoid seating surfaces were scored during fabrication or from previous fluid contamination. The suspect valves were replaced.
Other solenoid valves were monitored and repaired. PMH currently performs source inspection on each valve.
- b. Hydraulic fluid was found to be leaking past the piston seal of the Flotork actuator. PMH re-bronzed the pistons'and rehoned the cylinder barrel walls. To prevent recurrence, PfiH assigned a PMH unique part number in the bill of material which requires Flotork to rebronze the piston and the cylinder walls in actuators ordered by PMH.
- c. Four accumulators were observed to have porosity on the inner surface. PMH replaced the defective accumulators with tested accumulators. To prevent recurrence, each accuniulator is pressure tested prior to acceptance.
No problems have been identified subsequent to this corrective actior.
- 3. Review of the 10 CFR Part 50.55(e_) Condition )
l On March 23, 1987, Houston Light and Power Company, the owners of the South Texas Project Nuclear Power Plant (STP) reported to the NRC that during hot functional testing of Unit I, the pressure switches installed in the hydraulic actuators (HA) manufactured by PMH leaked oil. The HAs operate valves manufactured ;
by Control Components Incorporated (CCI) and as a unit function as j power operated relief valve (PORV). In this case, CCI is the I contractor, PMH is the subcontractor to CCI. Each HA is equipped with three pressure switches (PS) manufactured by-ITT NeoDyn (NeoDyn).
One PS monitors the nitrogen pressure in the hydraulic oil accumulator. The second PS monitors the oil pressure and starts the hydraulic oil pump motor when the accumulator pressure drops to a specified value. The third PS stops the hydraulic oil pump motor when the accumulator pressure reaches a predetermined value. Six of the eight oil PSs installed on the four HAs leaked oil. The four nitrogen pressure switches did not leak. The leaking oil PSs were replaced with the ones intended for Unit 2. STP ' returned three of the six failed PSs switches to PMH for tests and evaluation. The inspector reviewed the pertinent documents at PMH located in Orange, CA and observed tests performed on the returned PSs at NeoDyn located in Chatsworth, CA.
79 ,
l 7
ORGANIZATION: PAUL MONROE HYDRAULICS, INCORPORATED i
ORANGE, CALIFORhIA
)
l REPORT INSPECTION NO . 0Q0009 97 /517.01 RFRill TR ? par,F a M A
- a. Review of Chronology of_ Events l
The inspector reviewed the chronological sunrnary of communica-tions between STP and PMH and determined the following:
- 1) On February 28, 1987, the HAs performed satisfactorily as documented.in the PMH service report.
- 2) On March 5, 1987, STP initiated Nonconformance Report SJ-03470 dated March 5, 1987, documenting that a: total four PSs installed in four actuators leaked. The PSs were replaced, tested and determined acceptable.
- 3) On March 11, 1987, two additional PS leaked.
l i
- 4) March 16, 1987, PMH to STP inquired whether the PORV HAs were exercised as required by PMH maintenance manual during the storage period of years in the warehouse.
- 5) March 23,1987, telephone notes from STP to PMH indicated that a Nonconformance Report was initiated to document that the HAs were not exercised according to the manufacturer's reconar. ended maintenance instructions during storage in the warehouse.
C) On March 25, 1987, three of six PSs were shipped to PMH along with two failed servoamplifiers..
- 4. Observation of Tests at I,lT_NeoDyndC_hatswort_h)
On April 1, 1987, NeoDyn performed tests on 3 PSs returned from STP, identified as AIMS-PSH-7411, AIMS-PSL-7411 and CIMS-PSL-7431.
All three switches were type 100 P15780 and have a date code 84-09 indicating that they were manufactured in September 1984. In this type of PS, a teflon coated polymide diaphragm senses the process pressure and transmits it to an actuating mechanism. The diaphragm isolates the process from the motion transfer drive and is held in place between the pressure port, 0-ring assembly and the pressure plate. Hence, perforation of the diaphragm will permit oil to enter the motion transfer device and exit through the electrical conduit box.
! 80 l -- -
ORGANIZATION: PAUL MONR0E HYDRAULICS, INCORPORATED ORANCE, CALIFORNIA REPORT- INSPECTION Nn . QQonn197/n7.01 RFSill TS - PAGF R nf 6 During the test, all three pressure switches leaked when subjected to air pressure forcing remnant cil through the motion transfer device mechanism. The pressure switches were disassembled and the diaphragms were examined. Black particles were observed on two of the three diaphragms. The preliminary conclusion was that particles in the hydraulic oil deposited on the diaphragm and may have caused perforations. NeoDyn stated that the particles on the diaphragm and samples of the oil which was expelled from the switch during air pressure tests would be sent for chemical analysis.
The body and pressure plate of each switch would be further examined. The damaged diaphragms were replaced in the three PSs and the PSs were retested with air pressure. The PSs held 3000 psi without leaks and were verified to actuate to the high and low setpoints.
- 5. Review of HA Records 's The inspector reviewed the available documentr.tior at FhH on the HAs supplied to CCI. The documentation was in the form of final data packages (DP) for eight HAs with serial numbers PF 89270-500-01 through 08. The DPs indicate that HAs were tested at PMH during November 2 through December 0, 1984. Each DP contained the following:
- a. Certificate of Ccnformance certifying that the HA is in accord-ance with the CCI purchase order and applicable requirements of PMh quslity assurance program and that the HA has been qualified to the applicable requirements specified in specification 652/800/1, Revision O.
- b. An Acceptance Test Procedure for Linear Valve Modulating Operator" checklist was used to document the test results.
A quality control representative witnessed the test.s as evidenced by his signoffs. .
- c. Certificate of Conformance (CoC) from United Technologies, Springfield, Massachusetts dated November 21, 1984, certified that the accumulators conformed to the applicable design specifications. The CoC is dated several days after the HAs were tested. PMH clarified this discrepancy by producing a Nonconforming Material Report dated September 14, 1504, which identified that the CoCs did not accompany the accumulators. The CcCs were received after the tests.
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ORGANIZATION: PAUL MONR0E HYDRAULICS, If;CORPORATED ORANGE, CALIFORiilA REPORT INSPECTION L'n - ooonn197/97_n1 RFSl!! TS - PAGE 6 of 6
- d. The CoC from FORNACIARI/ PARKER dated November 27, 1984 certified that the cylinders conformed to the applicable design specifications. PMH stated that the CoCs received with the cylinders on September 24, 1984 were deficient and had to be returned. The revised CoCs were dated November 27, 1904.
- e. PMH received 27 pressure switches (PS) from NeoDyn on i September 17, 1984. During receipt inspection, the receipt inspector observed that the CoCs did not accompany the PSs. Inspection ISS No. 0047 dated September 24, 1984 documented this discrepancy and was closed on October 1, 1984 after the receipt of CoCs. Review of the documents irdicates that there was no apparent problem with the operators in general and that specifically the PSs were operable.
E. EXIT INTERVIEW: 4 The inspector met with individuals identified in Section F and discussed the scope and findings of the inspection.
I F. PERS0hS CONTACTED:
Paul Monroe Hydrau,lics, Incorpora,ted, Orange, California ,
F. Erlach, President
- R. Fischer, Chief Engineer J. Raymont, Director of Operations T. Cottrill, Quality Assurance Manager All the above individuals attended the exit meeting on March 31, 1987.
ITT heoDyn, Chatswort_h,, Ca,lifornia ,,
- J. F. Dregne, Regional Sales Manager
- L. Amper, Quality Control Manager
- L. A. Dunham, Engineering Manager
- S. R. Nanda, Director of Quality Assurance
- Denotes individual who attended the exit meeting on April 1, 1987.
82
- _ - _ _ _ _ _ _ _ _ _ _ - _-___ __--_________- . -_ L
ORGANIZATION: ROCKWELL ENGINEERING COMPANY, INCORPORATED BLUE ISLAND, ILLIN0IS REPORT INSPECTION INSPECTION NO.: 99900836/87-01 DATES: 02/19/87 ON-SITF H0llPO A CORRESPONDENCE ADDRESS: Rockwell Engineering Company, Incorporated ATTN: Mr. A. J. Rudis , Jr.
President 13500 South kestern Avenue Blue Island, Illinois 60406 ORGANIZATIONAL CONTACT: Mr. A. J. Rudis , Jr.
TELEPHONE NUMBER: 312-385-4600 NUCLEAR INDUSTRY ACTIVITY: Naterial Supplier - Structural Steel Components.
-i ASSIGNED INSPECTOR: _-_ -
' u 9
J. C.i Harper, Program Development and Reactive Date Inspection Section (PDRIS)
OTHERINSPECTOR(S):
APPROVED BY: ,. ,y [, [/Joj7 J.[.' Stone, Chief,PDRIS,VendorInspectionBranch Unt(
INSPECTION BASES AND SCOPE:
A. BASES: 10 CFR 50, Appendix B and 10 CFR Part 21.
B. SCOPE: The inspection was performed to follow up on a November 1986 Virginia Electric and Power Company (VEPC0) Part 21 report concerning laminations in steel columns produced by Inland Steel and supplied by Rockwell Engineering (RE) to North Anna for a new service water spray array system.
PLANT SITE APPLICABILITY: North Anna Power Station Unit Nos. 1 and 2.
l i
83
ORGANIZATION: ROCKWELL ENGINEERING COMPANY, INCORPORATED BLUE ISLAND, ILLIN0IS REPORT INSPECTION Nn - 00nn0A1r/A7-01 PFSt!t Ts- pAGE 7 of 6
.A. VIOLATIONS:
- 1. Contrary to Section 21.21 of 10 CFR Part 21, RE failed to adopt a Part 21 procedure for evaluating and reporting deviations.
(87-01-01)
- 2. Contrary to Section 21.31 of 10 CFR Part 21, RE failed to pass down the requirements. of Part 21 to sub-tier vendors involved in fulfilling the VEPC0 PO-ET 16183 for the new service water spray array systen supports. (87-01-02)
- 3. Contrary to Section 21.6 of 10 CFR Part 21, RE failed to post copies of Section 206 and a procedure for evaluating and reporting deviations pursuant to Part 21. (87-01-03)
B. NONCONFORMANCES:
- 1. Contrary to Criterion IV of Appendix B to 10 CFR Part 50 and the RE 0A Manual, Revision 9, RE did not pass down the requirements of 10 CFR 50, Appendix B to sub-tier vendors involved in fulfilling the VEPCO PO-ET 16183 for the new service water spray array system supports. (87-01-04)
- 2. Contrary to Criterion IX of Appendix B to 10 CFR Part 50 and the RE procedure " Qualification, Training and Testing of Quality Assurance Inspector," RECO 1900-1, Revision 3, and the recommended practices of SNT-TC-1A, two OA Inspectors with NDE responsibilities were certified without record of a vision examination. There was no objective evidence that a decertification examination was taken by one of the two CA Inspectors. (87-01-05)
- 3. Contrary to Criterion XV of Appendix B to 10 CFR Part 50 and to the PE OA Manual. Revision 9, at the time of the inspection RE had no procedures to address segregation of nonconforming materials.
(87-01-06)
D. OTHER FINDINGS OR COMMENTS:
Background
On November 7,1986, VEPC0 nntified the NRC that RE had supplied defective ASTM A36-84A wide flance structural members for the new service water spray array system supports. On September 15, 1986, a routine post weld 84
ORGANIZATION: ROCKWELL ENGINEERING COMPANY, INCORPORATED BLUE ISLANi), ILLIN0IS
= . .
REPORT INSPECTION NO - QQQnnR16/R7-01 RESULTS: PAGE 3 of 6 QC inspection initially revealed laminations in the web section of a steel column. RE was the seller of the steel columns to VEPC0 and Inland
( Steel was the supplier of the steel flange shapes to RE. On October 23, 1986, pre-installation ultrasonic testing (UT) of the steel from suspect J.x heat #17802 revealed several additional defective columns. RE was 10 notified and an Inland Steel UT-NDE Level II technician visited North
, Anna on October 28, 1986 and identified the defect as " pipe type indications."
h After review of production records for the defective steel, Inland isolated the pipe condition to heat #17802, ingots #11 and #12. Inland's position was that the excessive pipe in ingots #11 cnd #12 was probably an indirect result of an abnormality occurring during the pouring of the ingot. This abnormality was a full running stopper which essentially affects control over ingot pour height.
Conclusions An inspection of Inland by the NRC inspector confirmed that Inland has an adequate quality program and procedures in place for ingot processing including pouring heats, rolling ingots, ingot pit soaking, and cropping operations. Inland successfully isolated the pipe condition to heat 17802, ingots #11 and #12 by narrowing down the ingot production size, times, and weight. According to the Inland #17802 Heat Review, there was no recognizable cause for pipe on the ingots (i.e., dished top contcur, ingot eruptions, boiling, metal loss). It is conceivable that pour height control or top end turbulence may increase the probability of excessive pipe, however, in this case, the critical interface for ingot quality is the cropping operation. Therefore, the l origin of the pipe defect is academic since it should have been i eliminated during the cropping operation. The Inland Steel Application I Memo instructed the shearman to " cut out pipe" without mandating a reduced yield (for the circumstance a normal request). Once the ingot was rolled and sent to the shears for cropping, the shearman was not required to be conservative in shearing the ingot (since he was not required to clear the fish tail of the ingot). In light of these factors, it is probable that the shearman missed some pipe or secondary pipe in the ingot. It is also possible that missed pipe may have been interpreted as segregation. In many cases pipe is not visible on the surface but can only be detected by ultrasonics or radiography.
4 As part of their corrective action, Inland has committed to inquire about end uses on all orders specifying 10 CFR Part 21 (in cases where 10 CFR 50, Appendix B is not specified as in this case), and to supply all
' 85 1 l
ORGANIZATION: ROCKWELL ENGINEERING COMPANY, INCORPORATED BLUE ISLAND, ILLIN0IS REPORT INSPECTION Nn - 400nnn16/A7 M REStiLTS - PAGE 4 of 6 l orders specifying NCA 3800, N45.2-77, 10 CFR.50 or 10 CFR 21 as bloom- ,
cast (as a primary steel specification) and ingot-cast bottom-cut-only j (as an alternate steel specification). UT inspection will be mandatory only if ordered by the customer or in the specification. In this case, l UT inspection may have been the only method of discovering the pipe i in the finished product. Bloom-cast and ingot-cast bottom-cut-only j will reduce the probability of piDe or secondary pipe from being missed 4 during ingot shearing to almost nil since these production methods are f inherently less prone to the pipe defect. l It was determined from the inspection that RE did not specify the requirements of 10 CFR 50, Appendix B on the original purchase order (P0 29191) or the replacement order (RY-88962) but they did impose the requirements of 10 CFR Part 21 on Inland. As a result of not imposing Appendix B on Inland, RE was not effective in communicating the quality 1 requirements of the order. Therefore, Inland could justify filling the RE order in their normal and routine (quality requirements) manner, which i at the time was different from their equivalent Appendix B ouality program (from a procedural standpoint). Nevertheless, since Part 21 was imposed on the order, Inland should have been alerted to the possibility that the structurals were to be put into nuclear safety related service.
10 CFR Part 21 Issues ,
As part of a plant tour of RE, the NRC inspector, verified that the Part 21 notice was adequately posted, however, neither Section 206 nor a prncedure for Part 21 was posted. (See Violations 87-01-03.)
1 Upon review of the RE Quality System documentation, it was apparent that a procedure for evaluating and reporting devaitions did not exist. Therefore, at the time of the inspection there was no '
formalized method of evaluating potentially reportable Part 21 items or reporting O rt 21 items. (See Violation 87-01-01.)
RE fabricated the W10 x 49 members by using processes such as cutting, boring, welding and galvanizing before shipping to VEPCO. VEPC0 l imposed the requirements of 10 CFR Part 21 and 10 CFR 50, Appendix B as part of the original P0 ET-16183 on RE. RE subcontracted the galvanizing to the Empire Galvanizing (P0 29230). RE did not impose Part 21 or Appendix B on Empire (see Violation 87-01-02 and Noncoformance 87-01-04). Empire was on the RE approved vendors list when the members were galvanized. Welding rods, Lot #14832, type E711-1, were used for
- welding on the job. The rods were purchased from Weld Star (P0 29682),
l 86 i i
l ORGANIZATION: ROCKWELL ENGINEERING COMPAtlY, INCORPORATED BLUE ISLAND, ILLIN0IS REPORT INSPECTION Nn - QQQnn916/R7 01 RFSULTS: PAGE 5 of 6 Job No. 8624-STK. RE did not impose Part 21 or Appendix B on Weld Star j (see Violation 87-01-02 and Nonconformance 87-01-04). Weld Star was on ,
the list of RE approved vendors at the time that the rods were purchased. !
In addition, RE purchased beams and columns from the Central Steel and
' Wire Company via Northwestern Steel and Wire Company (P0 NSW-36583, dated l' November 5, 1985) to be applied to the VEPC0 order. Northwestern Steel and Wire was on the RE approved vendors list when the order was placed. However, RE did not impose the requirements of Part 21 or Appendix B on its P0 to Northwestern Steel and Wire Company. (See Violation 87-01-02 and Nonconformance 87-01-04.)
Qualification and Training Review of the RE is procedure " Qualification, Training and Testing of ,
Quality Assurance Inspector," RECO 1900-1, Revision 3 revealed that it )
was adequate. The qualifications of two 0A inspectors, two welders (#32 l and #24) and one QA auditor / certified welding inspector were reviewed.
Both of the QA inspectors were missino annual eye exams for at least one year and one of the QA inspector's certification had expired since there was no objective evidence of a decertification examination on file. (See Nonconformance 87-01-05.) The QA auditor / certified welding inspector's qualifications were well documented and correctly presented.
I Welders #24 and #32 welded the structural members for the VEPC0 order.
They both were correctly qualified according to AWS D1.1 (flat and horizontal positions) and to procedure 500-3, Revision 0, " Welding Procedure Specification for Flux Cored ARC Welding of Low Alloy and Carbon Steel Plates, Shapes and Forms." VEPC0 approved the procedure on September 24, 1986.
Plant Tour The NRC inspector inspected the shear work station, the saw work station, the burner area, the receiving inspection area, and the weld storage room. Up-to-date procedures were available at the shear and saw work stations as well as the burner area and receiving inspection area. The weld storage room was clean and all electrodes used for nuclear service had certifications and were properly stored.
Weekly calibration of the REC 0 2/2A 25 ft, tape measures were verified by the calibration status sheet.
87 l
1
ORGANIZATION: ROCKWELL ENGINEERING COMPANY, INCORPORATED BLUE ISLAND, ILLIN0IS REPORT INSPECTION NO .
QQQOOA1A/R7-01 RF9UITS: PAGE 6 of 6
.E. PERSONS CONTACTED:
A. J. Rudis, Jr., President, Pockwell J. Merich, Plant Superintendent, Rockwell R. A. Ruter, QA Manager, Rockwell i
j 88 i
I l
ORGANIZATION: SANbVIK SPECIAL METALS CORPORATION KENNEWICK, WASHINGTON l
REPORT INSPECTION INSPECTION NO.: 99900764/87-01 DATES: 03/16-19/87 ON-SITE HOURSt 27 CORRESPONDENCE ADDRESS: Sandvik Special Metals Corporation ATTN: Mr. E. R. Astley President Post Office Box 6027 Kennewick, Washington 99336 ORGANIZATIONAL CONTACT: Mr. K. Bowles, QA Manager TELEPHONE NUMBER: (509) EP6-4131 NUCLEAR INDUSTRY ACTIVITY: Nuclear fuel cladding supplier for Babcock &
Wilcox, Combustion Engineering, and Advanced Nuclear Fuel Corporation (ANFC) (formerly Exxon Nuclear Company).
ASSIGNED INSPECTOR: - -
Mn M # 87 R.L.Cilimberg,PrygramDevelopmentandReactive Date Inspection Sectioh (PDRIS)
OTHERINSPECTOR(S): None.
APPROVED BY: bhV8A\ f N]
. C. Stone, C [T)hp RIS, Vendor Inspection Branch ate INSPECTION BASES AND SCOPE:
A. BASES: 10 CFR Part 21, 10 CFR Part 50.
B. SCOPE: Observe fuel cladding fabrication activities, review Quality '
Assurance (QA) program implementation and follow-up Sandvik corrective action on previous NRC inspection findings.
i 4
PLANT SITE APPLICABILITY: Nuclear power facilities using CE, B&W, and ANFC fuel.
89 l
ORGANIZATION: SANDVIK SPECIAL METALS CORPORATION KENNEWICK, WASHINGTON REPORT NO_.
INSPECTION Q00n0764/A7-01 RESULTS: PAGF 2 of 6 A. VIOLATIONS:
None.
B. _NONC0_NFORMANCES:
1.
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Sections 3.4.1 and 3.5.2 of QA-GA-5, Revision 14, dated February 7,1986, the inside diameters (ID) of 19 hollows from inspected with the wrong go-no go gauge.(87-01-01) ingot number 224190Q were 2.
Contrary to Criterion V of Appendix B to 10 CFR 50 and Section 2.1.1.8 of Z121, " Alkaline Cleaning," Revision 10, dated June 3, 1985, the west end of Tank #3 did not exhibit the required identification concerning the filling of Tank #3 with fresh cleaning solution.
(87-01-02)
C. UNRESOLVED ITEMS:
None.
D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
1.
.(Closed) Violation (85-01-01)
Contrary to Section 21.31 of 10 CFR Part 21, SSM ordered zirconium alloy tube shells without specifying the applicability of 10 CFR Part 21 in the procurement documents for these purchased n:aterials.
The NRC inspector's review of SSM procurement documents representing orders to the suppliers of tube shells determined that the applicability of 10 CFR 21 has been specified by SSM since the last NRC inspection.
This item is considered closed.
2.
1 Closed) Nonconfonnance (85-01-01)
Contrary to Criterion V of Appendix B to 10 CFR Part 50 and Section 6.1 of the SSM QAM, current copies of some procedures were not available at work stations and expired copies of other procedures had not been removed from work stations.
The NRC inspector determined that current copies of procedures are available at work stations and expired copies of procedures have been removed. This item is considered closed. j 90
ORGANIZATION: SANDVIK SPECIAL METALS CORPORATION KENNEWICK, WASHINGTON REPORT INSPECTION NO - QQQnO76d/R7-01 RESUITS- PAGF 1 nf 6 E. INSPECTION FINDIAGS AND OTHER COMMENTS:
- 1. Entrance and Exit Meetings SSM management representatives were inforn,ed of the scope of the inspection during the entrance meeting. The inspection findings dnd observations Were summarized during the exit meeting on March 19, 1987.
- 2. _ Procurement The NRC inspector examined SSM procurement documents representing orders to suppliers of tube shells and hollows which are the starting material for the fabrication cf fuel cladding. This examination revealed that the requirements of 10 CFR 21 have been imposed on suppliers of shells and hollows since the last NRC inspection.
Review of SSM-1030, " Tube Reduced Extrusions (TREX) - Zirconium Alloy," Revision 9, dated February 16, 1983, confirmed the quality requirements for hollows that were imposed on suppliers.
The NRC inspector determined that the documentation required by SSM-1030 for several oroers was in compliance with SSM-1030, Revision 9. The suppliers used for these purchases were approved in accordance with SSM procedures.
- 3. Inspection ,of_ Starting Meterial The NRC inspector observed the inspection of 19 pieces of hollows from ingot Number 224190Q using Quality Assurance Procedure QA-GA-5, Revision 14. lhis procedure requires that hollows be inspected at attribute limits specified on form QC 053. QC 053 requires that the 10 e/ hollows be inspected with a go-no go gauge which measures 1.625 and 1.655 inches. The SSM inspector was using a go-no go gauge which measures 1.615 and 1.665 inches.
The SSM inspector stopped using the wrong gauge and began to use the specified gauge when questioned by the NRC inspector.
Nonconformance 87-01-01 was identified in this area.
- 4. Alkaline Cleaning The review of process specification Z121, " Alkaline Cleaning,"
Revision 10, dated June 3,1985, established that the date that Tank #3 was filled with fresh cleaning solution shall be written on 91
ORGANIZATION: SANDVIK SPECIAL METALS CORPORATION KENNEWICK, WASHINGTON REPORT INSPECTION No_- 94400764/87-01 RESULTS: PAGE 4 of 6 a piece of maskirig tape and placed on the west end of the tank. The NRC inspector observed that the west end of Tank #3 was devoid of tape or any other forth of the required information. The SSM operator and supervisor placed a tape with the required information on the tank when questioned by the NRC inspector.
Nonconformance 87-01-CE was identified in this area.
- 5. Fuel Clad _di,ng Fabrication and Inspection The NRC inspector observed 11 steps in the fabrication and inspection of zirconium alloy tt.bing for fuel cladding cind compared the proteaural requirements to the actions being performed by the operators. The following procedures and manufacturing steps were observed:
(a) Two reduction steps were performed according to Z181, " Breakdown Rocking," Revision 6, dated September 15, 1986 (b) Deep etching of hollows was performed according to Z140,
" Etching of Zirconium Hollows," Revision 17, duted February 12, 1987.
(c) Etching of tubes was performed according to Z411, "Preanneal Etching of Tubes," Revision 17, dated February 11, 1987.
(d) Three vacuum annealing steps were performed according to ZE30,
" Intermediate Hollows - Annealing," Revision 16, datea November 17, 1986.
(e) End chamfering of hollows was performed according to Z160, "Large Hollows - End Chamfering," Revision 7, dated October 11, 1982.
(f) Visual inspection of tubes in lot ABJ 22 was performed dCCording to NDT-V-50, "First Visual," Revision 7, dated March 8,1984.
(g) Ultrasonic test (UT) equipment was prepared for operation to detect flaws in tubing according to NDT-UT-10 " Setup of UT Flaw Inspection Systems for Inspection of Thin Wall Metal Tubing," Revision 7, dated November 23, 1983.
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92
ORGANIZATION: SANDVIK SPECIAL METALS CORPORATION KENNEWICK, WASHINGTON REPORT INSPECTION No - QQQOn764/R7 01 RESULTS PAGF 5 of 6 j I
(h) UT equipment was prepared for operation to measure dimensions of tubing according to NDT-3D-31, " Setup of the SSM Tubing Dimensional (3D) Test Stations," Revision 7, dated January 29, ,
1987. l l
The operators are qualified to work at a number of stations and the NRC inspector observed sevt.ral different operatcrs at each work station. The operators followed the procedures which were current end at the work stations in compliance with the SSM.QAM. No items of nonconformance or unresolved items were identified.
- 6. Internal Audits Internal audit reports for 1985 and 1986 were reviewed to determine conformance with Section 18 of the SSM QAM, Revision 8. Internal audits are performed in accordance with QA-GA-22, " Internal Audit Procedure," Revision 7, dateo November 3,1986, which implements all sections of the SSM QAM, Revision 8. Audit reports contain a review by the Quality Control Engineer of tube fabrication compared to an audit checklist generated by a word processor which outlines the requirements iri the QAM, procedures, process specifications, and instructions for each of 16 production and inspection areas in the plant. Each production and inspection area is audited every six months at the. rate of two areas per month.
The NRC inspector determined that the 1985 and 1986 internal audit reports are in compliance with the SSM QAM, Revision 8 and QA-GA-22, Revision 7. Corrective action (CA) and follow-up audits to verify CA were performed as required by the QAM and QA-GA-22 Revision 7.
1 93
i ORGANIZATION: SANDVIK SPECIAL METALS CORPORATION KENNEWICK, WASHINGTON l
REPORT INSPECTION I Nn_* QQ Qnr,764 / A 7-01 RESULTS: PAGF 6 of 6 -l i
F. PERSONS CONTACTED:
- K. Bowles
- D. Darsow J. Davis K. Garretson l C. Grando !
B. Howard D. Knutzen K. Langston T. Lee R. Logan 1
- J. Luebke l D. fiorwood I
- K. Redmann C. Stacey e
- J. Schemel l
- Attended exit rneeting.
1 l
94'
ORGANIZATION: SORRENTO ELECTRONICS SAN DIEGO, CALIFORNIA i
I REPORT INSPECTION INSPECTION DATES: March 16-20. 1987 ON-SITF Hnftp9 En NO.: 99900387/87-01 CORRESPONDENCE ADDRESS: Sorrento Electronics .
Division of GA Technologies Incorporated !
ATTN: Mr. I. Bijarchi President and General Manager P. O. Box 269025 San Diego, California 92116 ORGANIZATIONAL CONTACT: Mr. M. L. Jones - QA Director TELEPHONE NUMBER: (61!D 457-RR33 )
NUCLEAR INDUSTRY ACTIVITY: Sorrento Electronics' current activities are .
the design, engineering, and manufacturing of radiological monitoring systems.
ASSIGNED INSPECTOR: k C//MG J. J. Fetrosino, Program Development and Reactive Date I Inspection Section (PDRIS)
OTHERINSPECTOR(S): F. Victor, Brookhaven National Laboratory APPROVED BY:
Jamps C. Stone, Chief, PDRIS, Vendor Inspection Branch h/G/h Date ;
i INSPECTION BASES AND SCOPE:
A. BASES: Appendix B to 10 CFR Part 50 and 10 CFR Part 21.
B. SCOPE: This inspection was conducted to: (1) Verify QA program implementation, and (2) review a potentially reportable 10 CFR 21 issue that concerned an engineered safety feature actuation system deviation at the Palo Verde Nuclear Generating Station (PVNGS).
PLANT SITE APPLICABILITY: Hope Creek (50-354); Palo Verde (50-528/529/530);
andSouthTexas(50-498/499).
95
ORGANIZATION: S0RRENTO ELECTRONICS SAN DIEGO, CALIFORNIA ,
1 j
REPORT INSPECTION
{
)
un - coonnAn7/n7.01 R F(;til TS . PAGF 2 of B i
A. VIOLATIONS: j
- 1. Contrary to Section 21.21, " Notification of Failure to Comply or ,
Existence of a Defect," of 10 CFR Part 21, Sorrento Electronics (SE) has not adopted procedures that adequately assure that 10 CFR Part 21-potentially reportable deviations are evaluated and dispositioned as required. As a result,-SE failed to perform its required evaluation of deviations that were discovered on April 29, 1986 in the-engineered safety features actuation system (ESFAS) panels that were' fabricated by SE and supplied to the PVNGS facility (87-01-01).
B. NONCONFORMANCES:
- 1. Contrary to Criterion V, " Instructions, Procedures, and Drawings "
of Appendix B to 10 CFR Part 50, and SE's QA procedure number 9-01, f
" Verification of Special Process Controls," three out of the four 1 sub-tier vendor welder qualification records that were reviewed did not indicate that the required SE auditor verification activities hadbeenperformedasrequired(87-01-02). ;
f , .
- 2. Contrary to Criterion V, " Instructions, Proceduresi.and Drawings "- 1 of Appendix B to 10 CFR Part 50, and Section 5 of the SE Quality _l Assurance Manual (QAM), " Instructions, Procedures, and Drawings,"
SE failed to provide approved instructions to assembly personnel that were preparing coaxial cable ends and installir,g on the coaxial ,l l cable MHV type electrical connectors for use in SE's RD53 radiation !
monitoring systems detectors (87-01-03).
C. UNRESOLVED ITEMS:
During the next NRC inspection a review will be performed to evalute SE's determination of the root cause of the issues discussed in Section E.3 and E.6 of this report. (87-01-04/05)
D. STATUS OF PREVIOUS INSPECTION FINDINGS:
During this inspection we did not review any SE corrective actions that were implemented in response to the previous NRC inspection report Number 99900387/84-01. The SE corrective action will be reviewed during a future NRC inspection. 1 96 r
WM dV 1.
ORGANIZATION: SORRENTO ELECTRONICS / w~
SAN DIEG0, CALIFORNIA o
~
REPORT INSPECTION ,
wn - ooonn M7/A7 n1 RFSUITS; ,
PAGE 3 cf R E. OTHER FINDINGS OR COMMENTS: 3
- 1. _1_0 CFR Part 21 The SE 10 CFR Part 21 precedure was reviewed to determine its adequacy to provide compliance with the 10 CFR Part 21 evaluation and reporting requirements. In addition, the implementation of the. 1 5 10 CFR Part 21.6 posting requirements were eviiuated by inspecting '
the manufacturing areas at SE for contif ance. Discussionstere conducted to determine how SE assures that potentially repu table deviations are evaluated. It was determined that SEiharceaplied with its 10 CFR Part 21 posting requirements, and the SE procedure provides adequate reporting and notification clarityl J:wcVer, two aspects of the SE 10 CFR Part 21 procedure did not asw f. hat the m intent of our regulations were implemented, o,
^
(a) One violation was identified bewuse SE failed to assure that its '
potentially reportable deviatiors are evaluated (see Vloistion 87-01-01). This was evidenced.by one example where 5E failed ._
to perform its required evaluatice of an April 29, 1986 '
y engineered safety features actuation system (ESFM) ' component deviation at PVNGS.
(b) The second area of concern was that the 10 CFR Part 21 procedure' adapted by SE was open to an incorreef interpretation by SE a employees. The procedure could be interpreted such that before I an SE employee could report a deviation to supervision a l " substantial safety-hazard" evaluation would have to be performed.
This potential interpretation could discourage the SE workers from reporting deviations to supervisor for their review. The SE management recognizes that a weakness exists in this area n '
and has stated it was not the intent of SE to place any such ' ,
restrictions on its employees.
Sorrento Electronics had previously discussed the possibility of revising their 10 CFR Part 21 procedures. Since.the NRC inspector identified the interpretation and clarity concerns, SE has committed to review and revise their 10 CFR Part 21 procedure and posting notice.
- 2. QA Program Implementation -
2 Seversi inspections were performed in the SE production floor areas to evaluate the QA program implementation. However, during this i
97
m
. ~
ORGANIZATION: SCRRENTO ELECTRONICS
. SAN DIEGO, CALIFORNIA n
REPORT '
INSPECTION wn . oconn1n7/gg.01 RESUITS: PAGE 4 of 8 inspection, few SE safety-related work activities were being perfotned; therefore, it could not be determined how well SE was
~
( impiduenting its QA program at this time. Currently, SE is establishing a new production manufacturing facility in Mexico.
Nevertheitss, the following areas and processes were inspected eventhcugh the work activity was low:
(a) Coaxial cable cnd preparation activities were observed to evaluate if qualitative and quantitative criteria were being adequately controlled during in-process work activities.
Adequate controls were in place and were complied with in this area except for one aspect. Nonconformance 87-01-03 was identified in regard to not providing approved instructions for installation of an MHV electrical connector. It was observed that a worker who was preparing coaxial cable ends for electrical connector installations was not using any written instructions or technical sheets for cable end preparation. When asked for the instructions being used for this operation, the worker provided the NRC inspector with a vendor technical paper that indicated it was not to be used for fabrication. Therefore, at this work station, SE had not provided approved instructions.
(b) Audits and Surveys conducted by SE were reviewed. A three year audit schedule has been established and implemented for outside i audits. Additionally, the internal audit schedule was reviewed, and found to be in compliance with the SE QAM. However, during this review Nonconformance 87-01-02 was identified. 9-01, l paragraph 5.3.1, states, "When supplier audits are performed I per contract to assure control of special processes, the auditor shall verify and document that qualified personnel, '
equipment, and procedures are used." On those occasions when welding procediires are verified by an auditor who is not a qualified welding inspector, the welder qualification sheets are initially reviewed and signed by a welding engineer or qualified welding inspector and then verified and signed by the auditor. In three of four such records examined there was no documentation (i.e., countersignature) to substantiate that the qualification sheets had been reviewed by a qualified welding inspector, or welding engineer.
(c) Special processes - conformal coating and wave soldering areas.
Conformal coating is applied in accordance with SE Manufacturing Process Specification (MPS) 00007013. Several circuit boards were in the process of being coated and were witnessed by the 1
98 !
ORGANIZATION: SORRENTO ELECTRONICS l-SAN DIEGO, CALIFORNIA l REPORT INSPECTION en - ooonn'4A7/n7 01 RESULTS: __
PAGE 5 of 8 NRC inspector. Procedures were available, and in use in the work place. Acceptance criteria are listed in MPS 00007013 '
and are checked by Quality Control inspectors in accordance with Quality Control Instruction 207. Final inspection results are recorded on the final inspection sheets for each circuit -
board. In addition to the final inspection, periodic surveillance are conducted by Quality Control Inspectors on a random basis. Four of the most recent surveillance records -
were reviewed. Qualification records for one employee who was "
operating the conformal coating machine, and his supervisor, were examined and found to be current. #
Another special process, wave soldering, was also observed.
This operation was conducted by one operator, and assisted by his supervisor. Both the operator and his supervisor had ,
current qualification documentation. The wave soldering was accomplished in accordance with MPS 00007019. Procedures were
! present in the work place and in use. The operator exhibited l an adequate knowledge and understanding of the procedure. No p nonconformances were noted in special process operations.
I (d) Nonconforming parts, material and components - Nonconforming material is handled in accordance with Quality Control i Instruction 509. When nonconforming material reports (NMR) are filled out they become part of the routing package for each subassembly. Each NMR is listed on the Master Routing List (traveler) for resolution before a particular point in the manufacturing sequence. Fabrication of the subassembly cannot proceed until the NMR is satisfactorily dispositioned or a -
management decision is made to carry the NMR on to the next sequence.
Each action is documented on the traveler. Nonconforming material is tagged and segregated into locked bins until properly dispositioned. No nonconformances were observed in handling nonconforming material.
(e) Testing - The testing of one ATP-High Range Area Monitor Readout Module was witnessed. The test was conducted in accordance ,
with drawing 02181-0445, which outlines the test procedure. No nonconformancec were observed during the test.
(f) Qualification and training records were reviewed for eight Sorrento Electronics personnel. No nonconformances were noted and the records were easily retrievable.
1 99
ORGANIZATION: SORRENTO ELECTRONICS SAN DIEGO, CALIFORNIA REPORT INSPECTION wn - ooonn997/n7_n1 ,
RFSUITS- PAGE 6 of 8
- 3. Palo Verde ESFAS Cabinet Problem On April 28, 1986, the PVNGS Emergency Diesel Generator (EDG) load sequencer panel failed to initiate a pa.ticular sequence of automatic actions including EDG start due to deficient internal vendor wiring.
On April 29, 1986, the licensee initiated an inspection of vendor wiring and components in several electrical enclosures. During the inspection it was discovered that the AMP brand multiple-pin electrical connector assemblies used in the balance-of-plant (B0P) ESFAS control circuits had inadequate pin engagement. The licensee determined that the inadequate pin engagement could cause an open circuit when the connector assemblies are moved. Further inspection revealed several other vendor workmanship problems. For a complete understanding of this issue, the NRC inspector contacted PVNGS personnel and AMP representatives in addition to SE representatives.
It appears that an inadequate connector application review by SE during the design phase was probably the cause for the incorrect connector pin length. However, it appears that inadequate control by SE of its fabrication work activities at one point in time was another cause. This perception is based in part on the following deviations that were discovered:
(a) April 10, 1986 EDG load sequencer cabinets - A 10 CFR Part 21 report was submitted by SE. The issue involved an improper transistor grounding problem in the cabinets; (b) April 28, 1986, EDG load sequencer cabinets - Problems were discovered with SE installed internal control circuit wiring.
The deviations caused a loss of power initiation when internal wires were " jiggled;" and (c) April 29, 1986 ESFAS cabinets - Inadequate connector asrembly multi-pin engagement length, internal vendor wiring inadequacies, and miscellaneous electrical and mechanical workmanship problems.
This is considered an unresolved item pending SE's identification and NRC's review of the root cause of these deficiencies and whether other nuclear power plants may be affected.
- 4. Component Subassembly Fabrication During the inspection it was learned that SE purchases and receives certain raw materials and components at its San Diego Facility and then packages the material into manufacturing kits. The kits are then transferred to CIMEX, a sub-division of SE that is located in 100
ORGANIZATION: SORRENTO ELECTRONICS SAN DIEG0, CALIFORNIA REPORT INSPECTION Mn - 000nntA7/R7_n1 RFSill TS - PAGF 7 of P Tijuana, Mexico, for subassembly and component fabrication. After fabrication, the completed product is returned to SE, San Diego, for final inspection and testing.
- 5. Prior SE Management Commitments The SE President and General Manager previously committed to the NRC that an improved quality assurance awareness program would be established and carried out. Sorrento Electronics has established such a program, and it provides company wide periodic briefings to indoctrinate and reiterate the SE quality program to its employees.
- 6. Radiation Monitor Coaxial Cable 10 CFR Part 21 Report Recently, SE notified the Commission of an inadequate coaxial cable dielectric insulation value for its design application use with SE's radiological monitoring system ion chambers. Brief discussions were conducted with SE staff and the NRC inspector expressed his concern that the issue appears to be an SE design problem rather than a coaxial cable application problem. The NRC will continue to follow SE actions as the evaluation progresses.
F. PERSONS CONTACTED:
Name Organization
- M. L. Jones, QA Director Sorrento Electronics C. Fisher, Project Manager GA Technologies
- N. Porter, QA Manager Sorrento Electronics 4 J. Bakelarski, QAE Sorrento Electronics i E. McCoy, QAE Sorrento Electronics
- J. Couris, QC Manager Sorrento Electronics G. L. Logan, Logistics Manager Sorrento Electronics D. Koutahi, Purchasing Manager Sorrento Electronics G. Keeten, Buyer Sorrento Electronics S. Lupan, Engineer Sorrento Electronics
- S. G. Lewis, Logostics Director Sorrento Electronics
- T. A. Hoshenrose, Programs Director Sorrento Electronics .
- D. C. Nav, Engineering Director Sorrento Electronics l
- C. R. Phipps, Operations Director Sorrento Electronics P. Petiford, Resident DCASA US D0D B. Grey, QAE Sorrento Electronics
- Attended exit meeting.
101
ORGANIZATION: SORRENTO ELECTRONICS SAN DIEGO, CALIFORNIA 1
REPORT INSPECTION I on . oconntn7/n7_n1 RESULTS: PAGE 8 of 8 l
Name Organization l M. Chamberland, Fabricator Sorrento Electronics S. Oliver, Test Sorrento Electronics !
R. Newton, QA/QC Sorrento Electronics l J. Malone, Test Sorrento Electronics F. McCord, QAE Sorrento Electronics ,
R. Treat, Operator Sorrento Electronics '
J. Carstensen, Auditor Sorrento Electronics I
- Attended Exit Meeting.
t I
l 102
ORGANIZATION: U.S. TOOL AND DIE ALLIS0N PARK, PENNSYLVANIA REPORT INSPECTION INSPECTION DATES: 03/23-27/87 ON-SITE Hnt!Ps. 11n NO.: 99901082/87-01 CORRESPONDENCE ADDRESS: U.S. Tool and Die ATTN: Mr. Michael T. Rodgers President
- 4030 Route 8 l Allison Park, Pennsylvania 15101 ORGANIZATIONAL CONTACT
- Mr. Frank E. Witsch TELEPHONE NUMBER: 412-487-7030 i
NUCLEAR INDUSTRY ACTIVITY: Fabricator of spent fuel storage racks.
l I l
l l
ASSIGNED INSPECTOR: hIb A ist N ks 6 9 .
Claudia M. Abbate, Trogram' Development and Reactive "Da'te I Inspection Section (PDRIS)
OTHER INSPECTORS: James T. Conway, PDRIS Kenneth G. A pinwall, Consultant APPROVED BY: [. X M 8MP7 Japes C. Stone, Chief, PDRIS, Vendor Inspection Branch ' Da'te INSPECTION BASES AND SCOPE:
A. BASES: 10 CFR Part 50 Appendix B, 10 CFR Part 21.
i B. SCOPE: Observe the fabrication and testing processes regarding the fabrication of spent fuel storage racks. Welding, nondestructive testing, personnel training, quality control inspection, procurement, shop QA implementation and quality records were reviewed.
PLANT SITE APPLICABILITY: Callaway, Ginna, Kewaunee, LaSalle 2, Nine Mile Point 2, Seabrook, Shoreham, Vermont Yankee, Wolf Creek i
I I
103
ORGANIZATION: U.S. TOOL AND DIE ,
ALLISON PARK, PENNSYLVANIA
~
REFORT I INSPECTION NO.: 99901082/87-01 RESULTS: PAGE 2 of 15 ;
l A. VIOLATIONS- i
- 1. Contrary to Section 21.31 of 10 CFR Part 21, a review of documentation packages for spent fuel storage racks fabricated under ASME Code Section III, Subsection NF revealed that while 10 CFR Part 21 was imposed on U.S. Tool & Die (UST&D) by their customers, UST&D did not !
specify that 10 CFR.Part 21 requirements would apply on Purchase '
Orders (P0) 86-60208 to Columbia Electric Manufacturing; 84-1701 and
-1679 to Do All Pittsburgh; 86-61228 to Cromie Machine & Tool; l 82-1051 to Commercial Fasteners; 86-60802, -81208 and 87-70132 to l West Penn Laco; 82-103/ to Allegheny Ludlum Steel; 86-60620 to ,
Industrial Service' Centers; 83-1387 to Sandvik; 87-70115 to Weld 1 Star; 86-61112, -70103 and 87-70118 to Alloy-0xygen Weld Supply; ;
86-61218 to Metal Goods; and 86-60921, 87-70130 and -70208 to j Williams and Company. (87-01-01)
This is a Severity Level V violation (Supplement VII). . j
- 2. Contrary to Section 21.6 of 10 CFR Part 21, UST&D failed to post Section 206 of the Energy Reorganization Act of 1974. (87-01-02) }
This is a Severity level V violation (Supplement VII).
B. NONCONFORMANCES:
- 1. Contrary to Criterion X of Appendix 8 to 10 CFR Part 50, Section 14.3C of the UST&D Quality Assurance Program Manual (QAPM),
Revision 2, and Section 3.3 of Procedure 14.1, " Production Work Routing and Inspection Plan," Revision 0, no in-process examinations were being performed at the south shop per the " Production Work Routing and Inspection Plan," Drawing 8601-0, Revision 1, for spent fuel racks being fabricated for the LaSalle project. (87-01-03)
- 2. ' Contrary to Criterion X of Appendix B to 10 CFR Part 50, Section 10.2A of the UST&D QAPM, Revision 2, and Section 2.2.1 of Procedure 10.4, " Final Inspection," Revision 1, two undersize Welds were identi-fied by the NRC inspectors on a fuel rack which had been inspected and found acceptable by UST&D Quality Control. (87-01-04)
- 3. Contrary to Criterion XII of Appendix B to 10 CFR Part 50, Section 12.2B of the UST&D QAPM, Revision 2, and Sections 2.4 and 2.6 of Procedure 12.1, " Control of Measuring and Test Equipment," Revision i 2, a review of measuring and test equipment (M&TE) and calibration records revealed the following:
i 104
ORGANIZATION: U.S. TOOL AND DIE ALLIS0N PARK, PENNSYLVANIA REPORT INSPECTION NO.: 99901082/87-01 RESULTS: PAGE 3 of 15 i
- a. Documented evidence was not available to verify that the ,
calibration interval for M&TE was established in procedures. l l
b.- The four inner diameter (ID) Box Mandrels for the Kewaunee, q Vermont Yankee, and LaSalle projects in the south fabrication J l
shop were not identified with a permanent unique S/N applied by vibro-etching or with a durable label.
- c. A calibration interval was not established in the "Cardex" 'l system for the modified ID Box Mandrel for the LaSalle project in'the north fabrication shop. (87-01-05)
- 4. Contrary to Criterion IV of Appendix B to 10 CFR Part 50, Section 2.2A of the QAPM, Revision 2 Section 5 of ANSI N45.2 and Section 5 (Basic Requirements) of ANSI /ASME NQA-1, a review of 40 P0s for j materials and services related to spent fuel racks fabricated under l ASME Code Section III, Subsection NF indicated that quality requirements (e.g., QA Program) were not passed on to vendors for ]
the following P0s: 86-60746, -60813, -60814, and -61228 to Cromie i Machine and Tool; 86-6028 to Columbia Electric Manufacturing; !
84-1701 and -1679 to Do All Pittsburgh; 82-1068 to Capitol Pipe & I Steel Products; 82-1051 to Commercial Fasteners; 86-6082, -81208, i and 87-70132 to West Penn Laco; 82-1032 to Allegheny Ludlum Steel; !
85-51023 to Techalloy; 82-1311 and 83-1387 to Sandvik; 87-70115 to l Weldstar; 86-70103 to Alloy-0xygen Weld Supply; and 86-61218.to 1 Metal Goods. (87-01-06) !
- 5. Contrary to Criterion IV of Appendix B to 10 CFR Part 50, Sections 4.2B and 4.4B of the QAPM, Revision 2, Sections 2.2, 3.2 and 4.1 of Procedure 4.1, " Procurement Document Control," Revision 5, and Section 5.3 of NES Specification No. 83A2256, a review of P0s indicated that PO 87-70118 to Alloy Oxygen Welding Supply for stainless steel weld filler metal did not contain a delta ferrite ;
determination statement, and P0 66-60610 to WALC0 Corporation for markers and tape was not signed / initialed and dated by QA personnel.
(87-01-07) l
- 6. Contrary to Criterion V of Appendix B to 10 CFR Part 50, Section ,
5.2B of the QAPM, Revision 2, and Section 6.0 of ANSI N45.2, it was noted that a documented procedure / instruction did not exist to control tools (e.g., wire brushes, grinding wheels, hammers, etc.)
that were designated for use only on stainless steel material.
(87-01-08) .
'105
ORGANIZATION: U.S. TOOL AND DIE ALLISON PARK, PENNSYLVANIA REPORT INSPECTION NO.: 99901082/87-01 RESULTS: PAGE 4 of 15
- 7. Contrary to Criterion V of Appendix B to 10 CFR Part 50, Section 5.2B of the UST&D-QAPM, Revision 2, and Section 2.2 of Procedure.10.3, "Inprocess Inspection," Revision 3, Procedure 10.3.does not provide the QC. inspector appropriate guidance indicating the required random sample quantities or percentages needed to ensure a representative sample during in-process examinations. (87-01-09)
- 8. Contrary to Criterion XIV of Appendix B to 10 CFR Part 50, Section 14.2A of the UST&D QAPM, Revision 2, and Sections 2.3 and'3.2 of Procedure 10.3, "Inprocess Inspection," Revision 3, marking procedures used by the south shop QC inspector for in-process examination do not clearly identify items or components which have satisfactorily passed the required examinations or who examined the part. (87-01-10)
- 9. Contrary to Criterion VII of Appendix B to 10 CFR Part 50, ,
Section 7.2B of the UST&D QAPM, Revision 2, and Section 1.1.1 of '
Procedure 7.2, " Evaluation of Vendors," Revision 1, vendor evaluations had not been performed on Do All Pittsburgh and !
Columbia Electric Manufacturing who provide calibration services for UST&D. (87-01-11)
- 10. Contrary to Criterion II of Appendix B to 20 CFR Part 50, Section 2.5 of the UST&D QAPM, Revision 2, and Section 2.1 of Procedure 2.3,
" Training," Revision 3, no training records existed for two UST&D shop employees. (87-01-12)
- 11. Contrary to Criterion VII of Appendix B to 10 CFR Part 50, Section 7.28 of the UST&D QAPM, Revision 2, and Section 2.1 of Procedure 7.2, " Evaluation of Vendors," Revision 1 PO 86-60143, dated February 14, 1986, was placed with Industrial Service Center prior to the vendor evaluation which was performed March 12-13, 1986.
(87-01-13)
C. UNRESOLVED ITEMS:
None.
D. STATUS OF PREVIOUS INSPECTION FINDINGS:
! None. This was the first NRC/VIB inspection.of this facility.
106 1
ORGANIZATION: U.S. TOOL AND DIE ALLISON PARK, PENNSYLVANIA REPORT INSPECTION N0.: 99901082/87-01 RESULTS: PAGE 5 of 15 E. INSPECTION FINDINGS AND OTHER COMMENTS:
- 1. Entrance and Exit Meetings An entrance meeting was conducted on March 23, 1987 at the Allison Park, Pennsylvania office of UST&D. The purpose and scope of the inspection were discussed during this meeting. UST&D has two fabrication shops. The south shop, located in Glenshaw, Pennsylvania, receives the material and forms and welds the individual cells.
The north shop, located in Allison Park, Pennsylvania, assembles and welds the cells into fuel storage racks and performs the final l inspection. During the exit meeting conducted on March 27, 1987, the inspection findings and observations were discussed with UST&D personnel.
- 2. Spent Fuel Racks - huclear l l
Since 1982, UST&D has fabricated spent fuel racks for six commercial nuclear customers (see Table 1, page 15). Three projects are currently in progress: LaSalle 2, Kewaunee and Vermont Yankee. The customer's design is being used for the Vermont Yankee and Kewaunee projects, while UST&D's design is being used for the LaSalle project. To date, only five prototype cells have been fabricated for Vermont Yankee. A number of cells have been fabricated for Kewaunee, but they have not been assembled into a completed rack.
I The NRC inspector reviewed the customer's procurement packages for the nine projects including a detailed review of the P0s and/or the technical specifications for four projects: Nine Mile Point 2, Vermont Yankee, Kewaunee and LaSalle 2. For all four projects, it j was noted that the. requirements of 10 CFR Part 50, Appendix B and l 10 CFR Part 21 were referenced in the P0s and/or specifications. '
The Nuclear Energy Services (NES) specification for the Vermont Yankee project referenced the 1980 Edition of Sections II, III (Subsection NF), V and IX of the ASME Code, whereas the 1977 Edition of the same Sections were listed as applicable documents in the i Stone and Webster (S&W) specification for the Nine Mile Point 2 )
project. NDE personnel were to be trained and qualified per SNT-TC-1A, and inspectors were to be qualified to AhSI N45.2.6. 3 Weld filler metal was to be in accordance with Subsection NF of Section III and specification AWS A5.9. In general, the material specifications included ASTM A240, A276, and A564. Cleaning 107
ORGANIZATION: U.S. TOOL AND DIE ALLISON PARK, PENNSYLVANIA REPORT INSPECTION NO.: 99901082/87-01 RESULTS: PAGE 6 of 15 requirements were noted as Class B of ANSI N45.2.1; and handling, packaging, and shipping were to meet the Level C requirements of-ANSI N45.2.2. Record retention was in accordance with ANSI N45.2.9 and typical documents shipped to the customer included: certified material test reports for stainless steel and weld material, heat treatment certifications (17-4 PH material), NDE reports, inspection reports, nonconformance reports, repair and rework reports, and a certificate of conformance from UST&D.
l
- 3. Plant Tour The inspectors toured the north and south fabrication facilities at ;
various times in the company of UST&D officials. Receipt inspection. l press forming of stainless steel sheet and welding of cells were l activities noted in the south shop. Items witnessed in the north '
shop included machining, installation of poison material (i.e.,
boraflex strips), welding of support plates, PT examination, welding j of racks, in-process inspection, cleaning and final inspection. j I
During the plant tours, it was noted by the inspector that one hammer had a painted handle. UST&C personnel informed the inspector that the painted tools are used on projects which involve stainless steel-l while the unpainted tools are used on projects which involve carbon l
steel. There was no documented instruction / procedure to control the use of painted and unpainted tools.
Nonconformance 87-01-08 was identified in this area.
- 4. Production Work Routina & Inspection Plan (PWRIP)
A PWRIP., which is similar to a " shop traveler," is generated by the Project Manager for each nuclear job. The PWRIP is laid out as an E-size drawing and identifies, for both the north and south
~
shops, each production operation, CC inspection /in-process i examinations, and customer witness points and mandatory hold points.
Several notes pertaining to the requirements of the customer's specification, inspections 4acluding documentati'on, and hold points ;
are also included. Item No. 7 of the PWRIP states that "documen-tation is not always required" for in-process examinations. This does not agree with Section 14.2 of the QAPM which states, in part, i "Ir.spection reports will document all inspections and testing j delineated on the Production Work Routing & Inspection Plan." The 2
108
1 ORGANIZATION: U.S. TOOL AND DIE ALLISON PARK, PENNSYLVANIA l
REPORT INSPECTION NO.: 99901082/87-01 RESULTS: PAGE 7 of 15 i PWRIP, including changes, is approved by the Engineering and QA departments. The PWRIP came into effect upon the generation of the '
QAPM in December 1981, and the document was used on all the nuclear projects beginning with Wolf Creek. The inspector reviewed and verified that a PWRIP existed for the nine nuclear projects since 1982.. ,
During the inspection of shop activities, UST&D Drawing 8601-0, Revision 1, " Production Work Routing and Inspection. Plan," for the LaSalle project was reviewed. After the individual cell is tack-welded and jetline welded together, en in-process examination and weld inspection were to be performed. These two activities are outlined in Procedures 10.3 and 10.5 respectively.
When documentation or inspection reports were requested for the in-process examinations and weld inspections, none existee. No in-process examinations were being performed on the fuel racks in the initial stages of fabrication at the south shop.
Nonconformance 87-01-03 was identified in this area.
Upon further review of the procedure, it was noted that in-process examinations were defined as periodic random sampling type checks to determine that the shop is providing material which complies with the UST&D QA/QC program requirements. The shop QC personnel were unable to define, quantitatively or with a percentage, the number of periodic random sampling type checks to be performed by the inspector. '
Procedure 10.3 is inadequate in that it does not give sufficient i instructions to the QC inspector as to what quantity a random sample ,
is and does not ensure a consistent number of in-process examinations i among the different inspectors. '
Nonconformance 87-01-09 was identified in this area.
In addition, Procedure 10.3 does not require documentation of the in-process examinations, however, each part may be physically marked after an in-process examination has been performed. The option of marking components to indicate completion of an in-process examination is used by the south shop QC inspector. However, the l mark used is a check mark made on the component surface with a i
i marking pen. Several similar marks are also made on the component surface during manufacturing. There is no instruction or guidance in Procedure 10.3 for the QC inspector or other shop personnel on how to distinguish the inspector's markings from other manufacturing
, markings.
109
ORGANIZATION: U.S. TOOL AND DIE ALLISON PARK, PENNSYLVANIA REPORT INSPECTION NO.: 99901082/87-01 RESULTS: PAGE 8 of 15 Nonconformance 67-01-10 was identified in this area.
- 5. Welding and Weldin3 Machines Welding being performed on spent fuel storage racks for the LaSalle l and Vermont Yankee projects was observed. Three welders were I performing production work at the time of the inspection. The I following items were reviewed during the inspection: presence of I Welding Procedure Specifications (WPS) at the work area, preheat 1 and interpass temperature control, compatibility of WPSs to j production work and compliance with WPS essential and nonessential '
variables.
Welding machines were identified with a unique identification number for verification of amperage and voltage requirements.
Automatic welding equipment had attached documentation to alert the welding operators of amperage and voltage ranges and tolerances for the thickness of the material being used. In addition, each manual welding machine contained a controlled record showing each welder who used the machine, date of use, and WPS used. This record is used as a summary of weld performances for welder qualification records.
No items of nonconformance or unresolved items were identified :
in this area. !
- 6. Weld Inspection Final inspection of the completed spent fuel storage racks is outlined in Procedure 10.4, Revision 1, and includes a verification i of the physical dimensions of the completed rack to the approved shop drawing dimensions. .
During the inspection of completed fuel storage racks for the LaSalle project, visual and dimensional checks of fillet welds on fuel rack j pedestal mounting pads were performed by the NRC inspector. Drawing ;
8601-30, Revision 4, was used for the inspection. During the !
inspection, two fillet welds on the mounting pad were found to be less than the one-quarter inch weld specified on the drawing. The j l
rack was identified as UST&D rack number 11 (Conrnonwealth Edison rack number 2FC16C). A review of the final inspection checklist i revealed that these welds had been inspected and found acceptable by the UST&D QC inspector.
i k
i 110
(
ORGANIZATION: U.S. TOOL AND DIE ALLISON PARK, PENNSYLVANIA REPORT INSPECTION NO.: 99901082/87-01 RESULTS: PAGE 9 of 15 Prior to the exit meeting, the welds were repaired. The new welds were examined by UST&D QC personnel and found acceptable. The NRC inspector verified that the welds were acceptable per the drawing.
Nonconformance 87-01-04 was identified in this area.
- 7. Welder Qualifications and Training The welder cations (WPS),qualifications to applicable Procedure Qualification Welding Records (Procedure'Specifi-PQR) and design specification requirements were reviewed. Each welder record included his unique welder identification number, the WPS to which he was qualified and the supporting PQRs. Training records for four welders were reviewed. The records documented which training session the welders attended and the date (see Section 16 of this report).
The UST&D QA program requires that all personnel qualified to WPS 53 for automatic spot / fusion welding be trained to the applicable WPS, have practical experience and perform two test samples prior
! to production welding. The records of e'ight welders qualified to WPS 53 were reviewed and found in compliance with the require-ments mentioned above.
The review of welder qualifications and training resulted in no items of nonconformance or unresolved items; however, training was minimal in that welders are trained only to the WPSs and there is no i
evidence that welders are trained to new revisions of WPSs.
- 8. Weld Filler Material Control A review cof weld filler material control was performed using UST&D Procedure 9.2, Revision 4. The areas which were examined included:
- weld filler material storage areas, storage of filler material, marking of material (straight and spools), assignment of material inventory control (MIC) numbers and issuance / return records. These items were inspected at the weld filler material issuance stations at both the north and south shops. The review of weld filler material control indicated that Procedure 9.2 was being implemented.
No items of nonconformance or unresolved items were identified in this area, i 111
ORGANIZATION: U.S. TOOL AND DIE ALLISON PARK, PENNSYLVANIA REPCRT INSPECTION N0.: 99901082/87-01 RESULTS: PAGE 10 of 15
- 9. Nondestructive Examination (NDE) Personnel Qualifications The hDE qualification records were reviewed using UST&D Procedure 2.5', Revision 1. Certifications and qualifications of the UST&D Level III NDE consultant, the QC Manager and two QC irspectors were reviewed. Prior training hours; general, practical, and specific tests, including the amount of questions per discipline and test results; and a verification of annual eye examinations received by the personnel were reviewed. Records of training and indoctrination of personnel to the NDE procedures and program were also reviewed.
The review of documentation for NDE personnel qualifications resulted in no items of nonconformance or unresolved items; however, documented evidence of training of NDE personnel was minimal.
- 10. ControlofMeasuringandTe_s_t_ Equipment _(MT&E)
The NRC inspector reviewed applicable sections of the QAPM, one procedure and calibration records to determine whether M&TE was !
i properly identified, controlled and calibrated at specified intervals. Inspection areas in the north and south fabrication '
)
shops were inspected to review the calibration status of gages and measuring instruments fourd in these areas.
With the exception of four If box mandrels, the inspected equipment contained a vibro ~-etched S/N. The four mandrels were for the Kewaunee, LaSalle and Vermor,t Yankee projects and were located in the south shop. A Cardex 'ystem, which is. maintained by the QC Manager, is used to recoro calibration information. Each card identifies the type of equipment including S/N, calibration date, calibration frequency, the standard used, and the identity of the individual performing the calibration.
Equipment examined at the south shop included a vernier caliper, two micrometers, one thickness gage, one width gage, one tong test ammeter, and four ID box mandrels. At the north sho ammeter, three micrometers, one dial thickness gage,p, one one tong test vernier caliper, one bore gage, one radius gage set, one inside micrometer set, one optical comparator, one box mandrel (LaSalle project) and three micrometer standards were examined. In addition, the calibration status of the eight welding machines in the north shop was checked. The information contained on the calibration stickers on the items was in agreement with the applicable card.
112
.0 ORGANIZATION: U.S. TOOL AND DIE ALLISON PARK, PENNSYLVANIA REPORT INSPECTICh NO.: 99901082/87-01 RESULTS: PAGE 11 of 15 The two AC-DC tong test ammeters (S/N AX-48975 and -58261) are sent to Columbia Electric Manufacturing (CEN) for calibration every two years. Test reports from CEM indicated that the equipment was calibrated in February 1986 with standards traceable to the National Bureau of Standards (NBS). A Certificate of Calibration from Do All Pittsburgh indicated that a master gage block set (S/N 0-4) was calibrated in October 1584 with standards traceable to NBS. It was noted that a document did not exist to identify the specific equipment covered by the calibration program including the calibration frequency of such equipment.
Nonconformance 87-01-05 was identified in this area.
- 11. Procurement Document Control Forty-three P0s to 15 material manufacturers / suppliers, four P0s to a machining vendor, three P0s to two vendors for calibration services, and one P0 to a plating vendor were reviewed to assure that applicable technical and QA program requirements were included or referenced in the P0s. With the exception of one order, all the P0s were initialed and dated by a QA individual. The P0 in question was No. 86-60610, dated June 6,1986, to WALC0 Corporation for felt tip markers and nuclear grade cloth tape. Nineteen P0s did not invoke the requirements of 10 CFR Part 21 upon vendors - eight P0s to four manufacturers / suppliers of weld wire, three P0s to two calibration service vendors, one P0 to a machining vendor, one PO to a supplier of fasteners, and six P0s to manufacturers / suppliers of stainless steel. In addition, the requirement that a vendor have a QA program which was approved by UST&D was not included in nineteen P0s to vendors - four P0s to Cromie Machine and Tool, one PO to Columbia Electric Manuf acturing, two P0s to Do All Pittsburgh, one P0 to Capitol Pipe & Steel Products, one P0 to Commercial Fasteners, three P0s to West Penn Laco, one PO to Allegheny Ludlum, one P0 to Techalloy, two P0s to Sandvik, one FO to Weldstar, one P0 to Alloy-0xygen Weld Supply and one P0 to Metal Goods. It was also noted that for PO 87-70118, dated January 16, 1987, to Alloy-Oxygen Welding Supply for 200# ASME Section II SFA 5.9 Type 312 filler metal did not contain a "oelta ferrite determination" statement.
Violation 87-01-01 and Nonconformances 87-01-06 and 87-01-07 were identified in this area.
- 12. Docurrientation Packages (DP)
A DP did not exist for the Vermont Yankee and Kewaunee projects as finished racks were not yet completed for these projects. The 113
1 ORGANIZATION: U.S. TOOL AND DIE ALLIS0N PARK, PENNSYLVANIA REPORT INSPECTION NO.: 99901082/87-01 RESULTS: PAGE 12 of 15 inspector reviewed three DPs for Rack M120-24'for Wolf Creek and Racks No. 5 and 9 for LaSalle. For Wolf Creek, the DP consisted of a Bechtel Engineering and Quality Verification Document; Material.
Certification Data Sheet; CMTRs from Eastern Stainless Steel, Allegheny Ludlum Steel, and Colt Crucible for stainless steel products; a CMTR from Sandvik, who is a certificate holder, for ER 308 L weld wire; and a certificate of conformance-(C of C) from Commercial Fasteners for fasteners. Several forms for box identifi-cation / inspection, bottom plate identification / inspection' lead-in guide assembly to rack, and visual weld inspection reports were included. A Bechtel Supplier Deviation Disposition Request and a UST&D Nonconformance Report and Verticality Test Report were also part of the package. In addition, a UST&D C of C noted that all materials used in the fabrication of the reference rack assembly conformed to Bechtel specification 10466-C-175.
The DPs for the two LaSalle racks consisted of the follcwing:
Documentation Checklist, Shop Bill of Material (SBM), CMTRs, UST&D C of C, Weld Examination Records, Final Inspection Checklists (FIC),
Deviation / Variance Requests, Qualified Welders List, and a C of C
)
for the boraflex. The SBN listed the material inventory control (MIC) number as well as the material specification and supplier for '
each item (e.g., box _ half, bottom plate, pedestal body, etc.), The j FIC verified that the following activities were accomplished on a ~
finished rack: dimensional verification, MIC No. verification, visual weld inspection, cleaning, marking, boraflex verification, and the mandrel insertion test. j No items of nonconformance or unresolved items were identified in i this area.
- 13. Vendor Evaluations The NRC inspector reviewed the evaluations UST&D performed on its vendors and how approved vendors remain on the Approved Vendors 1.ist, This process is outlined in Procedure 7.2, " Evaluation of Vendors," Revision 1, dated March 12, 1987.
Eleven vendor evaluation packages were reviewed. The packages contained the original survey or audit checklie of the vendor and i reevaluations of the vendor. The reevaluation > consisted cf a Vendor Evaluation Questionnaire and a historical quality performance data review.
114
ORGANIZATION: .U.S. TOOL AND DIE ALLIS0N PARK, PENNSYLVANIA REFORT INSPECTION l NO.: 99901082/87-01 RESULTS: PAGE 13 of 15 l During the review,'it was noted that Do All Pittsburgh and Columbia Manufacturing Electric, who supply calibration services, have not been evaluated by UST&D, nor has a historical quality performance data record been maintained on either vendor. Also during the review, it was noted that Industrial Service Center had been audited by UST&D March 12-13, 1986, while P0 86-60143 to Industrial Service-Center had been placed February 14, 1986. This evaluation was not performed within the specified 14 days of issuance of the P0 as j required by Section 2.1 of Procedure 7.2. i l
Nonconformances 87-01-11 and 87-01-13 were identified in this area.
! 14. 10 CFR Part 21 I
! During the inspection, the NRC inspectors examined areas for the J l posting of 10 CFR Part 21 at both fabrication shops. It was noted j that 10 CFR Part 21, UST&D Procedure 16.2 and Public Law 96-295 Act of June 30, 1980, Section 223 were posted, but Section 206 of the Energy Reorganization Act of 1974 was not posted at the south shop.
Section 21.6 of 10 CFR Part 21 and UST&D Procedure 16.2, " Reportable Defects and Noncompliance (Nuclear Projects) 10 CFR 21," Revision 2, dated March 12, 1987, require Section 206 to be posted. Prior to :
the exit meeting, Section 206 and other applicable documents were l posted et both shops
! Violation 87-01-02 was identified in this area.
- 15. Training l The NRC inspector reviewed UST&D Procedure 2.3, " Training," Revision 3, dated March 23, 1986. The procedure requires all UST&D personnel performing quality related activites to be trained as applicable to their activity. The procedure also requires that training records be maintained.
The NRC inspector reviewed 13 employee training records and training i session records. The employee training records identify what training the employee has received while the training session <
records identify who attended the training, the date and an outline !
of the subjects which were covered during the training.
During this review it was noted that two machinists employed in :
the shop had training record files, but had not received any training.
Also noted were the facts that a training record, with no documented training, was present for a retired employee; several training 115
l i
ORGANIZATION: U.S. TOOL AND DIE ALLISON PARK, PENNSYLVANIA l
I REPORT INSPECTION N0.: 99901082/87-01 RESULTS: PAGE 14 of.15 l
records were incomplete in that hire dates and training session j subjects were missing; and two employees were trained outside the 45 (
day limit for new eraployees. In addition, UST&D only trains its- !
personnel to the quality control procedure which the employee will i be using without an overall QA/QC program training session. The !
personnel are not trained to new revisions of the procedures nor is .
there any type of refresher. training given to current employees l after the initial training course and after they have been employed -
for a lengthy amount of time.
Nonconformance 87-01-12 was identified in this area. !
- 16. Audits !
The NRC inspectors reviewed four internal audits. The Internal Project Management Audit, dated May 29, 1986, the Production Audit, dated August 13, 1986, the QA/QC Audit, dated November 12, 1986, and the Engineering Department Audit, dated March 4, 1987, were reviewed. ;
Each audit included a checklist, an audit report (summary of results),
and was performed by a qualified auditor. The audits which were reviewed were performed and documented in accordance with Procedure 18.1, " Audits," Revision 5, dated March 12, 1907.
No items of nonconformance or unresolved items were identified in this area. I F. PERSONS CONTACTED:
l P. Brinks, QC Manager, South Shop l F. DeSimmons, Foreman, North Shop W. Dickson, Director of Manufacturing E. Jablonsky, QC Inspector, North Shop i
L. Karenboyer, Welder
- R. Linder, Manager of Engineering
- E. March, Vice Chairman of the Board, Chief Executive Officer E. Reinhart, QC Manager, North Shop J. Rhoden, Project hanager F. Rhodes, Vice President, Manufacturing R. Rudisill, Welder
- M. Rodgers, President R. Stewart, Machinist
- B. Wachter, Vice President, Engineering and Research K. Weber, Assistant Project Manager l *F. Witsch, CA Manager
- Attended Exit Meeting 116
i ORGANIZATION: U.S. TOOL AND DIE ALLIS0N PARK, PENNSYLVANIA REPORT INSPECTION NO.: 99901082/87-01 RESULTS: PAGE 15 of 15 TABLE 1 Number of Date Utility /AE Project Type R,ac_ks/ Positions Completed SNUPPS Wolf Creek PWR-MDR 12/1328 1982 Kansas Gas & Non-Poison Electric /Bechtel
'SNUPPS Callaway PWR-MDR 12/1328 1982 l Union Electric / Non-Poison Bechtel Long Island Shoreham BWR-Non 15/1700 1983 l l Lighting / Stone Poison 8/756 '
l & Webster Water Flux 2/144 l Trap Control Rods Public Services Seabrook New Fuel 3/90 1983 Company Storage Racks Niagara Mohawk Nine Mile II BWR-MDR Stone & Webster Poison 17/2530. 1984 10/1519 Rochester Gas & Ginna Modified 6/420 1984 Electric PWR Checker-board to Poison Racks Commonwealth LaSalle II BWR Poison 20/4073 In Progress Edison Wisconsin Public '
Service Corp. Kewaunee PWR Poison 4/360 In Progress Nuclear Energy Vermont Services Yankee BWR Poison 10/2820 In Progress ~
117/118
ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION MONR0EVILLE, PENNSYLVANIA REPORT INSPECTION INSPECTION N0.: 99901043/87-01 DATES: 03/27/87 ON-SITE HOURS
- 6 CORRESPONDENCE ADDRESS: Westinghouse Electric Corporation Nuclear Services Integration Divison ATTN: Mr. T. Christopher General Manager Monroeville, Pennsylvania 15230 ORGANIZATIONAL CONTACT: Mr. J. Jalovich TELEPHONE NUMBER: 412-374-3470 NUCLEAR INDUSTRY ACTIVITY: Refurbishment of 480 volt circuit breakers used as Reactor Trip Breakers and dedicating commercial grade cor..ponents to class 1E components.
()
ASSIGNED INSPECTOR: Y ..
l< , 0, 2
K. R. Naidu, Program Development and Reactive Date Inspection Section (PDRIS)
OTHERINSPECTOR(5):
APPROVED BY: A #/3/4-7 J. C(/ Stone, Chief, PDRIS, Vendor Inspection Branch Dhte INSPECTION BASES AND SCOPE:
A. BASES: 10 CFR Part 21, Appendix B to 10 CFR 50,
- 8. SCOPE: Witness tests performed on OT-2 type push button switch which malfunctioned at Point Beach Nuclear Power Station Unit 2 on March 17, 1987.
PLANT SITE APPLICABILITY: Point Beach (50-301).
119
ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION MONROEVILLE, PENNSYLVANIA REPORT INSPECTION Nn - qqqn1na1/A7-01 RESULTS: PAGE 2 of 3 A. VIOLATIONS:
No violations were identified during this inspection.
B. NONCONFORMANCES:
No nonconformances were identified during this inspection.
C. UNRESOLVED ITEMS:
No unresolved items were identified during this inspection.
D. INSPECTION FINDINGS AND OTHER COMMENTS:
- 1. Background On March 17, 1987, Wisconsin Electric Power Corporation (WEP), the owner of Point Beach Nuclear Power Station, reported that a OT-2 type push button (PB) manufactured by Westinghouse malfunctioned during the monthly operability testing of the reactor trip breakers (RTB). The WEP procedure ICP 2.3, Revision 18, dated May 16, 1986, titled, " Periodic Test Reactor Protection System Logic," requires the operator to depress the PB to verify that the undervoltage trip i device (UYTD) actuates in response to a demand from the reactor trip system logic train A to trip RTB A. Depression of the PB prevents the shunt trip coil (STC) actuation while verifying operability of the UVTD. Subsequently, the PB is released to verify that the STC actuates. During the test at Point Beach Unit 2, when the PB was released, the STC did not actuate. The PB was replaced dnd the STC operability was verified successfully. The failed OT-2 type PB was brought to Westinghouse by a WEP representative.
- 2. Tests Performed on the Push Button On March 27, 1987, tests were performed on the PB at the Westinghouse's Seco Road facility. A written procedure was used to perform the following tests:
- a. Visual examination of PB.
- b. Contact resistance measurement and verification of continuity when the PB was depressed and released.
120
ORGANIZATION: WESTINGHOUSE ELECTRIC CORPORATION MONRDEVILLE, PENNSYLVAhlA W
REPORT INSPECTION NO.- 99901043/87-01 RESULTS: PAGE 3 of 3
- c. The movements of the operator button and plunger rod.
- d. Disassembly of the PB and examination of the contacts.
t No abnormalities were observed. The physical dimensions of the PB components were in accordance with the manufacturing drawings.
Representatives from Westinghouse and WEP concluded that the PB failure was a random non-repeatable event.
E. E
_1XIT INTERVIEW:
The inspector met with individuals mentioned in Section F and discussed the scope and findings of the inspection.
F. PERSONS CONTACTED:
Westinghouse Electric Corporation R. Steve, Project Manager, Point Beach J. Jalovich, Manager, Equipment Qualification Group A. K. Deb, Engineer F. Folino, Quality Assurance Engineer Wisconsin Electric Power Corporation J. Meyer, Electrical Engineer 121/122
ORGANIZATION: WYLE LABORATORIES HUNTSVILLE, ALABAMA REPORT INSPECTION INSPECTION l DATES: 02/03-04/87 ON SITF HnHRS- 19 NO.: 99900902/87-01 l CORRESPONDENCE ADDRESS: Wyle Laboratories Scientific Services and Systems Group ATTN: Mr. W. W. Holbrook, General Manager i Eastern Test and Engineering Operations 7800 Governors Drive Huntsville, Alabama 35807 ORGANIZATIONAL CONTACT: Mr. James Gleason, Manager, Nuclear Engineering TELEPHONE NUMBER: (205) 037-4411 i I
NUCLEAR INDUSTRY ACTIVITY: Wyle Laboratories, Huntsville, Alabama, provides a variety of nuclear services to the industry including environmental and seismic qualification testing of safety-related equipment, refurbishment and decertification of valves, valve and component flow testing, and mechanical and hydraulic snubber testing, decontamination and repair.
i l
ASSIGNED INSPECTOR: a w S. D. Alexander ( Special Projects Inspection h UBte l Section(SPIS)
OTHERINSPECTOR(S):
APPROVED BY: t] Ib W,%w kt 2 f 7 U. 'Polapovs. ChiW, SPI (, Vendor Inspection Branch ate /
INSPECTION BASES AND SCOPE:
A. BASES: 10 CFR Parts 21 and 50.49 and 10 CFR Part 50, Appendix B.
B. SCOPE: This inspection consisted of: (1) a technical evaluation of equipment qualification (EQ) test activities for safety-related equip-ment,(2)examinationoftestsetupandspecimensand(3) verifying that specifications and applicable regulatory requirements were met.
PLANT SITE APPLICABILITY: Susquehanna 1/2 (50-387/388), Dresden 2/3 (50-237/
249), LaSalle County Station 1/2 (50-373/374), Byron 1/2 (50-454/455)
(continued on next page).
123
i ORGANIZATION: KYLE LABORATORIES '
HUNTSVILLE. ALABAMA REPORT INSPECTION l Mn . QQQnnon?/97 01 I RFSill TS ? PAnF ? nf a
{
l PLANT SITE APPLICABILITY: (continued) Braidwood 1/2(50-456/457), Quad i Cities 1/2(50-254/265), and Zion 1/2 (50-295/304).
A. _ VIOLATIONS:
None.
B. NONCONFORMANCES:
None.
C. UNRESOLVED ITEMS:
None.
D.
STATUS OF PREVIOUS INSPECTION FINDINGS:
Not addressed during this inspection.
E.
0THER FINDINGS AND COMMEN_TS:
1.
_Backp)ound. In May 1986, Pennsylvania Power and Light Company (PP&L installed Marathon 1600 tbs in Limitorque valve actuators at Susquehanna Units I and 2 to replace unqualified tbs that were discovered during the licensee's followup inspections to IE Information Notice 86-03. {
In September 1986, PP&L EQ personnel became aware that the 1600s (and 1500s) had failed in Marathon sponsored EQ tests at Wyle Laboratories (hyle Report No. 45603-1 dated February,1982). The samples had been thermally aged for 444 hours0.00514 days <br />0.123 hours <br />7.34127e-4 weeks <br />1.68942e-4 months <br /> at 248 F. Failure occurred during loss-of-coolant-accident (LOCA) simulation at 365 F when the samples were energized with 528 Vac for an operability test using a motor external to the test chamber as a load.
The failure mode was shorting with the attendant inability to maintain the tbs energized at 528 Vac.
PP&L-sponsored testing of the 1600s was completed recently at Wyle for Susquehanna's drywell environment. These samples had been thermally aged for 185 hours0.00214 days <br />0.0514 hours <br />3.058862e-4 weeks <br />7.03925e-5 months <br /> at 248 F. Three Marathon 1600s were mounted in a NEMA-4 enclosure in the test chamber. Failure occurred on one of the samples in this test during the LOCA exposure at 360*F when the samples were energized with 528 Vac for the first operability test following spray initiation. The failure mode was a short circuit between the terminals of two 124
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ORGANIZATION: WYLE LABORATORIES J HUNTSVILLE, ALABAMA
]
i REPORT INSPECTION I WA . 'Q00noQn9/A7 n1 RF91lf TS ? Pf4F '4 nf a J
phases and short circuits from at least one terminal to ground on the mounting panel of the test enclosure across the insulated base of the TB.
- 2. Purpose. The purpose of this inspection was to gather information on the results of the above testing and environmental qualification testing of certain Amp Corporation electrical connector assemblies and Marathon Corporation terminal blocks performed by Wyle for..the Commonwealth Edison Company,
- 3. Review. The inspector reviewed documentation associated with the ;
- above testing, interviewed Wyle and Commonwealth Edison technical personnel and examined the test specimens. Also examined was the ,
Marathon terminal block test setup including the Loss-of-Coolant- !
Accident (LOCA) simulation chamber and the test NEMA enclosure in it; in which the three Marathon 1600 terminal block specimens I were still mounted.
l a. The Marathon test setup consisted of a spraytight.(but not l
steam-tight) steel electrical terminal block enclosure gounded to the LOCA chamber with 3 Marathon 1600s mounted on the inside of its back panel. The tbs were positioned as follows: two with their long axis oriented vertically mounted on either side of the center TB which had its long axis oriented horizontally.
The tbs were wired in series with three phases and leads from each one exiting the LOCA chamber such that they could be connected and disconnected externally and .inde. pendently for jumpering and or measurement of insulation resistance during the test. A 1-horsepower, 480 Vac, 3-phase motor was connected as a load for the operability checks. Each phase was protected by a 12 amp fuse and there was a leakage current monitoring circuit protected by a 4 amp fuse. The power supply protective device was set for twice the locked-rotor current for the load motor at rated voltage. The acceptance criterion was that it should not trip. The center, horizontal TB was the one that failed the test. It clearly showed signs of arcing from the energized terminals near the center of the block downward across the surface of the insulated base to the enclosure back panel at the lower side of the block base.
Review of the raw test data indicated that after the first six hours of steam exposure, chemical spray was initiated.
Less than one minute later, an operational accident performance test was initiated by energizing the tbs with 528 VAC. Within 125
J
(
i a
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ORGANIZATI0th ' WYLE LABORATORIES '
HUNTSVILLE, ALABAMA REPORT INSPECTION Nn - 00ConQn?/n7-01 RESULTS; $
PAGF 4 of 4 ,
1 seconds of energization, the 12 amp fuses in two of three i phases blew as well as the 4 amp fuse in the leakage current 1 monitoring circuit. '
j
- b. Review of documentation associated with the CECO sponsored 1 Wyle tests of the Amp connectors indicated that they had -
developed excessive leakage current during the test when energized with 132 VDC. {
, j Examination of the test specimens of the nylon insulated crimp type' butt-splice connectors (supplied with GE F01 electrical penetration assemblies) revealed c, lear signs of arcing through the connector insulation near'one of the j larger diameter end " bells" which accepts some, of the wire '
insulation.
- 4. Conclusions. Theinspector'sreviewconfirmedthabthetesting l had been conducted in accordance with specifications and ;;
regulatory requirements and that the testing was technically ?
adequate for its intended purposes. s ; l 1
The inspector further concluded that the results demonstrated the i inability of the Amp nylon insulated connectors.and one of the Marathon 1600 terminal blocks to meet their specified functional performance requirements under the simulated accident conditions, and that the failures appeared to represent environmentally induced failures; thus rendering them unqualified'for the applications for which they were tested.
The results of testing on the newer design, nuclear grade Amp connectors indicated that these connectors could be qualifed for the service and harsh environmental conditions under which they performed satisfactorily in the test.
126
1 SELECTED INFORMATION NOTICES l
l
SSINS No.: 6835 IN 87-19 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 April 9, 1987 l l
l IE INFORMATION NOTICE NO. 87-19: PERFORATION AND CRACKING OF ROD CLUSTER CONTROL ASSEMBLIES ,
f l
l Addressees: l All Westinghouse nuclear power pressurized-water reactor (PWR) facilities holding an operating license or a construction permit. ,
Purpose:
This notice is provided to infonn recipients of a potentially significant l safety problem that could result from the perforation and cracking of the rod l cluster control assemblies (RCCAs) in Westinghouse PWRs. It is expected that l recipients will review the information for applicability and consider action, j as appropriate, to preclude a similar problem from occurring at their facili- 1 ties. However, suggestions contained in this information notice do not consti- ,
tute NRC requirements; therefore, no specific action or written response is required, i l
Description of Circumstances:
An estimate that was intended to be conservative indicated that the RCCAs would I last for at least 15 years before the absorber cladding, a thin tube, would show excessive thinning as a result of sliding wear. These components were inspected at Point Beach Nuclear Plant, Unit 2, in 1983 after 13 years of operation. As a result of this inspection Point Beach reported on August 18, 1983 that sliding wear was minor, but one control rod had a 2-in. crack near the tip of the rod and severe fretting wear had occurred on several tubes.
Subsequent inspections at the Kewaunee and Haddam Neck nuclear power plants, which have been in operation for more than 12 years, confirmed the fretting wear. In addition, Haddam Neck reported tube cracking in 32 of 47 RCCAs.
In the event of a breach of the tubing resulting from wall thinning, perfora-tion, or cracking, the immediate effect is the introduction of activation products from the neutron absorber material into the reactor coolant. Although there are large margins, another concern is the potential reduction in shutdown margin and negative reactivity worth.
W 8704080095 128/I29
IN 87-19 April 9, 1987 Page 2 of 3 Discussion:
Each RCCA contains 16 rods. The rods at Point Beach, Kewaunee, and Haddam Neck were constructed with an outer tube of 0.019-in.-thick 304 stainless steel that retains the absorber material (80% silver, 15% indium, 5% cadmium). Some newer plants use hafnium as the absorber material, while others use boron carbide surrounded by a 0.038-in.-thick tube.
The control RCCAs are inserted or withdrawn to compensate for various reactivity changes'during operation of the reactor and can trip to provide shutdown capability. The shutdown RCCAs are fully withdrawn from the core when the reactor is critical.
At Kewaunee, marks of fretting wear about 1 inch in length, were found adjacent to the guide blocks that position the rods when the RCCAs are in their withdrawn position.
The 1-in.-thick stainless steel blocks are spaced on 12-in. centers and each rod in the cluster pgsses through all eight of the blocks. At Point Beach the tubing wore in two modes: fretting and sliding of the rods over the guide blocks during rod motion.
resulting from fretting and four c,. Five RCCAs at daddam Neck had wall thinning absorber material. All of the othersthese had were actually wearing into the fretting wear, but to a lesser extent.
The fretting resulted from flow-induced vibratory contact between the rods and the guide blocks during long periods of steady-state power operation. Vibration, is hydraulically induced by flow of the reactor coolant; therefore it is a continuous process when the reactor coolant pumps are in operation.
to Westinghouse Electric Corp. fretting wear encompassed one-third of theAccording circumference of the rod and the depth varied with the amount of time the RCCAs were in the withdrawn position.
At Point Beach significant number of short hairline cracks at the lower extremity of the tubing were observed near the end plug region of the rod. The cracks extendedmaterial.
absorber axially for 4 in, and penetrated the stainless tubing, exposing the No circumferential cracks were foun~d. Examination of the cracks showed that irradiation-induced swelling of the absorber was the principal cause of tensile stress in the cladding, which resulted 'in cracking after substantial irradiation.
Where excessively worn rods were found, they have been replaced. While the issue is being studied by NRC and the industry, several licensees have been given approval to slightly change the position of the fully withdrawn RCCA in order to distribute the wear among different locations on the tubing. Westing-house Electric Corp. reported that an increase in the amount of the silver isotope, Ag-110m.in the reactor water is a reliable indication of exposure of absorber material due to cracking or fretting wear.
130
IN 87-19 April 9, 1987 Page 3 of 3 The NRC is continuing review of the safe'ty significance of this information to determine whether further NRC action is warranted. 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.
l -
ard Jordan, Director Divisi of Emergency Preparedness and Engineering Response Office of Inspection and Enforcement Technical
Contact:
Paul Cortland, IE (301) 492-4175
Attachment:
List of Recently Issued IE Information Notices 131/132
SSINS Noo: 6835 TN 87-24 (fNITED STATES FUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WSHINGTON, D.C. 20555 l
June 8, 1987 NRC INFORMATION NOTICE NO. 87-?4: OPERATIONAL EXPERIENCE INVOLVING L0r ES OF ELECTRICAL INVERTERS Addressees:
All nuclear power reactor facilities holding an operating license or a con-struction permit.
Purpose:
This notice is provided to alert recipients of potential problems involving l electrical inverter losses that have led to unplanned plant transients and/or l inoperability or improper functioning of' safety-related and other important plant equipment. It is expected that recipients will review this information for applicability to their facilities and consider actions, as appropriate, to preclude similar problems from occurring at their facilities. However, sugges-tions contained in this notice do not constitute NRC requirements; therefore, no specific action or written response is required.
Past Related Correspondence:
IE Information Notice 84-80, " Plant Transients Induced By Failure of Non-Nuclear Instrumentation Power," November 8,1984 IE Rulletin 79-27, " Loss of Non-Class 1E Instrumentation and Control Power System BUS During Operation," November 30, 1979 IE Information Notice 79-29, " Loss of Nonsafety-Related Reactor Coolant Systen Instrumentation,During Oper6 tion," November 16, 1979 IE Circular 79-02, " Failure of 120 Volt Vital AC Power Supplies,"
January 11, 1979
Background:
Inverters in nucicar power plants provide "uninterruptible" vital ac electrical power to safety- and non-safety-related instrumentation and control systems.
Generally, loss of this function results in some type of undesirable system condition and/or plant transient, including unnecessary actuation of safety systems such as reactor protection and engineered safeguards systems; loss of indicators that provide plant status information; system disturbances, 8705290040 133
fN 87-2A 1une 4, 1987 Page 2 of 3 including reactor coolant system transients; improper response of the feedwater and steam generator water level control systems; loss of safety-related elec-trical equipment functions; damage to mechanical equipment; and challenges to operators and t.he remaining functional equipment. Such conditions and/or transients clearly have significant safety implications since they result in challenges to safety equipment and plant operations and/or a degradation of plant equipment.
The NRC case study report, AEOD/0605 dated December 1986, " Operational Experi-ence Involving losses of Electrical Inverters," includes the review of 94 licensee event reports (LERs), totaling 107 events involving inverter losses that occurred during 1982 through 1984. The study includes 35 additional events from the Nuclear Plant Reliability Data System (NPRDS) thet occurred in the same timeframe. These 142 events occurred at 51 distinct plants: 26 designed by Westinghouse, 11 by General Electric, 9 by Combustion Engineering, 4 by Babcock & Wilcox, and 1 by General Atomic. The total number of events included in the studv for each of the 3 years along with the number of reactor units which were operating during each of those years is summarized below.
Parameter 1982 _1983 1984 Operating Reactor linits 72 74 82 Inverter Loss Events 34 51 57 Losses / Operating Reactor 47 .69 70 As indicated above, the NRC has issued information on inverter losses since 1979; and industry groups have issued approximately 14 reports related to this issue.
Description of Circumstances:
The NRC case study report identified three potential failure mechanisms for inverters. One of these involves relatively high ambient temperature and/or humidity within inverter enclosures. This condition appears to result in accelerated aging of components that form a part of the inverter circuitry causing a significant reduction in component life expectancy and inverter loss.
Another mechanism for inverter failure involves the electrical interconnecting and physical arrangements for the inverter circuitry components. In some installations, these arrangements are such that when certain components fail, other components also may fail or degrade.
The third failure mechanism involves voltage spikes and perturbations. Many of the electrical loads in a plant have inductive characteristics. During plant operations that involve energizing and deenergizing these loads, voltage spikes and perturbations are generated. The solid-state devices in the inverter cir-cuitry are sensitive to these voltage spikes, and this has resulted in component failure, blown #Uses, and inverter losses. Additionally, secondary voltage per-turbations caused by lightning strikes or switching surges can have an adverse effect on inverter operation.
134
IN 87-24 June A, 1987 Page 3 of 3 Discussion:
The NRC case study report indicates that the failure mechanisms involving service condition parameters (e.g., ambient temperature and/or humidity and voltage spikes and perturbations) have common-cause implications. However, none of the events reviewed and evaluated in the report involved the simulta-neous loss of redundant inverter-powered buses.
The dominant cause of inverter losses was attributed to component failures.
Such components include diodes, fuses, silicon cor.2ro11ed rectifiers, capacitors, transistors, resistors, printed circuit boards, transformers and inductors. It also appears that major contributing factors for the occurrence of component failure events are high ambient temperature and/or humidity within inverter enclosures and electrical disturbances at the inverter input / output terminals.
In addition, incorrectly operating circuit breakers, improperly setting up test equipment, removing the wrong inverter unit from service, and improperly transferring power sources for a bus are some personnel actions that make them the second largest contributor to inverter losses.
It is suggested that licensees consider, monitoring of temperature and/or humidity internal to inverter enclosures and evaluating input and output voltages of the inverter unit during steady-state and transient conditions to assure that manu-facturer's recommendations are being considered. Additionally, to minimize the number of inverter loss events resulting from personnel actions, licensees might consider reviewing related maintenance and testing procedures and practices for inverters. Further, specialized training and practice sessions with involved plant personnel and veri'ication of appropriate sequence of steps to achieve desired related maintenance and tc' sting. activities also may be considered.
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.
har'es . Rossi, Director Division of Operational Events Assessment Office of Nuclear Reactor Regulation Technical
Contact:
Vincent D. Thomas, NRR (301)492-4414
Attachment:
List of Recently Issued NRC Information Notices 135/136
SSINS No.: 6835 IN 87-26 UNITED STATES ilVCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS WASHINGTON, D. C. 20555 June 11, 1987 NRC INFORMATION NOTICE NO. 87-26: CRACKS IN STIFFENING RINGS ON 48-INCH-DIAMETER UF CRINDERS 6
l Addressees:
All uranium fuel fabrication and conversion facilities.
Purpose:
This notice is being issued to alert recipients to a possible safety problem related to UF,. cylinders. It is suggested that recipients review the informa-tion and consYder actions, if appropriate, to preclude possible safety problems at their facilities. However, suggestions contained in this Information Notice do not constitute NRC requirements; therefore, no specific action or written [
response is required.
Description of Circumstances:
The NRC has received information from the manufacturer that some 48-inch-diameter UFg cylinders manufactured by the W. H. Stewart Company of Oklahoma City.
OkTahoma, have stiffening rings fabricated from a material (ASTM A306, Grade 75 steel) which is not specified in AflSI Standard N14.1-1982, or in a previous edition thereof and which is not an equivalent approved material under the ASME Boiler and Pressure Vessel Code,Section VIII, Division 1. The cylinders in question are identified by the following manufacturer's serial numbers:
2309 through 2333 2442 through 2617 2782 through 2828 However, because W. H. Stewart Company records are incomplete, other 48-inch-diameter cylinders manufactured after May 1975 may be affected. There-fore, this Infontation Notice is also applicable to any other UF have cause to believe contains stiffening rings made of ASTM A30h, cylinder Grade you 75 steel.
Discussion:
~
The Department of Transportation (00T) has published a rule (see 51 Fed. Reg. 46675, December 24, 1986) that after June 30, 1987, will reoutre UF shipping containers to meet the provisions of ANSI Standard N14.1-1982, or a6previous edition thereof [49 CFR Part 173.420(a)(2)l. The ANSI Standard specifies the design of UF 6 cylinders, including materials of construction for stiffening 8706050209 137
IN 87-26 June 11, 19'87 Page 2 of 2 rings. The ANSI Standard also pennits substitute materials to be used for stiffening rings provided they are equivalent approved materials under the ASME Boiler and Pressure Yessel Code,Section VIII, Division 1, which are compatible with fabrication of the cylinders involved. Shipment of cylinders that do not meet the ANSI Standard is not authorized under the new DOT rule scheduled to become effective on June 30, 1987.
Wedge-shaped cracks have been detected on some of the stiffening rings on identified cylinders.
The cause and safety significance of the cracks have not been determined. The stiffening rings may serve multiple safety and functional purposes; they are welded to the lifting lugs and may contribute to the strength of those lugs.
Therefore, licensees may want to consider inspecting the cylinders in question for cracks in the stiffening rings, lifting lugs, and associated weldments, including those weldments that join the stiffening rings to the cylindrical shell and to the lifting lugs. Extra care and caution should be used when lifting or handling any cylinder that has cracked stiffening rings, lifting lugs, or weldments.
DOT regulates the transport of UF, cylinders. Shipment of UF cylinders that 6
do not meet the ANSI Standard is not authorized under the new 00T rule scheduled to become effective on June 30, 1987. D0T should be consulted before you ship any of the identified cylinders off site.
No specific action or written response is required by this Information Notice.
If you have questions about this matter, please contact those listed below.
$0w = -
Richard E. Cunningham, Dire r Division of Fuel Cycle, Medical, j
Academic, and Commercial Use Safety 1 Office of Nuclear Material Safety i and Safeguards 00T Technical Contact on transportation questions:
Michael E. Wangler Telephone: (202)366-4498 NRC Technical Contact on all other questions:
Leland C. Rouse Telephone: (301)427-4309
Attachment:
List of Recently Issued NRC Information Notices 138
Attachment 1 IN 87-26 June 11, 1987 LIST OF RECENTLY ISSUED INFORMATION NOTICES 1987 Information Date of Notice No. Subject Issuance Issued to 87-?5 Potentially Significant 6/11/87 All nuclear power Problems Resulting from reactor facilities Human Error Involving holding an OL or CP.
Wrong Unit, Wrong, Train, or Wrong Component Events.
87-24 Operational Experience 6/4/87 All nuclear power Involving Losses of reactor facilities Electrical Inverters holding an OL or CP.
87-23 Loss of Decay Heat Removal 5/27/87 All PWR facilities During Low Reactor Coolant holding an OL or CP.
Level Operation 87-22 Operator Licensing Requali- 5/22/87 All research and fication Examinations at nonpower reactor Nonpower Reactors facilities.
87-21 Shutdown Order Issued Because 5/11/87 All nuclear power Licensed Operators Asleep facilities holding While on Duty an OL or CP and all licensed operators.
87-20 Hydrogen Leak in Auxiliary 4/20/87 All nuclear power Building facilities holding an OL or CP 86-108 Degradation of Reactor 4/20/87 All PWR facilities Sup. 1 Coolant System Pressure holding an OL or CP.
Boundary Resulting from Boric Acid Corrosion 86-64 Deficiencies in Upgrade 4/20/87 All nuclear power Sup. 1 Programs for Plant facilities holding Emergency Operating a CP or OL.
Procedures.
85-61 Misadministration to 4/15/87 All licensees Sup. 1 Patients Undergoing Thyroid authorized to use Scans byproduct material 87-19 Perforation and Cracking of 4/9/87 All Westinghouse Rod Cluster Control Assemblies power PWR facilities holding an OL or CP OL = Operating License CP = Construction Permit 139/140
SSINS No.: 6835 IN 87-30 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C.
July 2, 1987 NRC INFORMATION NOTICE NO. 87-30: CRACKING OF SURGE RING BRACKETS IN LARGE GENERAL ELECTRIC COMPANY ELECTRIC MOTORS Addressees:
All nuclear power reactor facilities holding an operating license or a con-struction permit.
Purpose:
This notice is provided to alert recipients to a potentially significant safety problem that could result in the loss of safety-related equipment, such as residual heat removal (RHR), core spray, and high-pressure pumps that are driven by large, vertical electric motors manufactured by General Electric Company (GEi. It is expected that recipients will review this information for applicability and consider actions, as appropriate, to preclude a similar problem at their facilities. However, suggestions contain'e d in this notice do act constitute NRC requirements; therefore, no specific action or written respon,se is required.
Background:
Felt blocks are used in large electric motors to keep the windings separated where they loop back at the end of the stator. The blocks are attached to a surge ring that is held in place by L-shaped surge ring brackets welded to the surge ring and bolted to the motor casing. Failure of these surge ring brack-ets and cracking of the felt blocks allows movement and wear of the end-turns, leading to a reduction in insulation resistance and possible motor failure. In addition, broken pieces of the surge ring bracket may enter the space between the stator and the rotor, resulting in electrical or mechanical motor degrada-tion.
Description of Circumstances:
During the 1985 outage at Peach Bottom Atomic Power Station, Unit 3, broken surge ring brackets were discovered in a 2000-horsepower (hp) RHR pump motor 8706250407 141
IN 87-30 July 2, 1987' Page 2 of 3 1
and a 600-hp core spray pump motor. Attempts to repair the brackets by welding were unsuccessful, and they were removed from the motors at GE's suggestion.
The licensee' for Peach Bottom is reevaluating the motor design and the need to replace the brackets.
During the week of November 10, 1986, the licensee for Cooper Nuclear Station inspected its 1000-hp RHR pump motors and discovered that three of four motors had broken lower surge ring brackets and cracked upper and lower end-turn felt blocks. In the 8 motor, five of six lower brackets were fractured. Repairs l were made and periodic visual inspections are being conducted. I GE informed the NRC on March 24, 1987 that the equipment affected in BWR plants includes the RHR, core spray, and high-pressure core spray system electric moto rs . GE recommended that annual inspections be perfonned until operating experience indicates that this is no longer necessary. At plants where the support brackets have been removed. GE stated that restoration actions snould be taken as soon as possible. Although GE notified BWR licensees of the Cooper Station event, it did not notify PWR licensees. A list, which was provided to the NRC by GE, of PWR facilities that have received large, vertical electric motors from GE is attached. It should be noted that other plants could have similar motors.
Discussion:
The surge ring bracket, a 1-inch-wide by 1/8-inch-thick L-shaped piece of ,
carbon steel, has been breaking at the ' sharp. bend. After the bracket breaks, a I 1-inch-long portion remains attached t'o the surge ring by a double fillet weld. I If this weld fails, the 1-inch by 1-inch by 1/8-inch piece of steel may move i inside the motor. Although tests conducted at Cooper Station showed signifi- l cant cyclic loading of the bracket when the motor was started, the bracket also was shown to be subject to vibration during steady-state operation.
At Cooper Nuclear Station, surge ring brackets of an improved design, which incorporated a larger bend radius, were installed and the motor insulation system was revarnished to fill and bond the cracks in the felt blocks. The licensee is visually inspecting the brackets on a regular basis. This examina-tion is conducted without disassembling the pump motor, using either a boroscope or a mirror inserted through the existing air vents. GE has recom-mended a complete disassembly and inspection at 10-year intervals to ensure the continued qualification of these motors.
142
IN 87-30 July 2, 1987 Page 3 of.3 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.
QCharesE.Rossi,Directorr%
Division of Operational Events Assessment l
Office of Nuclear Reactor Regulation Technical
Contact:
Paul Cortland, OSP (301)492-7190 Attachments:
- 1. List of PWR Facilities Known to Have Received Large, Vertical Electric Motors from General Electric Company
- 2. List of Recently Issued NRC Infonnation Notices .
Attachment 1 IN 87-30 July 2, 1987 Page 1 of 1 LIST OF PWR FACILITIES KNOWN TO HAVE RECEIVED LARGE, VERTICAL ELECTRIC MOTORS FROM GENERAL ELECTRIC COMPANY The plant numbers were not identified for multi-unit facilities.
MOTOR SIZE IN HORSEPOWER MNT Arkansas Nuclear One 800 Bellefonte Nuclear Plant 900 Farley Nuclear Plant 600 Harris Nuclear Power Plant 1300 Indian Point Station 400 Millstone Nuclear Power Station 500, 600, 700 North Anna Power Station 400 Palo Verde Nuclear Station 600 Salem Nuclear Generating Station 1000 Seabrook Nuclear Station 600, 800 Sequoyah Nuclear Plant 700 SNUPPS 1750 St. Lucie Plant 500, 600 Turkey Point Plant 500 Vogtle Nuclear Plant 700 143/144
l INDEX REPORT NUMBER PAGE FACILITY The Amerace Corporation 1 Union, New Jersey 99900296/86-01 Brand Industrial Services, Inc.
99901020/87-01 15 Park Ridge, Illinois Combustion Engineering, Inc. 25 Windsor, Connecticut 9990C002/86-01 General Electric Company 41 San Jose, California 99900403/87-01 Inland Steel Company 59 East Chicago, Indiana 99901081/87-01 ITT Barton Instruments Company 99900113/87-01 65 City of Industry, California Paul Monroe Hydraulics, Inc.
99900337/87-01 77 Orange, California Rockwell Engineering Company, Inc. 83 Blue Island, Illinois 99900836/87-01 Sandvik Special Metals Corporation 89 Kennewick, Washington 99900764/87-01 Sorrento Electronics 95 San Diego, California 99900387/87-01 U. S. Tool and Die Allison Park, Pennsylvania 99901082/87-01 103 Westinghouse Electric Corporation 119 Monroeville, Pennsylvania 99901043/87-01 Wyle Lab 0ratories 123 Huntsville, Alabama 99900902/87-01 145
INDEX (continued) l l
Selected Information Notices INFORMATION NOTICE # SUBJECT PAGE 87-19 PREFORMATION AND CRACKING 129 OF ROD CLUSTER CONTROL ASSEMBLIES 87-24 OPERATIONAL EXPERIENCE 133 INVOLVING LOSSES OF ELECTRICAL INVERTERS 87-26 CRACKS IN STIFFENING RINGS 137 ON 48-INCH-DIAMETER UFs CYLINDERS 87-30 CRACKING OF SURGE RING 141 BRACKETS IN LARGE GENERAL ELECTRIC COMPANY ELECTRIC MOTORS 146
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.......l.....:.....:.....l.....; -
.....:.....j.....:.....t.....l ....
.....l..... .....: ...[..... -...l COMBUSTION ENGINEERING l PWR FACILlilES WITH FUEL SUPPLIED BY CDMBUSil0N ENGINEERING I INCORPORATED I I
- .....:.....: ...;- - .....l .....:.....; ..... ....l ;.....:.....t..... .....:.....:.....
BENERAL ELECTRIC COMPANY I POTENTIAL I j l I i l 1 1 1 I l l l NUC ENER6Y BUS OPERATIONS! MULTI SliE I I ! : l l l 1
....:.....: .:.....l- -i.....:.....:.....:
INLAND STEEL COMPANY l l l I I I i ! ! ! ! l l 1 ! ! ! l l l
! ! I l ! I 1 i l ! l l 1 1 1 i ! l
.....l..... .....l.. .
.....l.....: _. : - .....:.....: . .i : -t.....: ...: ... _:.....l ITT BARTON INSTRUMENTS l ALL PLANTS USING ITT BATTON EDUIPMENT l COMPANY l i
---.............l.....:.....:.. .:.....l.....;- _:.....:.....l.....:..... ....
.!- i ..... ; -- :.....:.....;.....;
PAUL MONRDE HYDRAULICS, t i 1 l ! ! I l I I 1 ! ! ! I t i l i INCORPORATED ! ! I l 1 i i l t 1 1 l 1 i i l 1
.l.....;..... -:- y .....:.. .:.....g
.....l..... ; ....l.. .:.....].....:
ROCKWELL ENGINEERING 1 1 l l l l ! l I l l l 1 ! l ! l i 1 !
COMPANY INCORPORATED ! l ! ! 1 I I I l l l 1 1 I ! I I l l
.............. ... - i .. ... l .... . g -- :..... ..... ..... - _: ... .....l.. .] ....:.....:.....: l- ..:.....:.....- -!
SAXDVIK SPECIAL METALS I NUCLEAR POWER FACILITIES USIN6 CE, B&W, AND ANFC FUEL. I CORPORATION l 1
. .......;..... ..... _g.....: ...: ...:.....:.....:.. .:.....; ....: .: ...[.....l.....:..... .....:
SDRRENTO ELECTRONICS l 1 1 I I l l l l 1 I l ! ! ! ! l l 1 l l t i I I I i ! l l 1 l l 1 I I I i !
.....l.....: .;.....g.....:- .-
.. ...:..... ....t..... -
- .....!
- :.....l U.S. TOOL & DIE i I l I i i i l i I ! l l l 1 1 I i I i ! !
! 1 : I I l l 1 : I l l l 1- 1 I L t
...... .......;.....l- . : . ... . : .. .. . : ... .. . -- :.....:.....: ; ....g.....; : .. . : - i.....[ :.....
WESTINGHOUSE ELECTRIC l l l l l 1 i l l 1 l l ! ( I I i i l I CORPORAT10N 1 l ! l ! l l i l I l l 1 i i l I i i
.- ...........:.....:.....l.....;- :.....l.....;..... .....:.....:.....: ....;.....:.....
.....l.....:.....l.....l.....
WYLE LABORAf0 RIES i I 1 : I I I ! l I ! l l 1 I I I I i !
HUNTSVILLE,AL i 1 ! I I I I ! l t i I I I ! i l I
.l.....:.....l ... -
- ...: ....l..... :.. .l.....g.. .;.....:.. .;
148 I-APPLIES TO ALL FLANTS DOCKETNO.- APPLIES ONLY TO THE IDENTIFIED UNIT
VENDUR INSPECTION REPORTS RELATED TO REACTDR PLANTS e l 5 l S l 5! 3 : S S! S I S l T I T ! T i V lV I W ! W l W l W I I e A ! A i E i H ! O I T I U l U l H I R i U l E i 0 l A I A ! 0 ! N I I i .
e ! LI N i A i 0 1 ! I M l S I R I O!R I R l6 I i ! T l L ! P I O l e PLANTS I E I I B l R i T i L I M i 01 E I J l K I M l T l E I T l F I ! N !
l e i M i 0 1 R ! E E l U l E I U i E i A 1 EI O l L i R I S I l W l l e l l N I D l H I I l C l R 1 E I I N I Y l N I E : F l I C l P ! ! I e i 1 : 0 1 0 1 A I A : 1 I I H 1 M 1 I T I i 0 i B 1 R i P 1 & I e ! & l F l K l M l S ! EI l A i 1 I I P ! I I 1 R I A : E ! S 2 I e 2I R l ! l ! I I N I L I I OI Y l 6 l D 1 R I E i S 1 !
e l l E I I i P 1 1 1 1 N I E 1 i I 1 A l 2 I I K ! ! ! ,
e 1 l & 1 l R I & i 1 A I I I N I N I i 3 1 1 1 1 1 : I e ! I 1 1 2 : ! 0 l 2 l 1 l 1 1 1 T 1 K I I I & l ! 2 : I e l l 2 1 : I J l i l 1 1 S 1 1 I E I I l 2 l l 6 1 1 VENDORS e i I & 1 I I i 1 l & I ! ! 31 E I i 1 l l 3 i !
e 1 l 3 1 i ! ! l l 2 1 1 l l 6 I i ! 1 I l l l e ! t l l l & i l l 1&1 l 4 1 ! ! ! I l 1 i e ! l l ! 121 i i 121 1 : 1 I ! ! l 1 i
.......- : ... ; . ; ... ; ... ; . l ... l ... ; ... l ... ; ... ... l ... l ... ; . l ... ... ... ; ... l AAERACE CORPORATION i ALL NUCLEAR POWER STATIONS THAT USE ASASTAT ELECTRICAL l 3 i
i CONikOL REl,AY SERIES E-7000, E6P, ETR OR EML i
___:.....:.. .p....; -p....p. .; . . p....p. .. p. .p....p....:.....p. .p....; -p....p....:
BRAND INDUSTRIAL SERVICES: I I I I I I I I l I ! ! ! I I I I I l I 1 I I I I I I I I ! I I I i INCORPORATED (BISCO) l l l ! ! ! l l l 1 I l l l l l i ! !
.... p.... p....:..... p. . p.... p. .; .... .: p. .p. .:.....; ...:.....p....p.... ; ; ..
COMBUSTION ENGINEERING I PWR FAtiLITIES WITH FUEL SUPPLIED BY COMBUSTION ENGINEERING I INCORPORATED ! !
.. .. . . . p. ... p. .. . p. .p....:.. .: .... p.... p. . p. . p.... : ..... p. . ; p. .p. .p....p....;
6ENERAL ELECTRIC COMPANY I POTENTIAL l I I i but ENER6Y BUS OPERATIONS MULTI SITE : I I
--p....;- l -; - p ..p. .: .... p ...p. .p.... p....l_ -j -p....; ; p....l INLAND STEEL COMPANY l 1 l 1 ! ! ! 1 l l 1 1 I I I I I ! !
! l l 1 l ! 1 ! l ! ! I l 1 l l l i I
.........l p....p....l -! p....: .... p.... p.... p. .p.... p....p....p.... p....p....g -: :
ITT BART M INSTRUMENTS l ALL PLANTS USING ITT BATTON EQUIPMENT l COMPANY l l
.. -p.....p....l_ p....p....; ; . .l -l -l p.... p.. g -
p ...; p....;
PAUL MONRDE HYDRAULICS, 1 l 1 1 I I i ! l l 1 I I I I i i ! ! l INCORPORATED l l l l l 1 1 I i l I i ! 1 1 1 I l 1
_p....p. .p. .p.... - p....: . . p....l p. . p. .; -p....p....p....p....p....p....
ROCKWELL ENGINEERIN6 1 1 l 1 l l l l l l l 1 1 I l 1 l l l COMPANY INCORPORATED ! 1 l 1 i l l l l 1 1 I l l l l l 1 I
... .....p....;- -:.....p....l p. .: ....p.. p. p. .p. . p....:.....p.... p....j.....p.... p....]
SANDVIK SPECIAL METALS i NUCLEAR POWER FACILITIES USIN6 CE, B&W, AND ANFC FUEL. I CORPORATION l l
- : ..... l - l p.... p.... p....l .... p.... ...p....p....; p....:.....p....: ; t SORRENTO ELECTRONICS I I I i l I l ! l 1 ! l ! ! l l l l l l 1 I l t 1 I ! ! I ! ! ! ! ! l l l 1
- p.... p.. . p, . -
-l _- l l ....l p. . p. . p. . p.... p. ; p....p. .:.....p....l U.S. TOOL & DIE : l l I I I I I I I i 1 1 ! ! l i l 1 I l 1 l 1 l l l l l l 1 l ! I I I ! ! l ! ! l p.... p....:.. .l_ p.... p.... l . . p. . ; p....:.. . p. . p. . p._. p. .p.-- p....;- l WESTINGHOUSE ELECTRIC I i i ! ! l l 1 1 I I I I I l 1 ! l l CORPORATION 1 l l 1 l i l i i i l l 1 1 I l 1 i I
.. .................p....p. .:.....p....p....p.~.t . . p....p. .p .. p. p .p....p....p.... -p....p....!
WYLE LABORATORIES l' i I I l 1 ! ! I I I I I I i I i ! ! I I HUNTSVILLE,AL i i l I l i I ! I ! I 1 1 1 I l 1 I l
_.....p....;- p....p....p....p. .: ....p. .p. .;..... p....p....p....p. .p.-. p. .:.....: ;
149 I-APPLIES TO ALL PLANTS DOCKETNO.- APPLIES ONLY TO THE IDENTIFIED UNIT
I NRC PORM 336 U S NUCLE AR REGULAYoRY CoMMISSsoN t RE PoR T NUM8t N , Ass,aneir ay rioC. saa vo, No , ,, sori g fo' $$ BIBLIOGRAPHIC DATA SHEET NUREG-0040 j,#
Vol. 11, No. 2
$t t INST RUCTIONS ON MEVERSE 2 TifLL ANo SU9 tif Lt 3 LEAVE SLANK -'
Licensee ntractor and Vendor Inspection '
Status R Sort Quarterly R ort -- April 1987 - June 1987 ,0,,, ,,,,
. AUr oms, July 1987
. opt aeoaT issuno vsAs uo~T g Augyg6 1987
,no,0 o,~aoeaA~aA1so~NA EA A it,~o A oon ,, ,,,,,,, ,, c , 0~ U~i r ~Uu..a
. ,aca CT7.x Division of Reac r Inspection and Safeguards #
Office of Nuclear eactor Regulation ' " *
., G a 'N ' N" E "
l.
U.S. Nuclear Regul tory Commission .
Washington, D.C. '.555 jV w s,0~som~o ooo A~a Avio~ ~ Au A~o u A,ti~o goes33,,,,.. <,, c,, u.Tvno,arom Same as 7. above.
J Technical p4 si nmoo Covea eo ,,,,o- ,,o April 1987 - June 1987
.. , ,PL. . ...v~o,,,
g 13 AB5T R AC T <100 *ords or 'en/
\ ',
0 This periodical covers the re ltd of inspections performed by the NRC's Vendor Inspection Branch that V6 been distributed to the inspected organizations during the period rom April 1987 thru June 1987. Also, included in this issue are thed pults of certain inspections performed prior to April 1987 that werefnot 'ncluded in previous issues of NUREG-0040.
f j .
/
[ \
14 oOCUME NT AN ALV5#5 - a K E v J
R ol'o E SCR IP T ORS
\s -
16 A V AIL AS* LIT v STATEMENT Unlimited
$6 SECURITY CL A$$:F: CATION
,thos oeget Unc1assifieS
,rson reoem Unclassified 3 17 NVu8tA0FPAoES t6 PAiCE
- U, S.COVC RhmC hi PA lk f !NG Orr 1C0 :1987.191 697 : 60199
UNITED STATES speciat nevars.ctAss aArs NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 '05'*SME '
,,;^*r%9 E.e7 0FRCIAL BUSINESS PENALTY FOR PRIVATE USE, $300 1 1AN1NV V O M y ,
DC 20555 W~f3 gfg1NGTON
- --