ML20205A529
| ML20205A529 | |
| Person / Time | |
|---|---|
| Issue date: | 03/31/1999 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0040, NUREG-0040-V22-N04, NUREG-40, NUREG-40-V22-N4, NUDOCS 9903310009 | |
| Download: ML20205A529 (75) | |
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NUREG-0040 j Vol. 22, No. 4 I
Licensee Contractor and j Vendor Inspection l
4 Status Reaort ..
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Quarterly Report j
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l AVAILABILITY NOTICE l
Availability of Reference Materials Cited in NRC Publications NRC publications in the NUREG series, NRC regu- NRC Public Document Room l lations, and Title 10, Energy, of the Code of Federal 2120 L Street, N.W., Lower Level i Regulations, may be purchased from one of the fol- Washington, DC 20555-0001 lowing sources: < http://www.nrc. gov /NRC/PDR/pdr1.htm >
1 -800-397-4209 or locally 202-634-3273 l 1. The Superintendent of Documents U.S. Government Printing Office Microfiche of most NRC documents made publicly RO. Box 37082 available since January 1981 may be found in the Washington, DC 20402-9328 Local Public Document Rooms (LPDRs) located in
< http://www. access.g po. gov /s u_ docs > the vicinity of nuclear power plants. The locations 202 - 512 -1800 of the LPDRs may be obtained from the PDR (see
- 2. The National Technical Information Service Springfield, VA 22161 -0002 <http://www.nrc. gov /NRC/NUREGS/
<http://www.ntis. gov /ordernow> SR1350/V9/ipdr/html>
703 -487- 4650 Publicly released documents include, to name a The NUREG series comprises (1) brochures few, NUREG-series reports; Federal Register no-(NUREG/BR-XXXX), (2) proceedings of confer-tices; applicant, licensee, and vendor documents ences (NUREG/CP-XXXX), (3) reports resulting and correspondence; NRC correspondence and from international agreements (NUREG/lA-XXXX), internal memoranda; bulletins and information no-(4) technical and administrative reports and books tices; inspection and investigation reports; licens-
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microfiche, or, in some cases, diskette, from the <http://www. ansi.org>
Public Document Room (PDR): 212- 642 -4900 A year's subscription of this report consists of four quarterly issues.
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Vol. 22, No. 4 Licensee Contractor and Vendor Inspection Status Report Quarterly Report October- Cecember 1998 Manuscript Completed: March 1999 Date Published: March 1999 Division ofInspection Program Management Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Wcshington, DC 20555-0001 i s...../
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NUREG-0040, Vol. 22, No. 4 has been reproduced from the best available copy.
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l ABSTRACT l
This periodical covers the results of inspections performed between October 1998 and December 1998 by the NRC's Quality Assurance, Vendor inspection and Maintenance Branch that have been distributed to the inspected organizations.
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CONTENT 5 r
PAGE Abstract....................................................................................................................... iii I n t rod u ctio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . vil I n sp e cti on R ep o rts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 ACCUTECH (99901307/98-01 ) .................................. 2 Las Vegas, NV
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Nuclear Recearch Corporation (99901336/98-01 ) .................................. 15 Warrington, PA Valcor Engineering Corperation (99900728/98-01 ) .................................. 25 l
Springfield, NJ Westinghouse Electric Corporation (99900104/98-01) .................................. 41 Pensacola, FL Select Generic Correspondence on the Adequacy of Vendor ..................................... 66 Audits and the Quality of Vendor Products v
INTRODUCTION A fundamental premise of the U. S. Nuclear Regulatory Commission (NRC) licensing and inspection program is that licensees are responsible for the proper construction and safe and efficient operation of their nuclear power plants. The Federal government and nuclear industry have established a system for the inspection of commercial nuclear facilities to provide for multiple levels of inspection and verification. Each licensee, contractor, and vendor participates in a quality verification process in compliance with requirements prescribed by the NRC's rules and regulations (Title 10 of the Code of Federal Regulations). The NRC does inspections to oversee the commercial nuclear industry to determine whether its requirements are being met by licensees and their contractors, while the major inspection effort is performed by the industry within the framework of quality verification programs.
The licensee is responsible for developing and maintaining a detailed quality assurance (QA) plan with implementing procedures pursuant to 10 CFR Part 50. Through a system of planned and periodic audits and inspections, the licensee is responsible for ensuring that suppliers, contractors and vendors also have suitable and appropriate quality programs that meet NRC requirements, guides, codes, and standards.
The NRC reviews and inspects nuclear steam system suppliers (NSSSs), architect engineering (AE) firms, suppliers of products and services, independent testing laboratories performing equipment qualification tests, and holders of NRC construction permits and operating licenses in vendor-related areas. These inspections are done to ensure that the root causes of reported vendor-related problems are determined and appropriate corrective actions are developed. The inspections also review vendors to verify conformance with applicable NRC and industry quality requirements, to verify oversight of their vendors, and coordination between licensees and vendors.
The NRC does inspections to verify the quality and suitability of vendor products, licensee-vendor interface, environmental qualifimtion of equipment, and review of equipment problems found during operation and their corrective action. When nonconformances with NRC requirements and regulations are found, the inspected organization is required to take appropriate corrective action and to institute preventive measures to preclude recurrence. When generic implications are found, NRC ensures that affected licensees are informed through vendor reporting or by NRC generic correspondence such as information notices and bulletins, vii l
This quarterly report contains copies of all vendor inspection reports issued during the calendar quarter for which it is published. Each vendor inspection report lists the nuclear facilities inspected. This information will also alert affected regional offices to any significant problem areas that may require special attention. This report lists selected bulletins, generic letters, and information notices, and include copies of other pertinent correspondence involving vendor issues.
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INSPECTION REPORTS 1
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- November 25, 1998 1
Joe Casey, General Manager ACCUTECH 3873 W. Oquendo Road ,
Las Vegas, NV 89118
SUBJECT:
NRC INSPECTION REPORT 99901307/98-01
Dear Mr. Casey:
This letter addresses the inspection of your facility at Las Vegas, Nevada, conducted by Bill Rogers, Richard McIntyre, and Donald Naujock, of this office on October 26 through 29, 1998, and the discussions of their conclusions with David Rose, President of ACCUTECH, and other persons on your staff at the conclusion of the inspection.
Areas examined during the inspection are discussed in the ene!osed report. This inspection consisted of an examination of procedures and representative records, interviews with personnel, and observations by the inspectors. During this inspection it_was fcund that the corrective actions for previous violations and nonconformances, identified in insoection Reports 94-01 and 96-01, had been adequately implemented and documented, and we have concluded that those findings have been closed. In addition, during the review of your Quality Assurance program, within the scope of this inspection, we found no instance in which ACCUTECH failed to meet NRC requirements.
In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter and its enclosure will be placed in the NRC Public Document Room (PDR).
Sincerely, v.atR,.
Suzan . Black, Chief Quality Assurance, Vendor inspection and Maintenance Branch Division of Reactor Controls and Human Factors :
Office of Nuclear Reactor Regulation Docket No. 99901307
Enclosure:
Inspection Report 99901307/98-01 2
U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION Report No: 99901307/98-01
' Organization: ACCUTECH
Contact:
Steve Gauthier, Quality Assurance Manager (702)739-1966 Nuclear Activity: Manufactures and supplies safety-related fasteners to NRC Licensees Dates: October 26 29,1998 I
1 Inspectors: Bill Rogers, Lead Inspector, HQMB Richard McIntyre, HQMB Donald Naujock, EMCB l
Approved by: Richard Correia, Chief Reliability and Maintenance Section Quality Assurance, Vendor Inspection l and Maintenance Branch Division of Reactor Controls and Human Factors i
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Enclosure l
l 1 INSPECTION
SUMMARY
l On October 26-29,1998, the U.S. Nuclear Regulatory Commission (NRC) performed an inspection at the ACCUTECH facility in Las Vegas, Nevada. The inspection was conducted to review selected portions of ACCUTECH's quality assurance (QA) program, and its implementation, and the applicable programs and procedures used to supply safety-related fasteners to NRC licensees. Specifically, the inspectors reviewed activities related to the corrective actions taken in response to violations (NOVs) and nonconformances (NONs) issued in previous inspections and the processes, procedures, and implementing activities associated with the heat treatments performed in the oven installed in October 1996. The inspectors noted that ACCUTECH had been sold by its previous owner, B&G Manufacturing Co. Inc.,
to Heartland Precision Fasteners Inc., in September 1998.
The inspection bases were:
a 10 CFR Part 50, Appendix B, " Quality Assurance Criteria fc Nuclear Power Plants and Fuel Reprocessing Plants."
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- Reporting of Defects and Noncorr.pliance."
2 STATUS OF PREVIOUS INSPECTION FINDINGS 1 Violation 99901307/96-01-01 (Closed)
Contrary to the requirements of 10 CFR Part 21 (Part 21), ACCUTECH had determined that heat treatment lot TS7 contained defective fasteners but had not )
notified the NRC within two days or provided written notification within thirty days.
ACCUTECH had tested material from lot TS7, had determined that lot TS7 contained defective material, and had documented this conclusion on ACCUTECH l Form 17.002, " Reporting Defects and Noncompliance 10 CFR Part 21," which had !
been approved by the Branch Manager on March 28,1996. However, ACCUTECH l had not notified the NRC by telephone or telefax and ACCUTECH had not provided the written Part 21 notification to the NRC until May 8,1996.
As corrective action, ACCUTECH had initiated Corrective Action Report (CAR)97-004 which required that training be performed on the reporting requirements of 10 CFR Part 21. The inspectors reviewed the training session records and current Part 21 activities and concluded that ACCUTECH's corrective actions had been adequate and closed Violation 99901307/96-01-01.
Violation 99901307/96-01-02 (Closed)
Contrary to the requirements of 10 CFR Part 21, ACCUTECH had not adequately evaluated the materialin lots TS7 and K7 to determine if there were other potentially 2
> L defective lots and whether additional evaluations were required, and had not ;
adequately evaluated available information related to the heat treatment of lot K7 to identify the existence of an additional potentially defective lot (M2)
As corrective action, ACCUTECH had initiated CARS97-002 and 97-005 which required ACCUTECH to provide a Part 21 notification to the NRC for lot M2 and that traMing be performed on the requirements of 10 CFR Part 21 and SOP 17.002. The ;
inspectors reviewed the training session records, the current Part 21 activities, and the March .10,1997, Part 21 notification to the NRC for lot M2.-
8 The inspectors determined that ACCUTECH's corrective actions had been adequate and concluded that Violation 99901307/96-01-02 was closed.
Nonconformance 99901076/94-01-03 (Closed) I Nonconformance 99901076/94-01-03 was originally issued during a 1994 NRC inspection at Cardinal Industrial Products Corporation (CIPC). The corrective actions ;
were reviewed by the inspectors during the November 1996 inspection at !
ACCUTECH (CIPC was renamed ACCUTECH following a change of ownership). 1 The nonconformance remained open following the November 1996 ACCUTECH l inspection and required an additional response from ACCUTF .ti to address the original concerns of Nonconformance 99901076/94-01-03 described below.
Contrary to the requirements of Criterion Vil, " Control of Purchased Material, Equipment and Services," of 10 CFR 50, Appendix B, CIPC/ACCUTECH had not established a documented basis to substantiate that its destructive testing sampling plan for verifying critical characteristics provided reasonable assurance that dedicated commercial grade items met the applicable procurement document requirements.
The May 2,1997, ACCUTECH response had not addressed the sampling rationale provided by CIPC in 1994 -1995 correspondence to the NRC, but instead presented the "ACCUTECH Sample Plan Methodology" (ASPM) document for NRC review and acceptance. The response further stated that the ASPM generally met or exceeded the sampling plan confidence level described in the NRC's draft technical report, ;
" Sampling Plans for Dedicating Simple Metallic Commercial Grade items at Nuclear Power Plants," dated February 1997.
In a December 18,1997, NRC response to the May 2,1997, ACCUTECH letter, the NRC staff stated that it had not performed a technical and acceptance review of the ASPM document. However, the response provided comments on the portions of the ASPM document which had addressed the February 1997 NRC technical report. The December NRC response also stated that further information on the subject of sampling was contained in Draft Regulatory Guide DG-1070,." Sampling Plane Used for Dedicating Simple Metallic Commercial Grade items for Use in Nuclear Power Plants," dated September 1997. DG-1070 was an update to the February 1997 document and was available for public comment until January 30,1998.
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During the current inspection, ACCUTECH indicated that the ACCUTECH ASPM document had been superseded by SOP 7.002, "The Statistical Basis Used for ACCUTECH's Sampling Plans," Revision 1, dated April 20,1998, and SOP 12.001,
" Sampling Plans for Material Verification," Revision 10, dated September 29,1998.
SOP 7.002 described the statistical basis that supported ACCUTECH's sampling i plans for nondestructive dimensional control. SOP 12.001, provided the destructive ;
- testing sampling plans for chemical and mechanical conformance to the material l specifications. The primary basis for the oestructive sampling plans was contali ed in j ASTM F 1470, " Fastener Sampling for Specified Mechanical Properties and ;
Performance inspections." SOP 7.002 stated that the combination of both the ;
nondestructive and destructive sampling plans provided the statistical basis for :
ACCUTECH's overall sampling plans which were stated to provide at least a 95 percent confidence level that 95 percent of the parts would be in compliance. ;
i The inspectors also noted that the NRC was evaluating comments received on DG-1070 and had not issued any additional information or guidance on this sampling issue. In addition, EPRI Report NP-7218, " Utilization of Sampling Plans for CGI .
Acceptance," was currently being revised to address sampling sizes for destructive i testing, consideration of safety function and safety significance when selecting a )
sample size, and lot homogeneity considerations. Until subsequent industry information on sampling or DG-1070 is issued as a Regulatory Guide, HQMB will not ,
comment on specific vendor sampling plans for compliance to DG-1070 unless there ;
arc indications that inadequate material is being provided using the vendors sampling :
plans.
i The inspectors determined that ACCUTECH's corrective actions had been adequate l and concluded that Nonconformance 99901076/94-01-03 was closed. J Nonconformance 99901307/96-01-03 (Closed)
Contrary to the requirements of ASME NCA-3862.1(b) and NCA-3855.3 (b),
ACCUTECH provided Certified Material Test Reports (CMTRs) for ASME code i material supplied to NRC licensees, but did not reference or include the mill heat analysis as an identified attachment.
- The inspectors determined that since January 31,1997, ACCUTECH had been fumishing subcontracted supplier's mill certifications as an identified attachment to :
= the CMTR as required by SOP 6.001, " Control Of Certifications," Revision 1. dated October 21,1998, and that ACCUTECH had only been supplying products with r records which were complete and retrievable, in addition, all CIPC material which i had remained in the ACCUTECH warehouse had been scrapped in October 1998 I (following ACCUTECH's purchase by Heartland Precision Fasteners, Inc in !
September 1998). The inspectors determined that ACCUTECH's corrective actions ,
had been adequate and concluded that Nonconformance 99901307/96-01-03 was closed.
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o Nonconformance 99901307/96-01-04 (Closed)
Contrary to the requirements of Criterion Vil, " Control of Purchased Material,
- Equipment and Services," of 10 CFR 50, Appendix B, and Section 6.1, " Control of i Certifications," of ACCUTECH's Quality Systems Manual (QSM) Second Edition, Revision 0, dated November 29,1998, ACCUTECH's QA Department had not ;
provided adequate verification that material supplied to NRC licensees complied with the licensees' purchase order requirements. Specifically, ACCUTECH hr.d not been ,
- able to demonstrate that suppliers of material had been surveyed and audited to ;
verify their conformance with the above Appendix B requirements. l Since ACCUTECH had performed only limited physical testing of the material, and 7 the sampling plan used for testing the material did not have a documented basis to demonstrate that the material supplied under the purchase orders (POs) was
- equivalent to material purchased from suppliers qualified under Appendix B requirements, ACCUTECH had performed additional testing to the material remaining i
~ in inventory from which the applicable PO items in question had been produced. The ;
inspectors reviewed the results of the additional testing and the documentation that !
was sent to PECO and Wisconsin Electric describing this additional testing. .
. ACCUTECH also indicated that since January 31,1907, ACCUTECH had only been [ '
supplying products with records which were complete and retrievable and, in addition, the CIPC riaterial that had remained in the ACCUTECH warehouse had been scrapped in October 1998.' The inspectors determined that ACCUTECH's corrective ;
actions had been adequate and concluded that Nonconformance 99901307/96-01-04 1
was closed.
Noncor.formance 99901307/96-01-05 (Closed)
Contrary to the requirements of Criterion XVI,
- Corrective Action," of 10 CFR 50, Appendix B, and SOP 17.001, " Corrective Action," Revision 5, dated November 15, 1995, CIPC had failed to identify two nonconformances described in NRC Inspection Report 99901076/94-01 as conditions adverse to quality and to process these nonconformances as CARS in accordance with the corrective action process.
The inspectors verified that SOP 17.001 required the QA Manager to process all issues identified as conditions adverse to quality as part of the corrective action .;
process and to initiate CARS when required, in addition, the inspectors determined i that ACCUTECH had appropriately implemented this requirement by identifying the l violations a:1d nonconformances described in Inspection Report 99901307/96-01 as conditions adverse to quality and processed them as CARS in accordance with the I corrective action process. The inspectors determined that ACCUTECH's corrective ;
actions had been adequate and concluded that Nonconformance 99901307/96-01-05 ]
was closed, j i
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Nonconformance 99901307/96-01-06 (Closed) i Contrary to the requirements of Criterion V, " Instructions, Procedures, and Drawings," i of 10 CFR 50, Appendix B, ACCUTECH QSM, Revision 0, " Interface Activities for !
Material and Associated Documentation Which Has Not Been Legally Transferred l From CardinalIndustrial Products Corporation (CIPC) to ACCUTECH," and SOP I 22.001," Transference of Material and Associated Document Between CIPC and ACCUTECH," Revision 2, dated November 15,1995, no documentation existed to verify that ACCUTECH had implemented SOP 22.001 since July of 1995 and to ensure that CIPC material and documents had been properly reviewed.
The inspectors determined that SOP 22.002, " Stock / inventory Documentation Review Prior to Use," Revision 0, had been issued in March 1997 to ensure that all stock <
material and documentation for CIPC warehoused inventory was properly reviewed I and approved by December 31,1997, that this review had subsequently been ]
completed, and that the CIPC inventory had been segregated in a separate ACCUTECH warehouse. In addition, the inspectors also determined that CIPC inventory in ACCUTECH possession had been scrapped in October 1998 as documented on Nonconformance Reports 14099 and 14557. The inspectors noted )
that SOP 22.002 was no longer an active procedure since all warehoused inventory j reviews had been completed and the CIPC inventory had been scrapped. The '
inspectors determined that ACCUTECH's corrective actions had been adequate and j concluded that Nonconformance 99901307/96-01-06 was closed.
Nonconformance 99901307/96-01-07 (Closed)
Contrary to the requirements of Criterion Vil, " Procedures," of 10 CFR 50, Appendix B, and ACCUTECH SOP 17.002 " Reporting of Defects and ,
Noncompliance," Revision 6, dated November 15,1995, ACCUTECH had not l documented the Part 21 evaluation of lot K7 on Form CF 17.002.
As corrective action, ACCUTECH had initiated Corrective Action Report (CAR)97-010 which required that training be performed on the reporting requirements of 10 CFR Part 21 and the use of ACCUTECH Form 17.002. The inspectors reviewed the training sessions records and current Part 21 activities and concluded that ACCUTECH's corrective actions were adequate and concluded that Nonconformance 99901307/96-01-07 was closed.
3 INSPECTION FINDINGS AND OTHER COMMENTS 3.1 Heat Treatment Processes. Procedures and lmolementation
- a. Insoection Scooe During this inspection, the NRC inspectors reviewed selected fumace procedures and heat treatment documents and observed a furnace demonstration. The inspection 6
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scope included ACCUTECH's Reheat Furnace No. 4 (RF4), which had started operation in October 1996.
- b. Observations and Findinos RF4 Furnace Descriotion ACCUTECH personnel described the RF4 as a double chamber heat and quench vacuum furnace manufactured by C.I. Hayes inc, Type VSQ. The front chamber contained an oit quench tank and the back chamber an insulated heating box with three top and three bottom heating elements. The chambers were connected with rails that supported a removable 45" wide by 38"long charge table. When loaded, the operator manually pushed the charge table into the heating box and after heat treating the table automatically withdrew to a position above the quench tank. The operator could automatically or manually release the load from the table into the quench tank.
The temperature in the heating box was monitored by three thermocouples (TC). The furnace TC (FCE TC) was a calibrated type S thermocouple connected to a calibrated Honeywell DCP 700 programmable temperature controller that controls the temperature inside the heating box. The FCE TC was located in the center of one side of the box and extends severalinches into the heating volume. Two inches above the FCE TC was an over temperature Type S thermocouple (OT TC) connected to a calibrated Honeywell DL200H. The DL200H controller automatically shut off power to the heating elements when the temperature in the box reached a i predetermined set point. The temperature in the furnace load was monitored with a j calibrated type K thermocouple (Load TC) placed in the coldest location within the '
load. The inspector noted that during operation the Load TC temperature would lag the FCE TC temperature, and that the FCE TC temperature would be less than the OT TC temperature. Both Load TC and FCE TC temperatures were recorded on a ;
calibrated Honeywell DPR3000, strip chart recorder. l ACCUTECH personnel indicated that the vacuum pump was capable of lowering furnace pressure to a range of 60 to 200 microns. The furnace vacuum was monitored with a Televac il vacuum controller which would automatically disconnect power to the heating elements whenever the furnace vacuum exceeds 650 microns. i The function of the vacuum was to impede convection heat transfer and to minimize l scale formsilon on heat treated parts. Heat transfer between the chambers must be controllod to minimizing thermal effects on the oil in the quench tank. The operator stated that if the door on the heating box was left ajar during the heating cycle, radiant heat escaping from the box would raise the oil's surface temperature which, in turn, would increase oil vapor pressure above the 650 micron shut off set point.
Control of the Heat Treat Process The inspectors reviewed ACCUTECH's heat treatment procedures, SOP 18.001,
" Standard Operating Procedure for Operation of the Vacuum Fumace," Revision 9, 7
dated March 6,1998, and SOP 18.002, " Heat Treatment Equipment Calibration and Survey," Revision 7, dated March 6,1998, and selected process sheets. SOP 18.001 provided instructions to the operator on the type of information required for various furnace documents, the sampling for go-no go hardness tests, and general operation
, instructions. SOP 18.001 referenced other sources for more specific instructions, such as SOP 18.002 for calibrations and specification specific process sheets that provided soak time, soak temperature, and change in temperature requirements for different materials and part sizes. SOP 18.001 and associated referenced documents did not provide instructions for load distribution and enarge weight. ;
P SOP 18.002 provided instructions for verifying the accuracy of equipment used for centrolling and recording furnace temperatures and for profiling temperature distribution inside the heating box. The inspectors reviewed the surveys ACCUTECH had used to profile furnace temperature distribution at various soak temperatures.
These surveys measured soak temperatures at nine locations inside the heating box under simulated operating conditions. ACCUTECH had defined simulated operating conditions as using two baskets, one on top of the other, with each basket containing approximately 225 pounds of four inch long tubes stacked two inches deep. The last set of surveys had been performed on May 7,1998, at soak temperatures of 800'F, 1000*F,1400*F, and 1900*F. These surveys showed a maximum temperature difference of 28'F at a soak temperature of 800*F and 20*F at a soak temperature of 1900'F. The 800'F and 1900*F surveys identified the coldest spot in the heating box as being in the lower back, thermocouple side of the box. The cold spots at 800*F and 1900*F were 793*F and 1895*F, respectively.
ACCUTECH personnel informed the inspectors that production loads processed through RF4 are processed with load weights and distributions similar to those used for the surveys. For example, if the operator estimated that the load weight for two baskets would exceed approximately 450 pounds, the operato splits the load into two lighter loads and added filler material to each basket until the load weight was ;
approximately 450 pounds for each load.
Heat Treat Demonstration The inspectors observed a furnace demonstration on age hardening of four 2.0"-
BUNC heavy hex nuts made from ASME SB647 718 material that were being processed under Work Order (WO)100284. The fumace operator followed SOP 18.001 for general fumace operations and testings and specification specific
" Process Sheet AA" for heating parameters. The operator filled the charge baskets with filler material to simulate the load weight and distributions used in the May 7,1998, fumace surveys. The four nuts and Load TC were placed in the lower basket at the same location that coincided with the cold spot in the heating box. The baskets were stacked two high on the charge table and pushed into the heating box.
After closing the heating box door and checking it for tightness, the fumace door was closed and vacuum initiated. While the fumace vacuumized, the operator entered ,
heating parameters into the fumace controllers and completed the logs and documents required by SOP 18.001. The heating parameters consisted of an eight 8
I hour soak at 1325'F followed by an 87.5*F per hour cool-down for two hours and soaked for an additional eight hours at 1150'F followed by a nitrogen gas cool-down to below 150*F. The operator removed the nuts from the furnace and performed go-no go hardness tests which were acceptable (44Rc).
Heat Treatment Documentation The inspector reviewed heat treatment documentation from fifteen work orders that had been processed through RF4. On five work orders, the inspector noticed that the Load TC temperatures exceeded the FCE TC soak temperatures by 5'F,10'F, 25'F,35"F, and approximately 100*F. ACCUTECH had issued Nonconformance 14339 when the Load TC was approximately 100 F above the FCE TC soak temperature. Neither SOP 18.001 or the specification specific process sheets provided guidance on this anomaly and ACCUTECH depended upon the operator's knowledge to identify nonconformances based on high Load TC temperatures.
Although no unacceptable material was identified, the inspector informed ACCUTECH's personnel that the Load TC temperatures, which exceeded the FCE TC temperatures, were near the load material's transition temperature which could produce rejectable material in the load. The inspector also determined that seven of fifteen work orders had their loads split at RF4 (for weight and distribution) which was in agreement with ACCUTECH's statement that charge load weights and distributions are maintained in a similar ccnfiguration to those used during the calibration surveys in SOP 18.002.
- c. Conclusion The inspectors noted that although an anomaly had occurred during subcritical anneal and temper soak temperatures, no unacceptable material had been identified.
The inspectors concluded that the implementation of furnace procedures and furnace operator skill appeared to produce satisfactory products with RF4.
3.1 Review of Wisconsin Electric May 4-8.1998. Audit (for NUPIC members) of ACCUTECB
- a. Insoection scone The inspectors reviewed the May 4-8,1998, Wisconsin Electric joint utility Audit Report 98-019 (Nuclear Utility Procurement Issues Committee audP 4 OC)) of ACCUTECH to assess ACCUTECH's corrective actions process.
- b. Observations and Findinas As part of the audit, the NUPIC team had reviewed ACCUTECH's corrective actions related to the past NRC inspection report findings and verified that corrective actions had been taken for all NRC findings. As a follow up to past dedication issues at ACCUTECH, commercial grade dedication sampling was reviewed. The NUPIC audit report concluded that the ACCUTECH sampling plan for dimensionalinspection of suppliers which were not on the approved vendor list (AVL) compared favorably with 9
I NRC draft Regulatory Guide DG-1070. It further stated that the plan was designed to ultimately obtain the 95/5 confidence level referenced in DG-1070 and in some cases the sample size was greater than required by DG-1070 while in other cases, less items were sampled. The audit report stated that this appeared to be a result of the ACCUTECH sampling plan's lower number of rejects allowed while still accepting the lot.
The audit report also stated that the sampling plan for material and physical testing was less conservative than DG-1070 and takes credit for visual and dimensional inspections for ascertaining lot homogeneity. The testing sample size was based on the requirements of ASTM F1470 and F a more conservative approach was desired, additional testing at the customer's facility may be needed. The NRC inspectors did not see any proceduralindication that ACCUTECH was still relying on visual and dimensional inspections for ascertaining lot homogeneity as stated in the Wisconsin Electric audit report and ACCUTECH confirmed that this was not the case. Finally, the audit report correctly described that latest correspondence between the NRC and ACCUTECH concerning the ACCUTECH sampling plan basis and identified the NRC comments with respect to the DG-1070 document and the ACCUTECH sample plan methodology.
In addition, the NUPIC audit report stated that "ACCUTECH still considers it Cardinal's responsibility to notify customers of problems with items bought ano manufactured under Cardinal ownership," and that "they (ACCUTECH) will make a
" courtesy" notification to the affected customers." This position was not in agreement with the May 29,1997, NRC letter to ACCUTECH which stated that the NRC staff had j concluded that ACCUTECH was incorrect in the interpretation of its responsibilities I under Part 21 by characterizing them as " courtesy" actions; the Part 21 regulations require that the pertinent provisions of the regulations be implemented in the event of the discovery of a defect in a basic component; and that ACCUTECH was bound to take formal action pursuant to Part 21 and must do so because of the regulatory requirements it assumed due to its purchase of the CIP [ Cardinal) assets. Discussion with ACCUTECH management during the NRC inspection indicated that ACCUTECH l was currently in agreement with the NRC position on Pari 21 responsibility and did not agree with the position stated in the NUPIC audit report.
c. Conclusion
The inspectors concluded that NilPIC audit report 98-019, which documented the audit led by Wisconsin Electric, had mischaracterized two ACCUTECH positions related to sampling and Part 21 responsibility. However, NRC discussions with ACCUTECH management indicated that ACCUTECH's actual position on Part 21 was in agreement with NRC expectations and was acceptable. The inspectors also l
determined that ACCUTECH's position on sampling and verification of lot homogeneity was not as described in the NUPlC audit. (ACCUTECH's sampling ,
process and the NRC's position on sampling is described in Section 2 of this inspection report in the portion titled "Nonconformance 99901076/94-01-03 (Closed).")
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4 PERSONS CONTACTED l 1
.l David Rose, President Joseph Casey, General Manager Steven Gauthier, Quality Assurance Manager !
Joseph Lee Conrad, Quality Assurance Supervisor j 4
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ITEMS OPENED, CLOSED, AND DISCUSSED ftem Number Iygg Descriotions Closed 99901307/96-01-01 NOV Exceeding Part 21 Notification Timeliness Limit 99901307/96-01-02 NOV Inadequate Part 21 Evaluation -
99901076/94-01-03 NON Inadequate Documented Basis for ,
Destructiva Sampling Plan i 99901307/96-01-03 NON Incomplete Documentation Supplied to Customer 99901307/96-01-04 NON Inadequate Verification of Conformance 1
99901307/96-01-05 NON inadequate Corrective Action 99901307/96-01-06 NON Inadequate Review of Materialin Warehouse Stock 1 99901307/96-01-07 NON Inadequate Documentation of Part 21 Evaluation 1
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sP "fcw y 4 UNITED STATES
& E NUCLEAR REGULATORY COMMISSION E
'E WASHINGTON, D.C. 20666 4 001
% . ,,, $ December 22, 1998 Mr. Earl M. Pollock, President Nuclear Research Corporation 125 Titus Avenue, P.O. Box H Warrington, PA 18976
SUBJECT:
NRC INSPECTION REPORT NO. 99901336/96-01
Dear Mr. Pollock:
On November 23-25,1998, the staff of the U.S. Nuclear Regulatory Commission (USNRC) performed an inspection of the activities performed at the Warrington, Pennsylvania, facility of Nuclear Research Corporation (NRC). The enclosed report presents the results of that inspection.
The primary purpose of the inspection was to determine the extent of testing employed in the qualification of the ADM-606/616 series digital ratemeters and if such testing was accomplished in accordance with applicable regulatory requirements. During the inspection, the team specifically reviewed activities related to prototype testing, internal and external quality audits, conformance with licensee purchase order requirements for safety-related equipment, and overall compliance with USNRC requirements. During the inspection, the team did not identify any instance where NRC failed to meet such requirements.
In accordance with 10 CFR 2.790, of the USNRC's " Rules of Practice,' a copy of this letter and !
its enclosure will be placed in the USNRC's Public Document Room, Should you have any I questions concerning this inspection, we will be pleased to discuss them with you.
Sincerely,
( k >
Suzanne C. Black, Chief Quality Assurance, Vendor Inspection, and Maintenance Branch Division of Reactor Controls and Human Factors Office of Nuclear Reactor Regulation
Enclosure:
Inspection Report No. 99901336/98-01 U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION Report No.: 99901336/98-01 Organization: Nuclear Research Corporation 125 Titus Avenue Warrington, Pennsylvania
Contact:
Steven O'Brien Quality Assurance Manager Nuclear Activity: Provides radiation monitoring equipment to the nuclear industry.
Dates: November 23-25,1998 1
l inspection Team: Robert L. Pettis, Jr., HOMB/DRCH i Kenneth Heck, HOMB/DRCH Paul Loeser, HICB/DRCH Approved by: Robert A. Gramm, Chief Quality Assurance and Safety Assessment Section Quality Assurance, Vendor Inspection, and Maintenance Branch Division of Reactor Controls and Human Factors Office of Nuclear Reactor Regulation i
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Enclosure 1 INSPECTION
SUMMARY
From November 23-25,1998, representatives of the U.S. Nuclear Regulatory Commission (USNRC) conducted a performance-based inspection of the activities at the Warrington, Pennsylvania facility of the Nuclear Research Corporation (NRC). In conducting this inspection, the team emphasized technically directed observations and evaluations of NRC activities related to the manufacture and testing of nuclear safety-related radiation monitors used throughout the nuclear industry. As the technical bases for the inspect'on, the team relied upon the following:
Part 21, " Notification of Failure to Comply or Existence of a Defect," as defined in Title 10 of the Code of FederalRegulations (10 CFR)
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10 CFR Part 50, Appendix B, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants" NRC Quality Assurance Manual, Revision 0, approved March 22,1989.
2 STATUS OF PREVIOUS INSPECTION FINDINGS This was the first USNRC inspection performed at NRC.
3 INSPECTION FINDINGS AND OTHER COMMENTS 3.1 Backaround General Design Criteria 4 of the Code of FederalRegulations, Title 10, Part 50, Appendix A requires that instrumentation and control (l&C) equipment be compatible with the environmental conditions it could experience. To meet this criterion, I&C systems in nuclear power plants must withstand electromagnetic interference (EMI) and radio frequency interference (RFI) and power surges from external and internal devices such as transformers, solenoid coils, and relays, electrical surges induced by lightning and power switching transients, and radio frequency disturbances. Safety-critical protection and control systems must be appropriately hardened (e.g., shielded) against such interference. Therefore the expected environment in which these items must operate must be properly characterized to allow the electromagnetic immunity of I&C equipment to be assessed against appropriate levels of interference.
The Electric Power Research Institute (EPRI) has characterized the emission levels and expected types of interference at nuclear power plants. Based on these surveys, EPRI has developed guioelines for equipment susceotibility tests (EPRI Topical Report TR-102323, September 1904). The recommend a tests are included in standards defined by the military and commercial sectors and the levels are conservative based on the analyzed data. The USNRC staff has reviewed TR-10323 and has concluded that the EPRI guidelines provide an adequate method for qualifying digita l&C equipment for a nuclear plant's electromagnetic environment without the need for plant specific EMI surveys if the plant-specific electromagnetic environment is confirmed to be similar to 1
that identified in TR-10323. Nuclear utilities, operating under USNRC licenses, are ultimately responsible for the safety of equipment installed in their nuclear power plants.
Licensees generally limit the vendors from they purchase to those complying with regulatory requirements and periodically audit these vendors to ensure continuing compliance.
NRC maintains a quality assurance program in compliance with 10CFR50, Appendix B, 10CFR21 and American National Standards institute (ANSI) N45.2. The adequacy and effectiveness of NRC's quality assurance program is audited through the Joint Audit Program facilitated by the Nuclear Procurement issues Committee (NUPIC), whose membership represents operating nuclear power plants in the United States. NUPIC audits of NRC's quaiity program were conducted in 1995 and 1997; a follow-up audit is '
scheduled for December 1998.
l Nuclear utilities generally specify, as part of the contract, the specific EMI criteria and I standards to which purchased l&C equipment shall be qualified. NRC's customers may ,
elect to contract electromagnetic compatibility (EMC) testing to an independent I laboratory or through NRC, who subcontracts this testing to Radiation Sc!ences incorporated (RSI). NRC conducts full surveillance of all RSI testing and has performed ,
audits of RS!'s quality system in 1992 and 1998. NRC determined RSl's testing I activities to be in conformance with applicable NRC test procedures and quality I requirements.
The USNRC inspectors examined several of these procedures and verified that they specified EPRI TR-10323 guidelines and other applicable military and ccmmercial standards for EMI testing. In addition, RSI test reports were reviewed with respect to implementation of NRC test requirements and the adequacy of the RSI test report with regard to documentation of test environment, test samples, test equipment, and the l qualitative reporting of test results.
3.2 Review of EMC Testino of ADM-600 Series Radiation Monitorina Eauioment
- a. Insoection Scope 1 The inspection team reviewed NRC's EMC testing program for the ADM-600 series radiation monitoring equipment, including identification of EMC testing requirements and evaluation of complementary test records for conformance to 10CFR50, Appendix B.
NRC's ADM series is a microprocessor-based multifunction ratemeter which can accept a variety of inputs from different style detectors for configuration as an area or process monitor. Using a digital readout as well as an analog display for trending purposes, the ADM series of radiation detectors are used with a variety of probes for monitoring alpha, beta, gamma, X-ray, and neutron radiation.
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- b. Observations and Findinas b.1 EPRI Guidelines for EMI Testina The inspection team reviewed EPRI Topical Report TR 102323, " Guidelines for Electromagnetic Interference Testing in Power Plants," dated September 1994, which addresses the potential effects of electromagnetic emissions on digital equipment operation. Based on surveys at commercial nuclear plants, the report characterizes the environment in which digital equipment is expected to operate. The report recommends appropriate equipment testing standards such as Military Standard (MIL-STO) 641C and MIL-STD-641D for conduction and radiation susceptibility testing, in a safety evaluation of TR-102323, dated January 30,1996, the staff of the USNRC concluded that the EPRI guidelines provide an adequate method for qualifying digital instrumentation and control equipment for a plant's electromagnetic environment without
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I the need for plant specific EMI surveys if the plant-specific electromagnetic environment I is confirmed to be similar to that identified in TR-10323.
b.2 EMI Testina Performed by Radiation Sciences incoroorated Nuclear utilities gent. ally specify, as part of the contract, the specific EMI criteria for which instrumentation and control equipment shall be qualified. Nuclear licensees may elect to contract EMI/RFI testing to an independent laboratory or through NRC who subcontracts all EMI/RFI testing to RSI. Because RS!is not an 10 CFR 50 Appendix B service provider and does not accept 10 CFR Part 21 when imposed by purchase order, NRC assumes the responsibility for ensuring that RSI testing conforms to regulatory requirements. A 1997 NUPIC audit of NRC identified that applicable RSI activities such as control of calibrated test equipment and qualification of test personnel had not been reviewed by NRC during its audits of RSI. As a result of corrective action to address the findings, NRC performed an audit of RSI on January 28,1998.
The USNRC inspection team reviewed documentation provided by NRC that demonstrated the results of the RSI audit had been reviewed and accepted by NUPlC.
All RSI testing contracted by NRC is performed in accordance with approved test procedures and all testing is witnessed by NRC technical personnel.
b.3 Review of EMI Testina for ADM-600 Series Diaital Ratemeters The team selected several radiation monitors supplied by NRC to nuclear power plants for the purpose of reviewing EMI test results. The test reports were reviewed with respect to implementation of EMI requirements and the adequacy of the test report in ,
documenting general test requirements, test environment, test sample identification,' test ;
equipment calibration, and qualitative reporting of the test results. I 3
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. ADM-600/610 The ADM-600/610 series, developed in 1991, is an evolutionary design of ratemeters developed by NRC in 1975. In this design, processing can be distributed between two processors, one in the control room (rack installed ADM-600) and one in a location closer to the detector (ADM-610 in a NEMA-12 wall-mounted enclosure). In 1993, development began on the design of the ADM 606/616 series which has the same basic architecture as the ADM-600/610 but uses a 16-bit microprocessor.
The team reviewed EMl/RFI testing conducted by RSI in March 1992. The following tests were conducted in accordance with NRC Test Plan Document No. 200888.
. EMI Transient Susceptibility Conducted EMI/RFI Susceptibility Radiated EMI/RFl Susceptibility Hand Held Radio EMI/RFI Susceptibility The referenced test methodology used was MIL-STD-461C and the specific test conditions were detailed in the report prepared by RSI. The team's review did not identify any concerns.
- ADM-600A !
The team reviewed documentation of an ADM-600A digital ratemeter supplied to l PECO Energy Company as part of a control room ventilation radiation monitoring system. EMI/EMC testing was conducted by NTS in 1995 under contract to PECO and portions of the test (IEC 801-4 and IEC 804 5) were subcontracted to Chromeries Test Services under the oversight of NTS test personnel. The following tests were conducted under NTS Test Plan No. 60541-95N:
. Conducted Susceptibility, Method CS01,30 Hz to 50 KHz Conducted Susceptibility, Method CS02,50kHz to 400 MHZ
. Conducted Susceptibility, Method CS06, Spikes,
. Radiated Susceptibility, Method RS03, Electric Field,10 kHz to 1 GHz
. Electrical Fast Transients / Burst immunity, IEC 801-4
. Surge immunity, IEC 801-5 j i
The NTS report references the test methodology recommended by EPRI TR- l 102323. The team's review did not identify any concerns. l l
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. ADM-606 The inspection team reviewed the following documents associated with testing of the ADM-606 model digital ratemeter:
. Electromagnetic Interference / Compatibility Test Report, RSI-4521E/A2, prepared for Nuclear Research Corporation for the ADM-606/616 series Digital Ratemeters, dated August 1996, submitted by RSI.
. Electromagnetic Compatibility Testing for ADM-606 (V2), performed at RSI, Test Report RSI-5031E, dated August 17,1998.
. Nuclear Research Corporation Standard Test Plan for Measurement of Electromagnetic Interference for the Main Steam Line / N-16 Monitor System, NLM-100, TP-201462, dated March 27,1996.
. Nuclear Research Corporation Pre / Post EMI Test Procedure for Main Steam Line / 416 Monitor System, NLM-100, TP-201461, dated March 27,1996.
The documents identified two tests (1996 and 1998) performed for the ADM-606 digital ratemeter, The tests were a combinatiori of the M!L-STD-462C susceptibility tests, as specified in EPRI Topical Report 102323, dated September 1994, IEC 801 i series electrostatic discharge, fast transient and surge tests, and emissions tests as I specified in Federal Communication Commission (FCC) Regulation 47 CFR Part 15.
The following is a listing of specific tests performed by NRC:
From EPRI 102323: CS01, Conducted Susceptibility, Power Lead,30Hz to 50 KHz; CS02, Conducted Susceptibility, Power Lead,50 KHz to 400 MHZ; CS06, Conducted Susceptibility, Power Lead, Spike; and RS03, Radiated Susceptibility, Electric Field,10kHz to 1 GHz.
. From IEC 801: 801-2, Electrostatic Discharge, Contact & Air; 801-4, Electrical Fast Transient immunity; and 801-5, Surge immunity.
. From FCC Method FCC/B, Conducted Emissions and Method FCC/C, Radiated Emissions.
Before and after each test, the functionality of the unit was verified using NRC Pre / Post Test Procedure TP-201461. Each test was performed using NRC Standard Test Plan TP-201462. During the 1996 tests, the ADM-606/616 units passed the EPRI and the IEC tests, but the FCC tests showed a level of electromagnetic radiation higher than allowed by FCC Part 15 requirements. The design of the digital ratemeters was modified by adding EMI gaskets to the cover plate, adding a grounded conductive film to the display window, and adding
! grounded shields to all unused connectors at the back of the unit.
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in 1998, after these modifications were made, the units passed the tests. There was no attempt made to test or verify EPRI 102323 emissions level requirements as specified in Table B-1 and NRC did not claim to have met these emissions levels.
The team's review did not identify any concerns.
ADM-606M (V2)
EMI/EMC testing was conducted by RSI in August,1998, under contract to NRC.
The purpose of the tests were to qualify the ADM-606M (V2) to the requirements of the European Common Market based on similarities shared with the ADM-606M.
The RSI test report documented successful completion of the following tests:
. FCC Part 15 Class A Emission
- EN55011 Class A Emission
. EN61000-4-5 Surge (Criteria A 2 KV)
. EN61000-4-6 Conducted Immunity (Criteria A 10V)
EN61000-4-11 Voltage Variations (Criteria A Parts 1 & 2; Criteria B Part 3)
Review of ADM 600/616 Diaital Ratemeters for Baltimore Gas & Electric in addition to the team's overall review of the ADM-600 series digital ratemeters, the team selected for review ADM-600/616 digital ratemeters which were supplied to Baltimore Gas & Electric (BG&E) as part of a safety-related main steam line :
radiation monitoring system,.which includes functionality for monitoring N-16 )
effluent. The BG&E design specification reads as follows: )
The system shall meet the surge withstand requirements and the high frequency conducted transient requirements defined in EPRI TR-10323; and a requirement that all power and signal input lines be tested. i l
The entire radiation monitoring system, except for any electromagnetic storage i devices shall meet the radiated EMI susceptibility requirements defined in j EPRI TR-10323. All electromagnetic storage devices shall meet IEC 801-3 i with a 3V/m field from 27 to 1000 Hz.
An estimate of the emitted EMI shall be furnished from the supplier for the new system. This information r> hall be based on previously constructed system, I with similar configuration. In addition, the supplier shall certify that the emitted EMI meets the requirements of FCC Regulations 47 CFR Part 15 for Class B devices. ,
A critical onsite design review of the radiation monitoring cystems purchased under i this contract was conducted by B'3&E and Data Refining Technologies, Incorporated !
in July 1996. The report which documented this review concluded that the application and the prccess under which the radiation monitoring system wse developed satisfy the criteria set forth in IEEE/ANS 7.4.3.2-1993, " Standard Criteria 6
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for Digital Computers in Safety Systems of Nuclear Power Generating Stations." l NRC Test Plan No. TP-201462, which implements the BG&E test requirements, '
states "The following tests are in accordance with MIL-STD 461C, dated August i 1986, and MIL-STD 462, dated July 1967, as described in EPRI TR-102323, Appendix B, Paragraph 3.0:
. CS01, Conducted Susceptibility, 30 Hz - 50kHz.
. CS02, Conducted Susceptibility, 50 KHz - 00MHz.
. CS06, Conducted Susceptibility, Power Leads, Spikes.
. RS03, Radiated Susceptibility,10 KHz - 1GHz.
. IEC Publication 801-4, Electrical Fast Transient / Burst Immunity (1988). l
. IEC Publication 801-5, Surge immunity (1990). l l
. IEC Publication 801-2 for tabletop equipment using the following test levels:
. Air Discharge - 8 KV
. Contact Discharge - 4 KV l
. FCC Regulation 47 Part 15 for Class B devices. ;
. Radiation Emissions for Class B equipment EMI/EMC testing was conducted by RSI in August,1996 in accordance with the above test plan. In addition to these tests, BG&E witnessed and accepted comprehensive EMI/RFI tests on the skid-mounted units as part of factory accepta.1ce testing.
- c. Conclusions Based on review of documentation related to the tests described above, the inspection team concluded that the RSI tests implement the specifications of the NRC test plans; the test methodology, equipment, and results are well documented, and the test reports are signed by the appropriate level of RSI management. The inspection team also i concluded that the ADM-606/616 series digital ratemeters meet the applicable I susceptibility requirements as specified in EPRI-102323 and emissions requirements as ;
specified in FCC Part 15 Methods E and C.
4 ENTRANCE AND EXIT MEETINGS During the entrance meeting on November 23,1998, the USNRC team met with members of NRC management and staff and discussed the scope of the inspection. The team also reviewed its responsibilities for handling proprietary information as well as those of NRC. In addition, the team established contact persons within the management and staff of the applicable NRC organizations. The team discussed the results of the inspection with NRC management and staff on November 25,1998.
7 5 PARTIAL LIST OF PERSONS CONTACTED E. Pollock - President, NRC S. Panday Ph.D. V.P. Engineering J.~ Tomei - Engineering
-J. Gerfin Engineering S. O'Brien QA Manager l R. Letizia Engineering l
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I ps.8 40uq k UNITED STATES l 3
g NUCLEAR REOULATORY COMMISSION l o ,g WASHINGTON, D.C. 20066 4001 l
+4 0*
.,,,4 October 21, 1998 Mr. Jim Shieh, Quality Assurance Director Valcor Engineering Corporation 2 Lawrence Road Springfield, New Jersey 07081
SUBJECT:
NRC INSPECTION REPORT 99900728/98-01 AND NOTICE OF i NONCONFORMANCE j
Dear Mr. Shieh:
On September 8-11 and October 2,1998, the U.S. Nuclear Regulatory Commission (NRC) performed an inspection at the Valcor Engineering Corporation (Valcor) facility in Springfield, New Jersey. The enclosed report presents the findings of that inspection. The inspection was conducted to review selected portions of your quality assurance program, and its implementation, as it relates to the supply of solenoid operated valves to the nuclear industry.
This inspection specifically reviewed activities related to Valcor's review and analysis of valve failures at the Pennsylvania Power and Light Susquehanna Steam Electric Station (SSES), and ;
your evaluation performed in accordance with 10 CFR Part 21, including the susceptibility of I similarly designed valves.
During this inspection, the inspectors found the implementation of your quality assurance program failed to meet certain NRC requirements. Specifically, the inspection identifed that Valcor failed to prescribe written guidance within its procedures and associated drawings to control the application of a dry film graphite lubricant on its V70900-65-11 series solenoid operated valves.
This nonconformance is cited in the enclosed Notice of Nonconformance (NON), and the circumstances surrounding it are described in detail in the enclosed report. During the October 2,1998, portion of the inspection, the inspectors reviewed your corrective actions and steps taken to preclude recurrence and found them acceptable. Therefore, no further resoonse to the NON is required.
In accordance with 10 CFR 2.790 of ttdiffiC's " Rules of Practice," a copy of this letter and its enclosures will be placed in the NRC's Public Document Room.
Sincerely, GW W Suza C. Black, Chief Quality Assurance, Vendor inspection and Maintenance Branch Division of Reactor Controls and Human Factors Office of Nuclear Reactor Regulation Docket No. 99900728
Enclosures:
- 1. Notice of Nonconformance
- 2. Inspection Report 99900728/98-01 1
NOTICE OF NONCONFORMANCE Valcor Engineering Corporation Docket 99900728 Springfield, New Jersey Based on the results of an inspection conducted on September 8-11, and October 2,1998, it appears Stat certain of the Valcor Engineering Corporation (Valcor) activities were not conducted in accordance with NRC requirements.
Criterion V, " Instructions, Procedures, and Drawings," of 10 CFR Part 50, Appendix B requires ;
that activities affecting quality be prescribed by documented instructions and procedures and ;
shall be accomplished in accordance with those instrudions and procedures.
Contrary to Critericn V, Valcor failed to prescribe written guidance within its procedures and associated drawings 'o control the application of Acheson Colloids Company (Acheson) dry film graphite lubricant (Dat$ 156) on its V70900-65-11 series solenoid operated valves. i Additionally, Valcor fa: led to addrers a Dag@ 156 technical information sheet note that expressed a five mirate drying tirae caution before assembling components. (Nonconformance 99900728/98-01-01)
During the October 2,1998, portion of the inspection, the inspectors reviewed your corrective actions and steps taken to preclude recurrence and found them acceptable. Therefore, no ,
further response to the NON is required.
Dated at Rockville, Maryland this O l # day of October 1998 p
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s Enclosure 1
U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION Report No: 99900728/98-01 Organization: Valcor Engineering Corporation
Contact:
Jim Shieh, Quality Assurance Director (973)467-8400 Nuclear Activity: Manufacturer and supplier of flow control devices, including safety-related solenoid operated valves and replacement parts used in nuclear applications.
Dates: September 8-10 and October 2,1998 Inspectors: Gregory C. Cwalina, Senior Operations Engineer i Joseph J. Petrosino, Quality Assurance Specialist Harold L. Ornstein, Senior Reactor Systems Engineer 1
l Approved by: Robert A. Gramm, Chief I Quality Assurance and Safety Assessment Section Quality Assurance, Vendor inspection and Maintenance Branch Division of Reactor Controls and Human Factors l
l Enclosure 2 1 INSPECTION
SUMMARY
On September 8-10 and October 2,1998, the Nuclear Regulatory Commission (NRC) performed an inspection of the Valcor Engineering Corporation (Valcor). The inspection reviewed selected portions of the Valcor quality assurance program, and its implementation, as it relates :.o the supply of solenoid operated valves to the nuclear industry. Specifically, the inspection reviewed activities related to Valcor's review and analysis of reported valve failures at the Pennsylvania Power and Light Company's (PP&L) Susquehanna Steam Electric Station (SSES), and Valcor's evaluation performed in accordance with 10 CFR Part 21, including the susceptibility of similarly i des:gned valves.
The inspection bases were: ;
- Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocess!ng Plants." ;
. 10 CFR Part 21, " Reporting of Defects and Noncompliance."
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During this inspection, one nonconformance was identified and is discussed in Section 3.6 of this report. l 2 STATUS OF PREVIOUS INSPECTION FINDINGS I Violation 99900728/91-01-01(Closed) l During a February 1991 inspection of Valcor, the inspectors found that Valcor failed to l adopt appropriate procedures to provide for evaluating deviations or informing licensees or purchasers of the deviations. .
During this inspection, the inspectors reviewed Valcor's Part 21 implem9nting procedures (see Section 3.3) and found them to meet the requirements of 10 CFR Part 21 with regard to customer notifications.
Violation 99900728/91-01-02 (Closed)
During a February 1991 inspection of Valcor, the inspectors found that Valcor failed to ,
post copies of Section 206 of the Energy Reorganization Act of 1974 and failed to post
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its 10 CFR Part 21 procedure in a conspicuous location. '
During this inspection, the inspectors observed that the documents which Valcor had displayed were conspicuously located and met the requirements of $21.6, " Posting requirements," of 10 CFR Part 21.
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l l- 3 INSPECTION FINDINGS AND OTHER COMMENTS L
l 3.1 Backaround On September 20,1997, the Unit 2 Reactor Recirculation Pump Cooling Water inboard containment isolation valve failed its stroke time surveillance at the Pennsylvania Power and Light Company's (PP&L) Susquehanna Steam Electric Station (SSES) due to a reported failure of the associated solenoid operated valve (SOV). PP&L replaced the SOV, Valcor model V70900-65-11, and successfully stroked the valve. PP&L also ,
L initiated a Condition Report (CR-3173) to investigate the cause of the apparent failure of I the Valcor SOV (serial number (S/N) 43). ,
Valcor SOV S/N 43 had been in operatior, and continuously energized since April 1997. :
l Following removal from service, PP&L was unable to duplicate the apparent failure.
- Therefore, the SOV was returned to Valcor for their examination. Valcor reported that allinternal components were in satisfactory condition and within tolerances. Stroke '
testing was performed successfully, and the stroke and gap of the valve were dimensionally within their acceptable range. t PP&L investigated several proposed causal factors, including misinsta!Iation, misalignment and the possibility of internal binding due to manufacturing defects or the l presence of adhesive particles from installation The PP&L root cause analysis was reviewed by an independent testing laboratory, Altran Corporation, who also performed laboratory investigations and testing of the Valcor SOV. On March 20,1998, PP&L and '
Altran representatives met with Valcor to present the results of their root cause analysis.
l Following the meeting with PP&L and Altran, on March 25,1998, Valcor issued Nuclear Field Report (NFR) #017. This began Valcor's evaluation in accordance with 10 CFR Part 21 requirements. Valcor issued a Part 21 notification to the NRC on May 19,1998 (see Section 3.4).
l Altran issued Revision 0 of Technical Report No. 98115-TR-01, "RBCW Containment Isolation Valve HV-28792A2 Stroke Time Surveillance Failure Root Cause investigation," in April 1998. Revision 0 concluded that the most probable root cause for the failure was, " residual magnetism which existed in the Valcor Pilot SOV S/N 43 l following de-energization." l 1
The PP&L final root cause analysis identified several design deficiencies associated with the valves, as well as some internal PP&L issues. PP&L, in conjunction with Altran, concluded that the root cause of the valve failure was. " Defective Design with High
' Inherent Residual Magnetism, Unknown to the Vendor.. " Four contributing factors were also identified; 1- Inadequate control of the gap between the plunger and backstop l
- . 2- . Less than adequate qualification testing of a new design l
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I 3- Normal in-service creep deformation (of the backstop o-ring) 4- Particulate debris in the small clearance areas of the valve Based upon their evaluation and the independent root cause analysis performed by Altran, PP&L concluded' that all 48 Valcor solenoids, Models V70900-65-11 and 65-12, were defective due to a loss of o-ring resiliency which was necessary to overcome the effects of residual magnetism. As a result, Valcor, in consultation with PP&L, made the following changes to the design and manufacture of the 65-11 valve: a residual washer was added to prevent the plunger from coming in direct contact with the stop (Model 65-11 A) and, in order to meet SSES operational needs (i.e., oil in the air system which could attack the EPDM o-rings), Valcor changed the o-ring material to Viton (Model 65-11B)in some of the valves.
SSES experienced further problems with Valcor valves in the May/ June time frame.
First, on May 23,1998, a model 65-11 A failed to stroke when energized. An examination of the valve performed by Altran identified some foreign substances within the valve. However, Altran concluded _ that the failure was most likely caused by the residual washer "providing wedging action resistancs between the coils of the spring and the backstop." Later, on June 3,1998, SSES reported that 5 of 20 model 65-11B valves had failed bench testing. In this case one valve did not stroke upon demand, one exhibited unacceptable leakage, one showed a sluggish response and two exceeded the required pick-up voltage. PP&L again concluded that some of the failures were attributable to the residual washer interference.
Although Valcor testing could not duplicate the failures, Valcor widened the inside i diameter of the residual washer to allow more clearance with the spring (Model 65-11C).
Then, on September 15,1998, during bench testing at SSES,5 of 8 spare model 65-11C valves failed to stroke upon demand. On September 18,1998,4 of the 5 valves again failed testing. The latter test was witnessed by Valect, Altran and NRC employees. All 8 valves were sent to Altran for root cause analysis (see Section 3.5).
3.2 Valve Desian and Manufacturina
- a. Scope The inspectors reviewed design and manufacturing documenta associated with the Model V70900-65 valves to identify potential design and manufacturing deficiencies and whether other models within the series may be subject to the same operational failures j as the SSES valves.- Documents reviewed included design drawings and qualification 1 and test reports. I l
' During the investigation of this issue, several more interrelated condition reports were l issued by PP&L This inspection report will deal with the overall conclusions developed from all of the PP&L Condition Reports.
4
f
- b. Observations and Findinas The Model 70900-65 is a low weight, 3-way, direct acting, balanced poppet, nuclear- I qualified solenoid operated valve (see Figure 1). According to Valcor product literature, l it can be operated in a normally open or normally closed configuration, in any attitude. ;
Each valve is hand assembled, adjusted, and tested prior to final shipment.
1
///////gffr/////Apjp3gm ag:- Backstop
&l Y
l
[d -
}
- Plunger / Backstop l
l l [ ; Air Gap
/// -
h .
5 5 resi ual asher)
DE-ENERGIZED
' s N x -
- N '
- Plunger CONDITION :
W N /t \ :,
Spring /f' / / -kQ Pin i mN /
[ ;
~
Backseat $,
f- "
// '
O-Ring s : ; C Port
'(: ,
BPori ,
/
/ Poppet A Port /
l' MODEL V70900-65-11 SERIES I
Figure 1 5
I
~ he inspectors reviewed valve assembly procedure AP70900-65-4, " Assembly Procedure Solenoid Valve,3-Way Valcor P/N V70900-65-4," which was used to assemble the valves. This procedure was originally developed for the V70900-65-4 valves, which were the first of this model series to go into production. All subsequent valves in the series were constructed using this procedure. However, since the 65-11 is an AC powered valve, the assembly procedure was not completely applicable. Valcer aug~ snted t1e procedure with Engineering Order (EO) 4113 to produce the model 65-11 alves. Subsequently, Revision D (January 9,1997) incorporated the assembly steps from EO 4113.
The assembly procedure required taking measurements with the valve in both the energized and de-energized positions. The assembler then calculated the number of shims to be added to achieve the correct amount of poppet travel. Further measurements were taken to determine optimum plunger length. The plunger was then machined and the valve assembled and tested. The inspectors noted that Valcor did not consider the gap dimension as critical to valve operation, believing the assembly procedure resulted in a sufficient air gap to prevent contact between the plunger and backstop. Valcor did not recognize that the assembly procedure could allow the gap to shrink to zero, based upon dimensional tole'ances and o-ring compression. It was Valcor's position that the post assembly testing assured a proper dimensional fit.
Otherwise, improper operation or leakapa would have been detected.
Following the problems identified at SSES, in April 1998 Valcor created procedure AP70900-65-11 A-1, " Assembly Procedure Solenoid Valve, 3-Way Valcor P/N V70900-65-11 A and P/N V70900-65-11B", for the assembly of the AC valves (initially for the model 65-11 A and subsequently ievised to address the model 65-11B). This procedure was developed in consultation with PP&L and addressed the potential tolerance errors identified in the previous procedure. The inspectors noted that the model 65-11C is not addressed. Valcor explained that since the 65-11C valves were retrofitted from existing model 65-11B valves, and not built from scratch, a new assembly procedure was not necessary. Valcor stated that the procedure would be revised if model 65-11C valves are purchased in the future.
The inspectors noted that there are several valves within the V70900-65 series. The Model V70900-65-11 series is the only model that operates on AC voltage. All other models (65-4,5,6,10,12,16 and 17 have been produced) operate on DC voltages.
Although Valcor is not convinced that the DC models are susceptible to the residual magnetism problem, Valcor prudently decided to include the residual washer on all subsequent valves in the model 65 series. Procedure AP70900-65-4 was revised in May 1998 to delete it's applicability to the AC valves and identify all DC valves that are applicable. The procedure title was also modified to, " Assembly Procedure Solenoid Valve, 3-Way Valcor V70900-65-XX Series (D.C. Powered)."
Following the September 1998 bench testing failures and subsequent root cause analysis, the inspectors again reviewed Valcor's valve assembly procedures. A concern regarding the control of a dry film lubricant which is applied to the valve plunger and sometimes to the plunger tube was identified (see Section 3.6).
6 I
a;- )
1
)
i 1
?. c. Conclusions l l
The inspectors found that Valcor dio not treat the air gap as a critical dimension in the l original assembly procedures. Based upon a review of documents and discussion with the valve design and manufacturing personnel the inspectors concluded that Valcor's basis for assuming the assembly and testing procedure provided an adequate air gap was reasonable at the time. However, Valcor did not recognize that aging of the o-ring in the energized state could reduce its resiliency and increase its compression, thus potentially affecting valve operation. Following the PP&L and Altran investigation, Valcor revised the procedure to assure the existence of a positive air gap.
l 3.3 ' Review of Valcor's 10 CFR Part 21 Proaram and its lmolementation
- a. Insoection Scoos l
The NRC inspectors reviewed the program that Valcor established to implement the provisions of 10 CFR Part 21, and verified selected aspects of the Part 21 program implementation,
- b. Observations and Findinas 10 CFR Part 21 Procedure The inspectors reviewed Valcor Procedure S2110, "10CFR :1 Defects and Non-Compliance Reporting Procedure," Revision D, dated February 7,1997. The practice for the evaluation of deviations and failures to comply and the reporting of defects within the required time frames were found to be generally acceptable in S2110. The inspectors noted that procedure S2110 also required i customer problems to be identified on Valcor " field reports," and dispositioned in !
accordance with S2110. The responsibility for documenting the results of deviation evaluations were found to be included in procedure S2110. In addition, S2110 l contained the necessary documentation forms,i.e., Form A,~ Nuclear Field Reports !
(NFRs); Form B, Field Problem Data Sheets; and Form C,10 CFR Part 21 Reporting Evaluation Forms (10CFR21 Evaluation Form). ]
However, the inspectors noted some minor changes that were needed in procedure
- S2110 to remove ambiguities and provide additional specificity. For example, although
" Form C" was provided for the documentation of the results of an evaluation, an explanation of the evaluation process2 was not established in S2110. It was noted that the S2110 procedure used the Part 21 term, " defect" as defbsd in $21.3 of 10 CFR Part 21, in some sections of S2110 instead of the correct Part 21 term of " deviation." l
- Additionally, the inspectors noted that Valcor did not define the meaning of several 10 CFR Part 21 words which have a specific meaning that are unique to 10 CFR Part 21. ,
The weaknesses were identified and discussed with the Valcor Quality Assurance j
~
2 10 CFR Part 21,921.3," Definitions," states that " evaluation" means the process of determining whether a particular deviation could create a substantial hazard or determining i whether a failure to comply is associated with a substantial safety hazard.
{ 7
-33 . j
- . . - . . - - - . - - - ~ _ - - . . - . - _ - _ . - -
Director, and the QA Director stated that appropriate revisions would be incorporated in S2110 within 90 days after the completion of the exit meeting. l l
Proaram imolementation The inspectors reviewed records related to the Part 21 program implementation including: deviation / failure to comply evaluation packages, completed NFRs, Field Problem Data Sheets; and 10CFR21 Evaluation Forms. The completed NFRs dating back to 1990 indicated that Valcor was reviewing nuclear customer component problems to determine whether any issues were potentially generic, or potential deviations or failures to comply that would need to be evaluated in accordance with Valcor's Part 21 program. All of the NFRs reviewed reflected an adequate and reasonable disposition of the matter except for NFR #017, "V70900-65-11 7 Valve Reportably Failed to Drop-Out when De-energized," dated March 25,1998.
A review of NFR #017 and its associated evaluation package, including Valcor's 10CFR21 Evaluation Form, indicated that Valcor did not perform an evaluation as defined in $21.3 of 10 CFR Part 21. That is, although $21.3 of Part 21 requires an eveluation to be performed, Valcor only reviewed the issue to determine whether it (
applied to other products that it produced. However, since Valcor discussed the matter with the only domestic customer that used that particular series of SOVs, the intent of
$21.21(b) was addressed. Therefore, no violation was cited in this area. (See further discussion in Section 3.4)
- c. Conclusions The inspectors determined that Valcor's S2110 procedure was generally well written and l addressed the salient aspects of 10 CFR Part 21. However, some weaknesses l involving a lack of specificity may have caused an inadequate evaluation to be t performed for NFR #017.
3.4 Part 21 Evaluation of SSES Valy.g3
- a. S.QQQt r
The inspectors reviewed V% ar's Part 21 report and evaluation relating to the failures of the SSES valves to assure tr.at their activities met applicable regulatory requirements.
In addition, the inspectors examined whether Valcor properly considered the failure mode's impact on other similarly designed valves.
- b. Observations and Findinas ;
Following the initial PP&L problem in September 1997, Valcor was not convinced of the potential for residual magnetism. This was based upon their inability to duplicate the failure as documented in SKA 17979, " Failure Analysis of Valcor Valve V70900-65-11."
On March 19,1998, PP&L and Altran met with Valcor to present the results of the root ;
cause analysis. During that meeting, PP&L was able to demonstrate the possibility of l residual magnetism to Valcor. Subsequently, on March 25,1998, Valcor initiated NFR l
8
- 017, which documented the potential residual magnetism problem and started their Part 21 evaluation process NFR #017 stated that Valcor would revie'v the design to verify suitability and, if necessary, modify the design to increase the valve's margin to close. (The design modification is discussed in Section 3.2, above.)
On May 19,1998, Valcor determined the problem was reportable and issued a Part 21 report to the NRC on the same day. The Part 21 report stated that, "Following 6 to 18 months of continuously energized service,3 units of Valcor Model V70900-65-11 air pilot valves have reportedly failed to stroke closed immediately upon de-energization...Despite best efforts, the delays in closing have not been able to be replicated outside of the plant system. Delays in closing have ranged from 1 to 5 minutes (emphasis added)." The report went on to state, "The V70900-65-11 is an AC-powered version and is the oniv version (emphasis added) of the V70900-65 air pilot series subject to this effect."
1 The inspectors reviewed Valcor's Part 21 report and associated documentation. As '
stated above, Valcor was not able to duplicate the initial failure and did not consider l residual magnetism as a viable failure mode until the March 19,1998 meeting. /st that point, Valcor initiated their Part 21 process. The report to the NRC was made within the Part 21 timeliness guidelines. l The inspectors examined the basis for Valcor's statement that the closing delays ranged from one to five minutes. Valcor stated that the report was based upon failures later reported by PP&L, not the September 1997 failure (which Valcor still questions). Valcor informed the inspectors that the time delays were based on verbalinformation supplied by PP&L. The inspectors were not able to identify any documentation at Valcor to support this statement. However, a review of PP&L condition reports related to the later failures supports the times reported by Valcor.
The inspectors also reviewed the basis for Valcor's determination that only the AC valves were affected by the residual magnetism issue. Valcor explained that the AC and DC valves are essentially identical in construction with the exception of the solenoid coil (verified by the inspectors). Further, the model 65-11 utilizes a rectifier to convert the AC input voltage to DC at the coil. This reduces the effective solenoid strength.
Since the solenoid strength is reduced, AC valve construction typically results in a longer plunger to meet the valve's operating requirements. Valcor believed at the time that the longer plunger resulted in a smaller air gap in the AC valves. In addition, as mentioned above, the assembly procedure for the AC valve was slightly different, using different construction steps and different dimensions. Therefore, Valcor concluded that the DC valves would typically contain shorter tubes, larger air gaps and would not be susceptible to the effects of residual magnetism.
The Duke Power Company (Duke) had bought a number of model 6516 (DC) valves from Valcor through R. E. Hiller. Upon hearing of the Part 21 report, Duke contacted Valcor to determine if the problem existed in their valves. Valcor informed Duke that they believed the problem was restricted to the AC valves, but offered to provide the valve modification (residual washer and verify air gap). Valcor personnel modified the 9
l b
c valves at Duke's McGuire facility. During the modifications, Valcor measured the air l gaps and found that some negative air gaps (contact between plunger and backstop) existed. Valcor performed a Part 21 evaluation," Engineering Report Evaluation for Model Number V70900-65-4 Series D.C. Solenoid Valves," and concluded that the potential for residual magnetism did not constitute a significant safety hazard and was not reportable under Part 21. However, as a precaution, Valcor modified the assembly procedure to incorporate the residual washer and assure a positive air gap in all DC )
valves. At the time of the evaluation, the only DC valves affected were the model 65-12 valves sent to SSES and the model 65-16 valves sent to Duke. Valcor did not formally !
inform either customer for the_ following reasons: PP&L had informed Valcor that this ;
failure mode would not result in a safety problem at SSES due to their particular application for the model 65-12 valves and PP&L was already aware of the potential for i residual magnetism; the Duke valves had already been modified._ Valcor incorporated 4 the assembly change to their model 65-17 valves, which were in production.
The inspectors reviewed Valcor's Part 21 evaluation and discussed the issue with <
responsible staff. The inspectors determined that Valcor still believes the original design ,
of the valve, particularly considering the strength of the retum spring, sufficient to )
overcome the effects of residual magnetism, despite the strong argument provided in the Altran report 3. The . inspectors also determined that Valcor took a conservative approach and modified the valve design and assembly procedure to negate the potential for a residual magnetista problem, in addition, all affected parties were aware of the potential problem. Further, valves under construction were manufactured using the new .
design and assembly procedure. Although Valcor's acceptance of the concern for residual magnetism may have been nonconservative, their actions taken to preclude futJre occurrences were conservative and met the intent of NRC regulations. ,
' Subsequently, on September 15,1998, five of eight spare model 65-11C valves failed bench testing at SSES. The bench test was repeated on September 18,199et and witnessed by NRC inspectors, Valcor representatives and an Altran representative.
These failures are discussed below.
- c. Conclusions The inspectors concluded that, based upon the information available at the time, Valcor's determination that the DC valves were not affected by residual magnetism was ,
reasonable. In addition, when Valcor became aware of the potential for residual magnetism in the DC valves, they performed a Part 21 evaluation and modifed the valve design and assembly procedures.
i 1
- 8 Altran's investigation included an analysis of the spring capacity and the stored compression energy and resilience of the backseat o-ring.
10 36-
i e
3.5 Observations of Testino at Susauehanna and Altran Corooration l l
- a. hone The NRC staff inspectors witnessed subsequent testing of several Valcor model V70900-65 series SOVs which failed to stroke at the SSES facility during a bench test ,
on September 15,1998. The subsequent testing was conducted on September 18, i 1998, at the SSES facility and on September 23-25,1998, at the Altran Corporation for l diagnostic / root cause analysis. I
- b. Observations and Findings Backaround in July 1998, Valcor modified eight V70900-65-11 SOVs to address operational concerns that were identified at SSES by PP&L personnel. The modification )
included adding Viton O-Rings, changing the residual magnetism washer with one which had a larger inside diameter, and regapping the valves to achieve better operation (Model 65-11C). A review of an SSES Source Verification Report,98-026, regarding Valcor assembly / modification of four of these SOVs indicated that the Vendor Quality inspector requested that an "as received test be performed at SSES to determine the ;
acceptability of the valves prior to being placed into stock." However, instead of functional testing upon receipt at SSES, the SOVs were placed in a QC-hold status until they were functionally tested on September 15,1998.
Testina On September 15,1998, SSES personnel tested all eight Valcor Model 65-11C SOVs. Five of the eight valves failed to open, or stroke regardless of the various applied pressure and voltage. SSES personnel contacted the PP&L valve engineering personnel and a second test was scheduled for September 18,1998. The PP&L valve engineering personnel contacted a testing laboratory, Altran Corporation, to perform additional SOV testing in case the SOVs continued to fail to stroke. The valve engineering personnel also contacted Valcor for their assistance in developing test methodology, identifying possible failure mechanisms, and to witness the September 18, 1998, testing at SSES and any subsequent testing at Altran. NRC personnel also witnessed the testing.
On September 18,1998, the five SOVs that previously failed were tested again. Four of the five again failed to stroke when energized, and the fifth did stroke (opened) but it did not meet the pull-in voltage acceptance criteria. Manual force was applied to one of the SOV valve poppet assemblies to determine whether it would stroke, but the plunger assembly could not be moved. After the testing was completed, Valcor, Altran and PP&L personnel discussed the results of the tests and planned diagnostic and root cause actions that would follow at the Altran facility.
Starting on September 22,1998, testing was performed at the Altran facility using an Altran Test Plan that was developed using input from Valcor and PP&L. The four SOVs which failed to stroke on September 18,1998, again failed. Some of the failed SOVs were disassembled and inspected by various methods. Inspection of a plunger assembly under a scanning electron microscope (SEM) revested a thin coat of the 11 graphite base lubricant on one side of the plunger; while the other side showed a thicker coating. The lubricant is a commercially obtained graphite in alcohol suspension called l Dag@ 156. Disassembly of other SOVs showed the same disparity in the lubricant l coating on the plunger and cylinder of the solenoid. Testing with a load cell determined l that it took approximate 25 pounds of force to break the plunger assembly free in the SOV cylinder.
l Based upon the tests and observations, Altran, PP&L and Valcor concluded that the most credible failure mechanism was adhesion of the plunger to the wall of the solenoid cylinder. The cause appears to be assembly of the SOV without allowing adequate
! drying time for the Dag@ 156. Further testing and observations confirmed that this l condition was also apparent on other failed SOVs.
l l c. Conclusions The inspectors observed testing at Altran and agreed with the conclusion that the cause l of the SOV failure-to-stroke was a result of assembly of the SOV without allowing adequate drying time for the alcohol. Further testing and observations suggest that this ,
condition may have been the cause of previous failures at the SSES originally thought to be caused by the residual washer binding.
3.6 Solenoid Ooerated Valve Assembly
- a. Insoection Scooe The NRC inspectors reviewed records related to the assembly of V70900-65-11 series, 3-way,120-volt, alternating current solenoid operated valves (SOVs) to identify methods employed by Valcor to control the application of lubricants.
- b. Observations and Findinas The NRC inspectors reviewed the 3-way solenoid valve assembly procedures for Valcor V70900-65-11 series SOVs and conducted discussions with assembly, engineering and quality personnel regarding Valcor's SOV lubrication methodologies. The inspectors noted that three different lubricants for the V70900-65-11 assembly operations are required. Two of the lubricants are used for the elastomers in the SOVs and the third lubricant', Dag@ 156, is used for metallic part application. The inspectors noted that the application of Dag@ 156, applied to the plunger and solenoid cylinder, was not procedurally required or controlled.
. The inspectors noted that although Dag@ 156 was used on the V70900-65-11 series SOV components, Valcor had not established written guidance for its use. The
!
- The NRC inspector noted that Valcor used Acheson Colloids Company (Acheson) dry
- film graphite lubricant (Dag@ 156) on certain metallic components on its V70900-65-11 series l SOVs.
12 o
l l' - .-. -
inspectors found that Valcor's assembly procedures and associated drawings did not address the application or use of the lubricant. Additionally, the inspectors noted that a cautionary note contained in the technical literature of the Dan @ 156, recommended a five minute drying time, was neither relayed to, nor taken into consideration by the Valcor assemblers. The inspectors determined that Valcor's failure to prescribe written guidance within its procedures and associated drawings to control the application of Dag@ 156 on its V70900-65-11 series colenoid operated valves constituted a nonconformance with 10 CFR Part 50, Appendix B requirements.
(Nonconformance 99900728/98-01-01)
The inspectors noted that, prior to the October 2,1998, exit meeting, Valcor had developed and implemented revised procedures to address the discrepancy.
l
- c. Conclusions The inspection identified that Valcor failed to prescribe written guidance within its ;
procedures and associated drawings to control the app!ication of dry film graphite ;
lubricant on its V70900-65-11 series solenoid operated valves.
PARTIAL LIST OF PERSONS CONTACTED l
1 Valcor Enaineerina Corporation l
Dana F. Shave, Business Unit Manager Jim Shieh, Quality Assurance Director Barry W. Matiez, Engineering Manager Ravi Rustagi, Manager of Applications Engineering Anthony Sirianni, Assembly and Service Manager Jose A. Vega, Jr., Nuclear Project Engineer Joseph E. Sheriden, Project Engineer Altran Corooration William J. McBrine, Manager, Engineering Mechanics & Materials Van Christie, Materials Scientist Pennsvivania Power & Liaht Comoany Michael H. Rose, Supervising Engineer Laurence M. Olson, Senior Engineer - Valve Design 13
_ - . ~ . _ _ . - . - - ~ . . . _ . . - . . . . . . . - . . . . .
ITEMS OPENED, CLOSED AND DISCUSSED Closed i 99900728/91-01-01 VIO Failure to adopt appropriate Part 21 procedures l
99900728/91-01-02 VIO Failure to post Section 206 l I
99900728/98-01-01 NON Failure to adopt appropriate procedures I
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14
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_- _ _ ~. -
@ CEO o k UNITED STATES j ,j NUCLEAR REGULATORY COMMISSION
'" . WASHINGTON, D.C. 20066 4001
% # December 14, 1998 James C. Woeber, Director, Steam Generator Product Line Westinghouse Electric Corporation Pensacola Plant Nuclear Projects Division 8301 Scenic Highway Pensacola, FL- 32515-7810
SUBJECT:
NRC INSPECTION REPORT 99900104/98-01 AND NOTICE OF NONCONFORMANCE
Dear Mr. Woeber:
On September 21-25 and November 16-17,1998, the U.S. Nuclear Regulatory Commission (NRC) performed an inspection at the Westinghouse Electric Corporation steam generator '
manufacturing facility in Pensacola, Florida. The enclosed report presents the findings of that inspection.
The inspection was conducted to assess: (a) attributes and implementation of the Westinghouse Pensacola Plant (WPP) quality assurance program in the areas of control of ;
special processes, procurement control, and control of tubing operations, to ascertain whether i they met NRC requirements; (b) conformance to customer procurement requirements; (c) the corrective actions taken in response to a stop work order imposed by STP Nuclear Operating Company; and (d) implementation of Part 21 of Title 10 of the Code of Federal Reaulations.
Overall, the results of the inspection indicate that you have established appropriate program criteria for control of fabrication and examination activities, with implementation noted generally to be good. In addition, it was found that WPP was in compliance with the provisions of 10 CFR Part 21 and was effectively implementing the Assessment Recovery Program that was developed in response to the November 6,1997, stop work order from the STP Nuclear Operating Company. During the inspection, the inspectors determined, however, that WPP did not adequately implement its quality assurance program criteria for procurement of submerged arc welding flux to comply with NRC and customer requirements. Specifically, WPP accepted i containers of submerged arc welding flux from a vendor, for use in steam generator pressure boundary welding applications, which were- (a) indicated by submitted vendor documentation to have been produced by a different manufacturer to that specified by the procurement documents, and (b) were not identified with the name of the manufacturer as required by the relevant material specification.
This issue is cited in the enclosed Notice of Nonconformance (NON), and the circumstances surrounding it are described in detail in the enclosed report. You are requested to respond to the nonconformance and should follow the instructions specified in the enclosed NON when preparing your response.
J.C. Woeber In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter and its enclosures will be placed in the NRC's Public Document Room. ,
Sincerely,
( qw, ti L Suzann_e)C. Black, Chief Quality Assurance, Vendor inspection, and Maintenance Branch Division of Reactor Controls and Human Factors Office of Nuclear Reactor Regulation Docket No. 99900104
Enclosures:
- 1. Notice of Nonconformance
- 2. Inspection Report 99900104/98-01 cc:
Mr. T. F. Walker Manager, Procurement Quality STP Nuclear Operating Company P.O. Box 289 Wadsworth, TX 77483 P'
4 NOTICE OF NONCONFORMANCE Westinghouse Electric Corporation Docket No.: 99900104 Pensacola Plant Based on the results of an NRC inspection conducted on September 21-25 and November 16-17,1998, it appears that certain of your activities were not conducted in accordance with NRC requirements:
Criterion Vil of Appendix B to 10 CFR Part 50," Control of Purchased Material, Equipment, and Services," states, in part, " Measures shall be established to assure that purchased material, equipment, and services . . . conform to the procurement documents . . . ."
Criterion XV of Appendix B to 10 CFR Part 50," Nonconforming Materials, Parts, or Components," states, " Measures shall be established to control materials, parts, or components which do not conform to requirements in order to prevent their inadvertent use or installation.
These measures shall include, as appropriate, procedures for identification, documentation, segregation, disposition, and notification to affected organizations. Nonconforming items shall be reviewed and accepted, rejected, repaired or reworked in accordance with documented procedures."
Contrary to these requirements, the inspectors identified on September 23,1998, that submerged arc welding flux, ordered by Purchase Agreement E64876, had been accepted at receipt inspection, despite the following:
- 1. The submitted vendor documentation indicated that the submerged are welding flux was from a different manufacturer and of a different brand name to that specified by the procurement documents. Additionally, WPP Material Specification C523C04 was not appropriately revised to reflect the use of flux from a different manufacturer.
- 2. The submerged arc welding flux containers were not identified with the flux manufacturer's name as required by the procurement documents and WPP Material Specification C523C04 (Nonconformance 99900104/98-01-01).
Please send a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555-0001, with a copy to the Chief, Quality Assurance, Vendor inspection and Maintenance Branch, Division of Reactor Controls and Human Factors, Office of Nuclear Reactor Regulation, within 30 days of the date of the letter transmitting this Notice cf Nonconformance. This reply should be clearly marked as a " Reply to a Notice of Nonconformance" and should include for each Nonconformanco: (1) the reason for the Nonconformance, or if contested, the basis for disputing the Nonconformance, (2) the corrective steps that have been taken and the results achieved, (3) the corrective steps that will be taken to avoid further noncompliance, and (4) the date when your corrective action will be completed. Where good cause is shown, consideration will be given to extending the response time.
Dated atpockville, Maryland this N dayof December 1998 Enclosure 1
U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION Report No: 99900104/98-01 Organization: Westinghouse Electric Corporation Pensacola Plant 8301 Scenic Highway Pensacola, Florida 32515-7810
Contact:
James C. Woeber, Director, Steam Generator Product Line Nuclear Industry Activity: Manufacture of steam generators Dates: September 21-25 and November 16-17,1998 Inspectors; lan Barnes, Technical Assistant, Division of Reactor Safety Region IV William M. McNeill, Reactor inspector Division of Reactor Safety -
Region IV Robert L. Pettis, Senior Reactor Engineer Quality Assurance, Vendor inspection and Maintenance Branch Division of Reactor Controls and Human Factors Approved by: Robert A. Gramm, Chief Quality Assurance and Safety Assessment Section Quality Assurance, Vendor inspection and Maintenance Branch Division of Reactor Controls and Human Factors Enclosure 2
1 Inspection Summary Westinghouse Electric Corporation Pensacola Plant (WPP) currently holds ASME Certificate of Authorization N-1669 for manufacture of Class 1,2, and 3 vessels and piping systems, Class CS core support structures, and as a material organization supplying ferrous and nonferrous bars, threaded fasteners, forgings, plates, clad plate, flanges, fittings welded without filler metal, fittings made from NPT stamped tebular products, seamless and welded with (NPT stamped) and without filler metal tubular products, welding material, structural shapes, sheet, and rounds and hollows. This inspection was performed at the WPP manufacturing facility and was focused on manufacture of South Texas Project, Unit 1 replacement steam generators.
During this inspection, the inspectors assessed conformance of fabrication and examination activities to NRC, ASME Code, and customer requirements. Specific subject areas reviewed during the inspection were procurement control, control of special processes, control of tubing operations, implementation of 10 CFR Part 21, and corrective actions taken in response to a November 1997 stop woJ: order from STP Nuclear Operating Company (STPNOC).
The inspection bases were as follows:
Appendix B. " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to Part 50 of Title 10 of the Code of Federal Reoulations (10 CFR Part 50),
10 CFR Part 21," Reporting of Defects and Noncompliance," and WPP's Nuclear Quality Assurance Program Manual, approved March 18,1998.
Overall, the results of the inspection indicated that WPP had established appropriate program criteria for control of fabrication and examination activities, with implementation noted generally to be good. In addition, the inspectors found that WPP was in compliance with the provisions of 10 CFR Part 21 and was effectively implementing the Assessment Recovery Program that was developed in response to the November 6, 1997, stop work order from STPNOC. However, the inspection identified that WPP did not conform to certain NRC and procedural requirements pertaining to procurement of submerged arc welding flux. This nonconformance is discussed in Section 3.4.1.
2 Status of Previous inspection Findings
-The scope of this inspection was limited to the manufacture of replacement steam generators by the Westinghouse Pensacola Plant (WPP) for South Texas Project, Unit 1.
Although the NRC staff previously accepted WPP corrective actions for findings that resulted from inspection Report 99900104/92-01, dated May 21,1992, the inspectors did not verify implementation of the corrective actions during this inspection.
2
3 Inspection Findings and Other Comments 3.1 Insoection for Comoliance with 10 CFR Part 21
- a. Insoection Scoce The inspectors reviewed WPP's implementation of and compliance with the requirements of Title 10 of the Code of Federal Reaulations (10 CFR) Part 21, as applied to the manufacture of replacement steam generators for South Texas Project, Unit 1. The steam generators were being suppliecnin accordance with STPNOC Technical Specification 4R129NS1014 "Specificaticn t for Replacement Steam Generators," Revision 2, dated March 11,1998.
- b. Observations and Findinas b.1 Implementation of 10 CFR Part 21 The inspectors reviewed WPP Procedure PQ-02-007, " Identification & Reporting of Conditions Adverse to Safety," Revision 9, dated August 19,1998. The procedure was noted to require.the WPP Safety Review Committee to evaluate deviations and failures to comply, and to report the evaluation results to the Energy Systems Business Unit Safety Review Committee located in Monroeville, PA. The Energy Systems Business Unit Safety Review Committee was responsible for review of the WPP Safety Review Committee evaluation results and deterrhination of whether the deviation or failure to comply was reportable pursuant to the requirements of 10 CFR Part 21.
The inspectors noted with respect to Section 21.21 of 10 CFR Part 21, which requires the filing of an interim written report if an evaluation of an identified deviation or failure to comply potentially associated with a substantial safety hazard cannot be completed within 60 days from discovery, that paragraph 6.4.3 of Procedure PQ-02-007, Revision 9, required notification of the Chairman of the Energy Systems Business Unit Safety Review Committee if it could not be conclusively determined within a timely fashion whether a condition adverse to safety existed. WPP explained that this reflected the understanding that the discovery period, defined by Westinghouse as the period of information gathering and completion of the documentation, associated with a given issue is not governed by a time limit, but instead it is the evaluation by the Safety Review Committee of such issue that is governed by the 60-day reporting requirement.
Once the issue is reported to the Energy Systems Business Unit Safety Review Committee and a Potential Issue is opened, the 60-day requirement of Energy Systems Business Unit Quality Policy / Procedure 21, Revision 2, would be in effect. The d '
inspectors found this explanation acceptable and consistent with the provisions of Energy Systems Business Unit Quality Policy / Procedure 21, Revision 2.
The inspectors also verified WPP's compliance to the posting requirements of Section 21.6 of 10 CFR Part 21, which requires the posting of current copies of the regulation, including adopted procedures, and Section 206 of the Energy Reorganization Act of 1974. The inspectors noted that WPP had posted copies of Revision 15 of the 3
Westinghouse Electric Company poster, dated March 4,1997, which satisfied the intent of the regulation. The poster was viewed by the inspectors in numerous locations throughout the facility, incluoing the administration building lunch area, manufacturing floor locations, and lobbies to both manufacturing and administration buildings.
b.2 Review of incorporation of 10 CFR Part 21 Requirements in Procurement Documents The inspectors selected several WPP purchase agreements for review that were applicable to basic components that were to be utilized in the South Texas Project, Unit 1, replacement steam generators. Included in the review were purchase agreements placed with: (a) Japan Steel Works (channel head forgings); (b) Ansaldo Energia S.P.A.
(upper shell barrels); (c) Creusot-Marrel, Inc. (transition cone forgings); (d) Kobe Steel, Ltd. (tube sheet forgings); (e) Sandvik Steel (seamless tubes); and (f) Equipos Nucleares SA (channel head cladding, machining and assembly). In addition, one purchase agreement placed with Forgemasters Engineering Ltd. was reviewed, pertaining to tube sheet forgings for Shearon Harris replacement steam generators.
Each purchase agreement was found to reference a Quality Note which corresponded to an inspection code for a particular supplier activity. The Quality Notes, numbered HA0051 through HA0058, contained the applicable specific quality system requirements, including compliance to Section 21.31, " Procurement Documents," of 10 CFR Part 21.
The Quality Notes also contained technical and documentation requirements related to nondestructive examination, certificates of compliance, and certified material test reports.
Item 1 of Quality Note HA0051, " Material for ASME Code and Safety Related Applications," invoked the reporting requirements of 10 CFR Part 21 for all U.S. firms; Item 2 required non-U.S. firms to report all deviations to WPP for evaluation and disposition; and item 3 required compliance to NCA-3800 of the ASME Boiler and Pressure Vessel Code and the applicable portions of 10 CFR Part 50, Appendix B, quality assurance program requirements, as qualified by WPP or the ASME by issuance of a Quality System Certificate.
The inspectors examined the subject material of several of the Material Disposition Reports that had been received by WPP from the above suppliers in the 1995-1998 time frame, and reviewed the 10 CFR Part 21 files to identify if any of the Material Disposition Reports had resulted in a deviation requiring the opening of a Potential issue. This review identified two issues which had been evaluated as a potentially reportable condition. The inspectors noted no other Material Disposition Reports in the sample reviewed that appeared to warrant evaluation for reportability. The first issue, Material Disposition Report 372765, pertained to the detection by Equipos Nucleares SA of an imbedded foreign ebject during nondestructive examination of replacement steam generator shell barrels that had been furnished by Creusot-Marrel for the Joseph Farley Nuclear Station. At the request of WPP, the issue was reviewed as a potential Part 21 by the Chairman of the Energy Systems Business Unit Safety Review Committee. On September 18,1998, WPP concluded that the issue had been adequately addressed by the Safety Review Committee and that no condition adverse to safety existed. The inspectors concurred with this conclusion.
4
s The second issue related to an evaluation of bobbin and rotating pancake coil indications at tube support plate intersections that were detected in the pre-shipment -
inspection of the Shearon Harris replacement steam generators. The indications were !
indicative of the presence of dings (i.e., mechanical damage causing a localized reduction in tube diameter ) and were most frequently located at the top and bottom of the 9th tube support plate. The WPP evaluation concluded that the potential for stress corrosion cracking at the dings was judged to be negligible and that no corrective action ,
was required. This conclusion was based primarily on: (a) absence of detected stress corrosion cracking, after up to 18 years of operational service, in steam generators tubed with thermally treated inconel 600 tubing and which contained dings larger than those present in the Shearon Harris replacement steam generators; (b) a similar absence of detected stress corrosion cracking in steam generators containing thermally treated Inconel 690 tubing, the tubing material selected for the South Texas Project, Unit 1, replacement steam generators, after up to 8 years of operational service; and (c) the superior corrosion resistance exhibited by thermally treated inconel 690 tubing in laboratory corrosion testing. On September 18,1998, the WPP Safety Review Committee concluded that based on the evaluation of the bobbin indications, no condition adverse to safety existed. The inspectors concurred with this conclusion.
b.3 Review of Anti-Vibration Bar End Cap Laser Weld Failure A review was performed by the lead inspector during November 1617,1998, of the - -
l WPP evaluation of a failure on October 29,1998, of an anti-vibration bar end cap assembly during staking and dimpiing to an anti-vibration bar (Steam Generator 4, Shop Order 12272). The end cap assembly consists of a channel-shaped component to which an end piece is attached by laser welding to each of the channellegs. The machine operator and inspector heard a noise during the staking and dimpling operation, that was subsequently established to have been caused by the separation of one side of the attached piece because of a lack of penetration of the laser weld into the channelleg. The welding of the anti-vibration end cap assembly was performed l approximately 1 year before the staking and dimpling operation, with the exact date unable to be established due to the non-assignment of unique identification numbers for individual assemblies.
The inspector reviewed WPP Report " Root Cause Analysis for the STP RSG AVB End .
Cap Laser Weld," Revision 1. This analysis concluded that the most probable cause of the failure was improcer power settings by the welding operator, resulting from a mixup in settings between that used for anti-vibration bar end caps and that used for welding non-nuclear diaphragms. The root cause was attributed to be the periodic production interruptions experienced during weldin', South Texas Project, Unit 1, anti-vibration bar end caps for Shop Order 12272 (Steam Generator 4), as a result of higher scheduler priority for diaphragms. Additional contributing factors were determined to be the absence of independent verification of power settings and inadequacies in the visual inspection process used for verification of weld penetration. The inspector reviewed the defined visual inspection criteria and confir'med that they were inadequate for establishing conformance of the weld to drawing requirements. The inspector also 5
considered, however, that quality assurance staff or inspection personnel should have
' previously identified these inadequacies. Overall, the inspector considered the root cause analysis conclusions to be reasonable. Re-inspection of end cap assemblies was initiated in the four South Texas Project, Unit 1, steam generators, with gas tungsten arc weld repairs performed for identified lack of penetration of laser welds.
The anti vibration bar end cap design used for South Texas Project, Unit 1, replacement steam generators was previously used for steam generators for Point Beach, Shearon Harris, and Kori. An analysis of design loads on the anti-vibration bar end cap assembly was performed, which conservatively estimated the maximum axial and transverse loads to be, respectively,100 lb and 300 lb. To establish the acceptability of installed anti-vibration bar end cap assemblies in completed stearn generators, pull-out tests (i.e.,
measuring the axialload required to pull the anti-vibration bar out of the end cap) and pull-off tests (i.e., measuring the transverse foad required to separate end cap components) were performed for different fabrication conditions. The test conditions evaluated for the pull out tests included one and two unacceptable laser welds, no end piece, and one tack weld with no laser welds. The minimum axial load value obtained was approximately 1800 lb for the no laser welds with one tack weld condition, a factor of 18 relative to the 100 lb maximum design load. The test conditions evaluated for the pull-off tests included one and two unacceptable laser welds, no laser welds coupled with an autogenous tack weld each side, and no laser welds coupled with a filler welded tack each side. The minimum transverse load value obtained was 2695 lb for the no laser welds coupled with a filler welded tack each side condition, a factor of approximately 9 relative to the maximum design load. The WPP Safety Committee concluded that the anti-vibration bar end cap failure was not reportable under 10 CFR Part 21. The inspector concurred with this conclusion and considered that the WPP approach to evaluation of the failure was excellent.
- c. Conclusions The inspectors concluded from the review and evaluation of a sample of material dispositien reports and purchase orders and the evaluation performed on anti-vibration end cap assemblies, that WPP was effectively implementing the provisions of the Energy Systems Business Unit quality program and was in compliance with the requirements of 10 CFR Part 21.
3.2 Procurement Control
- a. Insoection Scooe The inspectors reviewed the WPP technical and quality requirements that were imposed by purchase order and material spec. ication on the vendors selected for manufacture of pressure boundary components for South Texas, Unit 1, replacement steam generators.
Components included in the review were the tubing, feedwater and auxiliary feedwater nozzles, the upper and lower shell barrel forgings, the transition cone forging, the elliptical head forging, and the tube plate forging. The WPP material specifications examined during this review were B163C23, " Thermally Treated Alloy UNS N06690 6
l
1 (Alloy 690) Tubing for South Texas Unit No.1 Replacement Steam Generators (Section ill-NB, SB-163, Code Case N-20-3)," Revision D; A508C25, "SA 508 Class 3a Nozzle Forgings (Section lil NB)," Revision D; A508C23, "SA-508 Class 3a Shell (Ring)
Forgings (Section Ill-NB)," Revisions C and D; A508C24,"SA-508 Class 3a Elliptical Head Forgings (Section Ill-NB)," Revision C; and A508C20, "SA-508 Class 3a Tube Plate Forgings (Section ill-NB)," Revision D.
In addition, the inspectors examined the conformance of the vendor certification, for the Steam Generator 4 (Shop Order 12272) pressure boundary items, to the requirements of the procurement documents, ASME Sections il and lil Code, and STPNOC Specification 4R129NS1014, Revision 2.
A limited review of welding material procurement and receipt inspection was also performed which has been documented in Section 3.4.1 (Control of Welding) of the inspection report.
- b. Observations and Findinas l The inspectors found that the WPP material specifications, utilized for the procurement l of steam generator pressure boundary materials, were consistent with ASME Sections !!
I and Ili Code requirements and the technical and quality requirements contained in STPNOC Specification 4R129NS1014, Revision 2. The vendor certification for tha Steam Generator 4 (Shop Order 12272) pressure boundary items was ascertained to be in compliance with ASME Section lli Code and WPP material specification sampling and testing requirements,
- c. Conclusions The inspectors concluded from the review of WPP material specifications and vendor documentation that: (1) present conformance to ASME Code and STPNOC Specification 4R129NS1014, Revision 2, requirements was good; and (2) specification reviews performed as part of the corrective action response to the stop work order from STPNOC had been effective.
3.3 Control of Tubina Ooerations
- a. Insoection Scone The inspectors reviewed WPP Drawings 6488E99, " Steam Generator Upper Shell Detail Assy," Sub 3; 6488E93, " Steam Generator U-tube Insertion and Assy with Helium Leak Test," Sub 4; and 6488E68, " Steam Generator Tube Plate and Lower Barrel Weld Assembly and Machining," Sub 4. The inspectors also reviewed Detailed Manufacturing Procedures: DMP-5563, " Steam Generator Tubing Tube Tack Expansion," Revision 22; DMP-6255, " Foreign Object Control Procedure for a Complete Steam Generator,"
Revision 06; and DMP-6401, " Steam Generator Hydrau':c Tube Expansion," Revision 8; ano Quality inspection Procedures: QlP-310. " Tub, msertion Inspection Requirements," Revision 16; and QlP-3605, "Hydroswage Equipment Calibration,"
7
Revision 01. The team observed the insertion of tubes, tack expansions, and crevice expansions of tubes.
- b. .Qbservations and Findinos The team found that procedural controls for tubing operations were satisfactory and consistent with the requirements of STPNOC Specification 4R129NS1014, Revision 2.
The team also observed insertion of tubes into Steam Generator 3 (Shop Order 12271, Traveler Operation 60), and noted that the tubing operation was, in general, proceeding smoothly and without problems. Tubing of the steam generator was performed in a
' evel "A" clean room, the highest level of foreign object control. Level "A" controls imposed limitations on personnel access and tools in the area and required daily sweeping and mopping of the area. Personnel performing tubing operations were also required to wear head coverings, rubber shoe covers, gloves, and coveralls.
The inspectors noted certain conditions which, although not considered to represent a significant foreign object or tubing damage concern, appeared to be inconsistent with the overall program cleenliness controls. An abrasive cloth used to remove minor markings from tubes was observed to not be discarded after it fell on the floor of the clean room. When a new shipping box of tubes was introduced to the area, the unpacking process resulted in loosening of some staples and some splintering of the wood on the box edges. Staples held down plastic fiber strapping that secured the tubes during shipping. Some staples had only one leg in the wood because the straps were torn rather than cut. Some staples could be removed with only finger pressure.
After the inspectors identified these undesirable conditions, WPP staff taped over the staples and splinters and noted the conditions in a log book so that subsequent shifts would become aware of the conditions. In addition, WPP staff held a " stand down" meeting with the work crews to rei'1 force sensitivity to the Level "A" requirements.
The inspectors observed the performance of hydraulic tack expansions of tubing in Steam Generator 3 (Shop Order 12271) and full-depth hydraulic tube-to-tube sheet expansions in Steam Generator 2 (Shop Order 12270), Operations 070 and 060, respectively, on their travelers. The tack expansion secured the tube for welding, and the crevice expansion eliminated the crevice between the outside diameter of the tube and the inside diameter of the tube sheet holes. Tne team observed the expansion processes and observed the quality control checks made of the expansion processes.
These checks included verification of the pressures used for expansions, plug gauging the tube inside diameters after expansion, and measurement of the expansion depth into the tube. The crevice expansion diameters were sampled and measured automatically. An online computer recorded and plotted control charts of the measurements for acceptance. These measures appeared to assure expansion of the tubes to the proper depth.
- c. Conclusions The inspectors found the overall WPP tubing criteria and practices to be consistent with the requirements of STPNOC Specification 4R129NS1014, Revision 2. Corrective 8
l
actions were immediately taken in response to inspector observations regarding use of abrasive paper, loose staples, and wood splinters.
3.4 Control of Soecial Processes 3.4.1 Control of Wclding
- a. Insoection Scoce The inspectors reviewed Procedure DMP-5881," Shop Control of Welding Materials ASME Boiler Pressure Vessel Code," Revision 20, to determine the measures used to control order, receipt, conditioning, and issue of welding materials. The inspectors toured the welding material crib and observed the material identification practices and the temperatures of storage ovens ( for covered electrodes and submerged arc welding flux). In addition, the inspectors reviewed the issue records for E7018 and E9018 covered electrodes, SFA 5.23 Class EM-2 electrode, and Oerlikon OP-121-TT submerged arc welding flux and examined the applicable cedified material test reports and material specifications. The material specifications reviewed were: (1) C051C02,
" Carbon Steel Electrode SFA 5.1 Class E7018 for Shielded Metal Arc Welding (SMAW)," Revision C; (2) C05502, " Low Alloy Steel Electrode SFA 5.5 Class E9018M for Shielded Metal arc Welding," Revisions B and C; (3) C523C04, "Oerlikon Type OP-121TT Flux for Submerged Arc Welding (SAW) Low Alloy Steel," Revision C; and (4) C523C05, " Low Alloy Steel Electrode SFA 5.23 Class EM-2 and Oerlikon OP-121TT Flux for Submerged Arc Welding (SAW)," Revision A. The inspectors 1 visited the warehouse where bulk welding materials were si ed to examine the F. ..,ufication on submerged arc welding flux containers and review storage conditions.
Limited welding activities were in process on the South Texas Project, Unit 1, replacement steam generators during the onsite inspect:on. The ;nspectors were able to observe performance of; (1) tube-to-tube sheet weldinj in Steam Generator 4, Shop Order 12272, using Welding Procedure Specification 41641, Revision 13; and (2) performance of a shielded metal arc weld buildup (Detali H) on the inside diameter of the "H" upper shell barrel for Steam Generator 2, Shop Order 12270, using Welding Procedure Specification 4222 Revision 5.
- b. Observations and Findinos Weldino Materials The inspectors determined that appropriate technical controls had been developed for shop control of welding materials, with no problems noted with respect to storage oven temperatures, identification of covered electrodes, and issue control. Review of certified material test reports for E7018 and E9018 covered electrodes showed the chemistry and mechanical properties were in conformance with the requirements of the respective WPP material specifications and ASME Sections 11 and 111 Code.
9
i The inspectors observed, however, during examination of the submerged arc welding flux storage ovens, that a single WPP heat code had been assigned to what appeared potentially to be multiple lot numbers of flux (i.e., the flux ovens were marked as containing Oerlikon OP-121-TT flux, assigned Heat Code SZ47, from Manufacturing Loi Nos.1719-7-3,1719-7-4,1719-7-5,1719-9-6,1719-10-1,1719-10 2,1719-14-6, 1719-15-1,1710-15 2, and 1719-15-3). Additionalinspection regarding the procurement, receipt inspection, and qualification testing of this flux (which is used for submerged are welding of the steam generator low alloy steel pressure boundary weld joints) identified the following:
The applicable procurement document for the submerged arc welding flux was ascertained by the inspectors to be WPP Purchase Agreement E64876, dated June 28,1996. This purchase agreement required the vendor, Welding Engineering Supply Co., to supply Oerlikon Type OP-121TT automatic submerged arc welding flux per WPP Material Specification C523C04, Revision C, and the requirements of the attached Quality Note HA0051.
WPP Material Specification C523C04, "Oerlikon Type OP-121TT Flux for Submerged Arc Welding (SAW) Low Alloy Steel," Revision C, stated that its purpose was to establish the requirements for Oerlikon Type OP-121TT flux for submerged arc welding to comply with ASME Section 11, Part C, and ASME Section Ill, Division 1, Subsection NB, paragraph NB-2400. The material specification included in its requirements that: (1) certified test reports should include a statement of cCmpliance to the purchase order and the specification and revision letter; and (2) each unit paukage of flux should be identified with material classification and specification numbers, supplier's name and trade designation, and lot number.
The inspectors noted that WPP Internal Quality Release 055814, dated November 11,1996, identified the flux received in response to WPP Purchase Agreement E64876 as Oerlikon Type OP-121TT, Heat No. 1719-3-6. The Certificate of Compliance issued by Welding Engineering Supply Company for the WPP Purchase Agreement E64876 order identified, however, the type of flux as Hobart 121/SD3. A Hobart Certificate of Compliance for the WPP purchase agreement number was also provided to the inspectors. This document indicated the type of flux to be Hobart 121 and identified that the flux had been produced under Quality System Program FMQP 01, Revision 4, audited and approved by WPP as meeting ASME Section lll Code, Subsection NCA-3800, and the applicable portions of 10 CFR Part 50, Appendix B. Both Certificates of Compliance indicated the lot number of the flux was 1719.
The inspectors questioned WPP personnel concerning the supply and acceptance of a Hobart manufactured submerged arc welding flux, when the applicable procurement documents required Oerlikon Type OP-121TT flux to be furnished.
The inspectors were informed that WPP became aware that Oerlikon (a Swiss manufacturer of welding products) had discontinued its ASME Quality System Certificate as a materials organization, thus necessitating, for use of the vendor 10 l .
l l
l l
products in ASME Section lli Code welding applications, an audit by WPP to verify that a quality assurance program was in effect which conformed to the provisions of ;
paragraph NCA-3800 in Section lli of the ASME Code. Upon ascertaining that l Hobart, a domestic manufacturer of welding products, was licensed to Jse Oerlikon formulations in the manufacture of welding consumables, WPP opted to perform an NCA-3800 audit at Hobart. This audit was performed on August 1-2,1996, and the vendor conditionally approved pending receipt of corrective actions for two identified findings.
A copy of the Oerlikon licensing agreement was obtained by WPP from Hobart at hw inspectors' request. The inspectors noted during review of this agreement that
- required the licensee to use its own trade names for the products and did not permit the licensee to use Cerlikon brand or trade names. No information was provided by WPP personnel with respect to why Purchase Agreemerit E64876 and WPP Material Specification C523C04 were not appropriately revised to reflect the use of the Hobart manufactured product The failure of receipt inspection to identify that the flux identity (contained in the submitted vendor documentation) did not conform to that specified in the procurement documents, as required by Criterion Vii, " Control of Purchased Material, Equipment, and Services," and Criterion XV, " Nonconforming Materials, Parts, and Components," of Appendix B to )
10 CFR Part 50, was identified as Nonconformance 99900104/98-01-01.
~
l
. Paragraph 3.1 of WPP Material Specification C523C04, Revision C, included in its I requirements that flux containers be identified with the supplier's name, trade I designation, and lot number. The inspectors noted, however, during examination of containers in warehouse storage, that the information on the container identification labels did not include the flux manufacturer's name and showed the trade designation to be OP-121 TT. The failure of receipt inspection to identify that the flux containers did not contain the manufacturer's name was identified as an additional example of Nonconformance 99900104/98-01-01.
. The intpectors observed that the identification labels on flux containers in warehouse storage showed the same three number type format (e.g., 1719-5-1) for lot identity as was previously noted on flux storage oven documentation. The inspectors requested information on what this format represented, in that the scope of WPP testing of this flux (with respect to the requirement in paragraph NB-2420 of the ASME Section ill Code to test each combination of heat of bare electrode and lot of submerged arc flux) appeared to be based on the assumption that all of the flux containers were from a single lot. WPP contacted Hobart to obtain the information and were informed that the three numbers represented, respectively, lot number, batch number, and mix number. An initial complete chemical lot was made up which was then subdivided into batches, followed by further subdivision into smaller quantities for mixing with a binder. The inspectors were informed, in response to questions regarding uniformity of flux composition from batch to batch, that a weld test pad was made for each batch and chemical analysis performed.
Copies of the batch analyses for Lot 1719 were obtained from Hobart by WPP.
11 J
,- The inspectors concluded from review of the analyses that the values were reasonably uniform from batch to batch.
During review of mechanical properties of submerged arc weld deposits made using Lot 1719, the inspectors noted that the scope of testing performed by WPP was not consistent with the requirements of Materials Specification C523C05, " Low .
Alloy Steel Electrode SFA 5.23 Class EM-2 and Oerlikon OP-121TT Flux for '
Submerged Arc Welding (SAW)," Revision A. The specification required performance of additional Charpy-V and drop weight tests to those required by
, ASME Section lli Code. The scope of actual notch toughness testing performed ,
^
was limited to that required by ASME Section lli Code. After identification of the -
discrepancy by the inspectors, WPP revised Material Specification C523C05 to eliminate the requirement for the additional Charpy-V and drop weight tests.
Production Weldina The inspectors noted good compliance to welding procedure specification requirements [
during observation of gas tungsten are tube-to-tube sheet welding and performance of a shielded metal arc weld pad buildup.
- c. Conclusions The inspectors found that welding material storage and issue were satisfactorily controlled, with good compliance to welding procedure specification requirements noted during limited observationc of production welding. A nonconformance was identified ,
with respect to the procurement of submerged arc welding flux.
3.4.2 Post-Weld Heat Treatment
- a. l_n soection Scone 4
The inspectors reviewed the conformance of two WPP post-weld heat treatment procedures to the requirements of the ASME Section lli Code. The selected procedures were: Procedure DMP-6181, " Post Weld Heat Treatment Steam Generator Upper Assembly and Closure Seam," Revision 14; and DMP-6420, " Heat Treat Procedure including Preheat, Interpass, Gouging, Hydrogen Baking and Post Weld Heat Treatment," Revision 03. The inspectors examined the post-weld heat treatment records for two South Texas Project, Unit 1, replacement Steam Generator 3 l assemblies (i.e., Upper Shell Assembly, Serial No. 8962J; and Transition Conc to Lower
~ Shell Barrel B Assembly, Serial No. 8910J) to verify conformance to procedural requirements. In addition, the inspectors reviewed Quality inspection Procedure )
OlP-3146, " Post Weld Heat Treat Data Sys (SG&PRZ)," Revision 09, to verify that ,
appropriate provisions had been made to monitor accumulated post-weld heat treatment time with respect to the qualification times that had been used in the testing of base materials, welding materials, and welding procedure qualifications.
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_ ..-. . - - . -- ..- - -...~~- . - - - -- --- -
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1 5 b. Observations and Findinas -
1 4
The inspectors verified that the procedural (DMP-6181, Revision 14, and DMP-6420, Revision 03) requirements for hold time and permissible heating rate, cooling rate, and
.!- temperature gradient were consistent with ASME Section ill Code requirements. The
' inspectors noted from review of the procedures that significant numbers of j thermocouples were used to monitor assembly temperatures, with the selected j attachment locations providing a high degree . I assurance of required temperatures and 3 uniformity being achieved. No problems were identified during review of the post-weld j heat treatment records for the South Texas Project, Unit 1, replacement Steam Generator 3 Upper Shell and Transitior' Cone to Lower Shell Barrel B assemblies, w;tn
. good procedural compliance noted. The inspectors determined from review of Procedure OlP-3146, Revision 09, that appropriate provisions had been made for monitoring accumulated post-weld heat treatment time with respect to the qualification times used during testing of base materials, weld materials, and welding procedure qualifications. The inspectors additionally noted that the utilization of a hydrogen baking
{ practice on completed welds significantly reduced the number of required pov ./ eld heat
! treatment cycles; thereby, maintaining accumulated post-weld heat treatment times well
! below the qualification times used for materials and welding procedure qualification j testing.
7 4 c. Conclusions
] The procedural requirements for conduct of post-weld heat treatment on South Texar
- Project, Unit 1, replacement steam generator assemblies were consistent with the ASME Section 111 Code, with good procedural compliance noted during review of post-weld heat treatment records. Appropriate actions had been taken to monitor j component accumulated post-weld heat treatment time with respect to qualification j times used in materials and welding procedure qualification testing, i
j 3.4.3 Nondestructive Examination
- a. Insoection Scagg The inspectors reviewed the conformance of WPP nondestructive examination procedures to the requirements of the ASME Section V Code and OTPNOC Specification 4R129NS1014, Revision 2. Included in this review were Nondestructive Examination Procedures 8001, " Guidelines for RT/UT Layout of Weld," Revision 01; 8105, " Solvent Remover Liquid Penetrant Exam," Revision 16; 8106, " Water Washable Liquid Penetrant Exam," Revision 13; 8139, " Water Washable Liquid Penetrant Exam,"
Revision 02; 8567, " Ultrasonic Examination of Austenitic Welds," Revision 05; B573,
" Ultrasonic Examination of Welds," Revision 17; 8141, " Solvent Remover Liquid Penetrant Exam," Revision 01; 8330,
- Radiographic Procedure," Revision 01; and 8609,
" Visual Examination of S/G Welds," Revbion 01.
The inspectors witnessed the ultrasonic examination of Weld P (Elliptical Head to Upper Shell Barrel J) in Steam Generator 3 (Shop Order 12271) for South Texas Project, 13 56-
Unit 1. The inspectors also reviewed the radiographic film for the following South Texas Project, Unit 1, steam generator welds: (1) Steam Generator 2 (Shop Order 12270),
Weld C (Lower Shell Barrel A to Lower Shell Barrel B), G (Lower Shell Barrel B to Transition Cone), and Y (Lower Shell Barrel A to Tube Sheet); and (2) Steam Generator 1 (Shop Order 12269), Weld R (Feedwater Nozzle to Upper Shell Barrel H).
As of the onsite inspection, radiographic examinations of Welds S and SO (Manways to l Upper Shell Barrel J) had not been performed. As a result, the radiographic film was reviewed for the corresponding welds in replacement Steam Generator 3 for Shearon Harris (Shop Order 12245).
- b. Observations and Findinas The inspectors found that the nondestructive examination procedures were consistent with the requirements of ASME Section V Code and STNOC Specification 4R129NS1014, Revision 2.
The inspectors verified during review of the Steam Generators 1 and 2 radiographic film
~ that WPP had used the correct image quality indicator (penetrameter) and film type, and that film density values were acceptable. No disagreements were noted with the WPP film interpretations. In general, the radiographic results indicated that the welds were of high quality and free of apparent defects. Minor exceptions were a few views which showed acceptable levels of porosity and slag inclusions. The inspectors noted during review of radiographic film for Welds S and SO in Steam Generator 3 (Shop Order 12245) that, because of the section thickness variations in the manway welds, certain views required " double film viewing" (i.e., film overlapping) to achieve an acceptable film ,
density. The use of " double film viewing" was not documented, however, on the I radiographic inspection reports (reader sheets). The inspectors noted no problems at the locations where " double film viewing" was required to achieve an acceptable film density, but considered that its use should be appropriately documented on the reader sheets. WPP re-reviewed the radiographs for the Shearon Harris replacement steam generator manway welds subsequent to the onsite inspection, and documented when double film viewing was used.
As part of the observation of the ultrasonic examination of Weld P in Steam Generator 3 (Shop Order 12271), the inspectors witnessed calibration of the instrument using Calibration Block SV80. The inspectors verified that the calibration block was manufactured from the same material type as the vessel being inspected, the correct reference level was used, and the instrument exhibited required linearity and alignment.
Axial and circumferential scans of Weld P with 60*,45*, and O' search units were observed. The examination personnel were noted to conform to the procedural requirements with respect to couplant, search units, instrumentation, settings and techniques. Scanning was performed at 14 dB above the reference level and with the proper overlap, coverage, and speed. Intermittent rejectable indications were found by WPP personnel during these examinations, which were located at the weld center line about 0.5 inches from the weld root. The indications were appropriately documented in an inspection report.
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- c. Conclusions -
The inspectors found that the WPP nondestructive examination procedures were consistent with the requirements of ASME Section V Code and STPNOC Specification 4R129NS1014, Revision 2. Performance of radiographic and ultrasonic examinations was noted to conform to procedural requirements. An observation made regarding documentation of the use of " double film viewing" during interpretation of manway weld radiographs was appropriately responded to by WPP quality assurance staff.
3.5 Corrective Action Proaram - Stoo Work Order Backaround On November 6,1997, STPNOC issued a stop work order pertaining to all Westinghouse activities associated with the procurement and fabrication of South Texas Project, Unit 1, replacement steam generators. The order applied to all Westinghouse and Westinghouse contractor facilities worldwide, but excluded replacement steam generator licensing activities. The basis for the order resulted from identified engineering and manufacturing deficiencies and STPNOC concerns regarding the effectiveness of WPP oversight of its vendors The most significant engineering deficiency resulted in the nozzle openings for the .
feedwater an 2 auxiliary feedwater nozzles being cut at incorrect locations in the "H" upper shell barrels for the four STP, Unit 1, replacement steam generators. The barrels were fabricated and machined by Ansaldo in accordance with Detail F of WPP Shell Fabrication Drawing 6488E61, Revision 4, which was revised by Engineering Change Notice 33447, dated November 1996, to show the final placement and dimensions of the FW and AFW nozzle openings. The view presented in Detail F of the upper shell barrel was a " Bottom View" looking up at the replacement steam generator However, WPP General Arrangement Drawing 6438E99, Revision 4 utilized the typical" Top View" convention for the replacement steam generator, and indicated: (1l the feedwater nozzle to be located at 29' from the 7" axis (South) towards the "W" axis (East), and (2) the auxiliary feedwater nozzle to be located at 25 ' from the "Z" axis towards the "Y" axis (West). The use of different conventions resulted in the nozzle locations chown in Detail F on the fabrication drawing sent to Ansaldo being 180* from the locations required by the general arrangement drawing. The error was identified by WPP after receipt of the shell barrels from Ansaldo and prior to further fabrication operations.
Material Deficiency Reports 373042,373043, 373044, and 373046, dated November 5, 1997, were issued, with corrective actions taken including cutting two of the upper shell barrels to make a new one and ordering of replacement upper shell barrels.
The most significant vendor error was documented in Material Deficiency Report 367490, dated April 30,1997. In this instance, Ansaldo cut a manway opening in an upper shell barrel 3 inches above the required location. The final disposition was '
to scrap the barrel.
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Contingent upon release of the stop work order, STPNOC requested that Westinghouse develop a total assessment / recovery plan that included a historical confirmation that all work to date, and resultant work products, wees in compliance with established requirements and were technically and prograt matically correct. By letter dated January 9,1998, WPP submitted its Assessme it Recovery Program (ST-W2-NOC-000358, Revision 2) to STPNOC for review and approval. The program consisted of two parts- (1) a historical review a f manufacturing, quality assurance and purchasing; and (2) a root cause analysis of e +nts in the areas of supplier oversight, cause and prevention, quality assurance and . .anagement oversight, design assurance, and program compliance. Imp;ementation of one Ass?ssment Recovery Program was accomplished by use of review teams. Reviews of manufacturing routings and records, purchase orders and quality requirements, material heat code packages, and nonconformance documents were performed by WPP technical services and product assurance personnel. One team, the " Historical Design Engineering Review Team,"
consisted of individuals who were not involved in the WPP operation. The stop work order was lifted by STPNOC in January 1998 following review and approval of the Assessment Recovery Program.
- a. Insoection Scoce The inspectors reviewed work products of the Assessment Recovery Program and interviewed participating personnel. In addition, a review was performed of : (1) the assessment results documented in Report NSD-E-MSI-98-039, " Historical Review Design Engineering, South Texas Project, Replacement Steam Generators " Revision 1, dated July 1998; and (2) the root cause analysis documented in Report OS-98-0083, dated January 20,1998, which was transmitted to STPNOC by WPP Letter l ST-W2-NOC-000363, dated January 27,1998.
- b. Observations and Findinos Report NSD-E-MSI-98-039, Revision 1, was found to address four review subjects (i e.,
design and manufacturing drawings, material specifications, process specifications, and engineering design). The inspectors noted that the report made several recommendations to WPP management in regard to the design document verification process. The inspectors ascertained from review of the root cause analysis report that WPP had concluded that managerial oversight was at the center of the cause. WPP management made a commitment to establish on a forward basis a commitment to total quality programs, which included absolute compliance to technical specification and procedural requirements and appropriate overview of WPP quality system performance to identify, sliminate or resolve problems either intemally or at the supplier.
The inspectors reviewed several work products of the Assessment Recovery Program corrective actions which included: (1) a material specification review (PEN-98-326, dated July 29,1998); (2) a report discussing the cobalt and carbon content of welding consumables for South Texas Project (NPD/E/ PEN-98-311, dated July 29,1998), which was prepared to verify if cobalt and carbon contents of stainless steel and Inconel filler materials meet STPNOC Technical Specification requirements; and (3) a check list for 16 drawings (NPD/E/ PEN 98-331, dated August 3,1998) and a check list for Process Specifications (NPD/E/ PEN 98-337, dated August 6,1998), both documenting the resolution of comments made in the historical review.
The inspectors also reviewed the implementation of all five design assurance corrective actions. Revisions made to procedures as a result of the corrective action plan included: (1) revisions to PQ-02-031, Revision 5, which added a conformance matrix; (2) revisions to PQ-02-006, Revision 11, which required all drawings be independently checked by drafting personnel; (3) revisions to PQ-02-033, Revision 2, which
,plemented various calculation note requirements; (4) revisions to PQ-02-003, devision 8, which required an impact assessment of design changes on engineering drawings;,and (5) revisions to numerous procedures to incorporate scratch criteria to ensure that WPP, and its customers, have a clear understanding of the expectations in this area and the level of scratches considered acceptable. The procedures reviewed by the inspectors, to verify effective implementation of the scratch criteria, included the following: WPP Process Specification 87111WU, "Special Engineering Requirements During Tube Bundle Assembly," Revision G; WPP Material Specifications B163C20, Revision G, and B163C27, Revision E, used, respectively, for thermally treated Alloy 690 tubing for Arkansas Nuclear One, Unit 2 and Farley Units 1 and 2 replacement steam generators; Sandvik Control Procedures (for Arkansas Nuclear One, Unit 2) 5404, " Visual Standards," Revision 2, 5488, " Inspection of U-Bent Tubes," Revision 2, and 5482, " Inspection of Straight Tubes," Revision 1; WPP Procedure DMP-5563,
" Steam Generator Tubing, Tube Tack Expansion," Revision 22; and WPP Quality inspection Procedure 3150, " Tube Insertion Inspection Requirements," Revision 16.
- c. Conclusions Based on the review and evaluation of the WPP work products that were generated in response to the STPNOC stop work order, the inspectors concluded that WPP was effectively implementing the provisions of the Assessment Recovery Program which was submitted to the utility on January 9,1998, and was in compliance with the applicable provisions of 10 CFR Part 50, Appendix B.
3.6 Entrance and Exit Meetinas At the entrance meeting on September 21,1998, the lead inspector discussed the scope of the inspection, outlined the areas to be inspected, and establishea interfaces t with WPP management. In the exit meetings on September 25 c.d November 17, ,
1998, the lead inspector discussed the findings and observations of the inspection.
Documents reviewed during the inspection which were identified as containing -
proprietary information included WPP material specifications, Report :
NSD-E-MSI-98-039, Revision 1, and Report QS-98-0083, dated January 20,1998. No information was included from these documents in the inspection report that was t considered proprietary.
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PARTIAL LIST OF PERSONS CONTACTED Westinahouse Pensacola Plant J. Alexander, Controller J. Allen, Senior Quality Assurance Engineer S. Anderson, Purchasing Manager D. Arnott, Works Engineering Manager G. Bieberbach, Project Manager, South Texas T. Coriale, Product Assurance Manager R. Eubanks, Area Manager, Manufacturing R. Frisbey, Human Resources Manager L. Garrett, Engineering Aide D. Jennings, Quality Control Supervisor M. Kachmar, Engineering Manager J. Land, Design Engineer L. Lavoie, Area Manager, Manufacturing
< O. Machado, Manager, Nuclear Projects and Marketing K. Merritt, Quality Assurance Manager K. Olmstead, Quality Assurance / Reliability Engineer ,
U. Schneider, Manager, Technical Services !
C. Schaishuhn, Plant Manager M. Weatherly, Quality Assurance Engineer / Nondestructive Examination Level 111 J. Woeber, Director, Steam Generator Product Line
)
STP Nuclear Ooeratino Comoany C. McIntyre, Director, Engineering Projects R. Rehkugler, Director, Quality C. Sist, Resident inspector 4 Arkwricht Mutual Insurance Comoany W. Jones, Authorized Nuclear inspector l ITEMS OPENED l Ooened l
99900104/98-01-01 Para. 3.2 NON Procurement of submerged arc welding flux l
18 LIST OF DOCUMENTS EXAMINED )
i
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South Texas Project Nuclear Ooeratino Comoany Soecification 4R129NS1014. Replacement Steam Generators, Revision 2 Westinahouse Process Soecification 87111WU,"Special Engineering Requirements During Tube Bundle Assembly," Revision G Westinahouse Material Soecifications A508C20, SA-508 Class 3a Tube Plate Forgings (Section ill-NB), Revision D A508C23, SA-508 Class 3a Shell(Ring) Forgings (Section lil-NB), Revisions C and D i A508C24, SA-508 Class 3a Elliptical Head Forgings (Section Ill-NB), Revision C A508C25, SA-508 Class 3a Nozzle Forgings (Section ill-NB), Revision D B163C20, Thermally Treated Alloy UNS N06690 (Alloy 690) Tubing for Arkansas Nuclear One Unit No. 2 Replacement Steam Generators (Section ill-NB, SB-163, Code Case N-20-3), Revision G B163C23, Thermally Treated Alloy UNS N06690 (Alloy 690) Tubing for South Texas Unit No.1 Replacement Steam Generators (Section lil-NB, SB-163, Code Case N-20-3, Revision D B163C27, Thermally Treated Alloy UNS N06690 (Alloy 690) Tubing for Farley Units No.1 and 2 Replacement Steam Generators (Section lil-NB, SB-163, Code Case N-20-3), Revision E C523C04, Oerlikon Type OP-121TT Flux for Submerged Arc Welding (SAW) Low Alloy Steel, Revision C C523C05, Low Alloy Steel Electrode SFA 5.23 Class EM-2 and Oerlikon OP-121TT Flux for Submerged Arc Welding (SAW), Revision A CO51CO2, Carbon Steel Electrode SFA 5.1 Class E7018 for Shielded Metal Arc Welding (SMAW),
Revision C C055CO2, Low Alloy Steel Electrode SFA 5.5 Class E9018M for Shielded Metal Arc Welding, Revisions B :
and C Quality Insoection Procedures 3150, Tube insertion Inspection Requirements, Revision 16 l 3605, Hydroswage Equipment Calibration, Revision 01 l
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1 8573, Ultrasonic Examination of Welds, Revision 17 8141. Solvent Remover Liquid Penetrant Exam, Revision 01 8330, Radiographic Procedure, Revision 01 8609, Visual Examination of S/G Welds, Revision 01 Sandvik Control Procedures 5404, Visual Standards, Revision 2 5488, inspection of U-Bent Tubes, Revision 2 5482, inspection of Straight Tubes, Revision 1 i
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r Drawinas 6488E99, Steam Generator Upper Shell Detail Assy, Sub 3 6488E93, Steam Generator U-tube Insertion and Assy with Helium Leak Test, Sub ' 4 6488E68, Steam Generator Tube Plate and Lower Barrel Weld Assembly and Machining, Sub 4 6488E71, Replacement Steam Generator Outline, Revision 4 i Detailed Manufacturina Procedures DMP-0200, General Welding Specification, Gas Tungsten Arc Welding - GTAW ASME Boiler and Pressure Vessel Code Section IX and Section lil, Revision 22 DMP-0700, General Welding Repair Procedure ASME Boiler and Pressure Vessel Code Section IX and Section Ill, Revision 19 DMP-5524, Heat Treatment Procedure including Preheat, Interpass, Gouging, Hydrogen Baking and Posti Weld Heat Treatment, Revision 50 DMP-5563, Steam Generator Tubing Tube Tack Expansion, Revision 22 DMP-5881, Shop Control of Welding Materials ASME Boiler Pressure Vessel Code, Revision 20 DMP-6255, Foreign Object Control Procedure for a Complete Steam Generator, Revision 06
. DMP-6401, Steam Generator Hydraulic Tube Expansion, Revision 8 DMP-6420, Heat Treatment Procedure including Preheat, Interpass, Gouging, Hydrogen Baking and Post !
Weld Heat Treatment, Revision 03 DMP-6424, Instructions for Assembly and Inspection of Anti-Vibration Bars in Delta 94 Replacement Steam Generators, Revision 01 4
Nondestructive Examination Procedures
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8001, Guidelines for RT/UT Layout of Weld, Rev:sion 01 T
- 81C5, Solvent Remover Liquid Penetrant Exam, Revision 16 t
8106, Water Washable Liquid Penetrant Exam, Revision 13 8139, Water Washable Liquid Penetrant Exam, Revision 02 -
8567, Ultrasonic Examination of Austenitic Welds, Revision 05 t
20 1
8573, Ultrcs:nic Extminiti:n of Welds, Revision 17.
8141, Solvent Remover Liquid Penetrant Exam, Revision 01 8330, Radiographic Procedure, Revision 01 8609, Visual Examination of S/G Welds, Revision 01 Sandvik Control Procedures 5404, Visual Standards, Revision 2 i 5488, inspection of U Bent Tubes, Revision 2 5482, inspection of Straight Tubes, Revision 1 l
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Selected Generic Correspondence _on the Adequacy of Vendor Audits and the Quality of Vendor Products identifier Titig Information Notice 99-038 Metal-Clad Circuit Breaker Maintenance issues identified by NRC !
Inspectors
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NRC FORM 336 U.s.MJCLEAR REsULATORY COMeesslON
{249) 1. KEPORT NUMBER NRCM 1102, (Assigned by NRC, Add Vol., Supp., Rev.,
32n 32or and Addendum %% If any.)
CIBLIOGRAPHIC DATA SHEET
'8**"'***"'*"***"'"'
- 2. TITLE AND SuBriTLE NUREG-0040 Vol. 22, No. 4 Licensee Contractor and Vendor Inspection Stitus Report 3. DATE REPORT PUBLISHED Quarterly Report uoNTN ygAn October - December 1998 I March 1999
- 5. AUTMOR(S) 6. TYPE OF REPORT Quarterly
- 7. PERIOD COVERED (inctusve celes) j October - December 1998 8 PERFORMING ORGANIZATION . NAME AND ADORESS (rNRC. proude Dwson. omco or Regen, u 5 Nucasar Reguiefory C - . and mating acuress; acontractor proude name and meene edeess }
Division of Inspection Program Management Office of Nuclear Reactor Regulation
{
U. S. Nuclear Regulatory Commission W;shington, DC 20555-0001
- 9. SPONSOR'AG ORGANIZATION - NAME AND ADDRESS (rNRc type "same as abovoi # contractor. prowde Nec Dwman, osce orRegen, u.s Nucasar Reguntary comnnssen.
and meene ed$essi Same as above
- 10. SUPPLEMENTARY NOTES
- 11. ABSTRACT (200 words or Asss)
This periodical covers the results of inspections performed by the NRC's Quality Assurance, Vendor inspection and Miintenance Branch, that have been distributed i the inspected organlZations during the period October through December 1998.
- 12. KEY WORDS/DESCRIPTORS (Usf words orphrases mar ad assst researcreers a Aocating me sporfi 13. AVARA8iuTY STATEMENT Vendor inspection unlimited 14 SECURITYCLASSIFICATION (Tks Page) unclassified 5 Report) unclassified
- 15. NUMBER OF PAGES
- 16. PRICE EC' FORM 335 (249)
Pas form was electrorucally produced by Emo Federal Forms. Inr.
Printed on recycled paper Federal Recycling Program
64 LICENSEL CONTRACTOR Al%UTNDUR OCTZEER-CECEM.BER 1998 UNITED STATES SPECIAL STANDARD MAIL NUCLEAR REGULATORY COMMISSION POSTAGE AND FEES PAID WASHINGTON. DC 20555-0001 PERM NO G-67 OFFICIAL BtiSINESS PENALTY FOR PR VATE USE,5300
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