ML20236V479

Insp Rept 99900404/98-02 on 980615-17.Violations & Nonconformance Noted.Major Areas Inspected:Insp Conducted to Determine Quality Activities Performed by Westinghouse Conducted Under Appropriate Provisions of 10CFR50,App B

ML20236V479
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Issue date:	07/29/1998
From:	NRC (Affiliation Not Assigned)
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l U. S. NUCLEAR REGULATORY COMMISSION l

OFFICE OF NUCLEAR REACTOR REGULATION l

! Report No.: 99900404/98-02 Organization: Westinghouse Electric Company l Pittsburgh, Pennsylvania 15230

Contact:

Mr. Nicholas J. Liparulo, Manager Equipment Design and Regulatory Engineering Department Nuclear Industry Nuclear steam supply system design, compenents and Activity: services l

Date: June 15-17,1998 inspectors: Robert L. Pettis, Jr.,HQMB/DRCH Joseph J. Petrosino, HOMB/DRCH Melvin S. Holmberg, Region 111 Approved: 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 Enclosure 3 9908030268 990729 "

PDR GA999 EPfWEST 99900404 PDR

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1 INSPECTION

SUMMARY

The purpose of the inspection was to determine if quality activities performed by Westinghouse were conducted under the appropriate provisions of the Westinghouse 10 CFR Part 50, Appendix B, quality assurance program of record, and 10 CFR Part 21.

The inspection bases were as follows:

Appendix B, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to Part 50 of the C_ ode of Federal Regulations (10 CFR Part 50) 10 CFR Part 21, " Reporting of Defects and Noncompliance."

Westinghouse Energy System Business Unit (ESBU) Quality Management System (QSM) document (quality assurance topical report), Revision 2, dated May 29,1997.

During the inspection, the team identified the following instances where Westinghouse failed to conform to NRC requirements:

1.1 Violation Violation 99900404/98-02-01 was identified and is discussed in Sections 3.1 and 3.2 of this report.

1.2 Nonconformances Nonconformance 99900404/98-02-02 was identified and is discussed in Section 3.1 of this report.

Nonconformance 99900404/98-02-03 was identified and is discussed in Section 3.1 of this report.

Nonconformance 9990040A98-02-04 was identified and is discussed in Section 3.1 of this report.

2 STATUS OF PREVIOUS ,'NSPECTION FINDINGS Unresolved Item 99900404/98-01-01 (Closed)

During the January 12-13,1998 NRC inspection at the Westinghouse Repair and Replacement Services facility in Cheswick, Pennsylvania, the inspection team did not complete its review of an identified problem with a device used on Westinghouse 480-volt circuit breakers, direct trip actuators (DTAs). The DTA is typically used on model Type DB and DS breakers with either an Amptector or the newer Digitrip solid state overcurrent trip devices. The problem concerned a particular DTA model which incorporated a design change that used glued joints that held together 2

the core, permanent magnet and toc pole pieces. The Browra Ferry nuclear plant informed Westinghouse that a DTA had failed after several cycles. Westinghouse sold Amptectors and Digitrip devices equipped with DTAs as safety-related components. Eaton Cutler-Hammer manufactures the DTAs as commercial grade components and supplies them to Westinghouse who dedicates them for use in safety-related applications.

The team reviewed the circumstances and held discussions with the Westinghouse staff and determined that the subject DTAs (with glued joints) were supplied only to the Browns Ferry and Vermont Yankee nuclear plants for use in safety-related breakers; that Westinghouse informed Browns Ferry and Vermont Yankee; and that the particular glued joint method is no longer employed by the manufacturer. The team concluded that Westinghouse appropriately informed the applicable licensees and satisfactorily executed its 10 CFR Part 21 responsibilities. As a result, this

' issue is considered closed.

Unresolved item 99900404/98-01-02 (Ooen)

During the January 1998 inspection, the team asked whether Westinghouse reconciled its original j

seismic qualification report for a model DB-50 problem at the Indian Point-2 (IP2) nuclear power plant to determine whether the reduction in mass or the extra winding on the trip pan spring could adversely affect the operation of the breaker. Westinghouse stated that there was conservatism in ]

the design, but they had not yet reconciled the change to the qualification test report. They also stated that an upper limit of 31-ounces of force to raise the trip bar a quarter-inch is used as a limiting criterion at the factory and is published in vendor literature. However, Westinghouse did not publish a lower limit operational design criterion of 14-ounces which it uses intemally for manufacturing and refurbishment guidance at the factory to determine full operability. The Westinghouse staff stressed that the 14-ounce value is not a designed or analyzed number because they did not perform seismic testing to validate the value. Westinghouse stated that it plans to perform seismic testing in the near future to substantiate its lower limit guidance and that it does not believe that either the removal of the trip pads from the trip bar or the extra winding on  !

the trip pan spring is a major contributor to the failures of the IP2 DB-50 breakers. Westinghouse }

personnel informed the inspectors that it plans to perform seismic and evaluation tests to confirm that the trip pad removal and the extra winding on the trip pan spring do not contribute to unreliable operation of DB type breakers.

It is noted that although this potentially generic issue was identified to Westinghouse in October 1997, as of July 1,1998, it has not performed testing to substantiate the seismic adequacy of its circuit breaker functionality during a design basis event. Therefore, this matter will remain open and be reviewed during a future inspection.

Unresolved item 99900404/98-01-03 (Ocen)

' On May 7,1997, in a letter to the NRC, Illinois Power Company's (IPC), stated that five arc chutes (part number 56F417G02 for Westinghouse model 50DHP350,1200-ampere,4160-Vac circuit

' breakers) at its Clinton Power Station had inadequate clearance between the rear are horn of the arcing contact and the lower edge of the molded case of the arc chute, in a letter dated July 2, 1997, IPC reported the issue to the NRC under 10 CFR Part 21.

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9 L Westinghouse Trip / Conference report Q&ES 98-0519, prepared April 6,1998, indicated that Westinghouse personnel traveled to the Clinton site and satisfactorily assembled the arc chutes assemblies with their associated breakers. An associated Westinghouse letter (RRS/DSE(98)-087, dated March 23,1998) indicated that the problem was that "whomever attempted to assemble the arc chute to the breaker did not understand the assembly and had misconstrued the problem. The rear arc chute was never the problem, it was a bent seismic tie down bracket on the front of the arc

' chute." Subsequent to this inspection, the NRC inspectors contacted Clinton personnel and were informed that they did not agree with the Westinghouse conclusion. Additionally, Clinton personnel stated that when they first contacted Eaton Cutler Hammer, they stated they had never heard of a similar problem with the are chute / arc horn assembly, but later (in person) indicated that they were aware of the problem. The Clinton personnel also stated that the Westinghouse vendor manual for the circuit breaker did not address the are chute problem. As a result, this issue will remain open and will be reviewed during a future inspection to assess the potentially generic implications of an inadequate vendor manual that does not address proper assembly.

Unresolved item 99900404/98-01-04 (Ocen) l The team reviewed the adequacy and effectiveness of records associated with circuit breaker l issues,10 CFR 21.21.(a)(1) compliance, and related procedural requirements and guidance. The team also reviewed Westinghouse Energy Systems Business Unit (ESBU) Policy / Procedure 21.0,

! " identification and Reporting of Conditions Adverse to Safety," Revision 2, dated May 29,1997, and associated administrative instructions which included " retired" Westinghouse Nuclear l Technology Division's Administrative Manual (AM) F-3.0, " Potential Deviation or Failure to Comply .

(PD) Process," Revision 1, dated July 1,1995. The team reviewed several 10 CFR Part 21 related I i

component issues identified by various licensees as potentially reportable.

The team identified previously that no references existed to PDs in the procedure. Westinghouse l

agreed that there was no section in the procedure that delineated the process used to handle PDs.

During the last inspection the team noted that Westinghouse did not have any procedures or instructions which delineated the required actions for processing PDs. However during this inspection, the team.was informed that Westinghouse had been using the retired Administrative Manual (AM) AM-F-3.0, for PD disposition. Additionally, the team determined that it had not rc . iewed a balanced sample of PDs and potential issues (Pis) to appropriately evaluate the effectiveness of the implementation of Westinghouse's 10 CFR Part 21 program. As a result, this matter will remain open and will be reviewed during a future NRC inspection.

Unresolved item 99900404/98-01-05 (Closed)

This issue concerned the failure of DS-416 circuit breakers as a result of the dry lubricant being j removed from the mechanism during a refurbishment at a non-Westinghouse approved '

refurbishment facility. The team reviewed records, technical bulletins and cor. ducted discussions with the Westinghouse staff and determined that the two technical DS circuit breaker issues were not related and were appropriately dispositioned. As a result, this issue is considered closed.

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3 ' INSPECTION FINDINGS AND OTHER COMMENTS 3.1 10 CFR Part 21 Prooram lmolementation

. a. Insoection Scone The team evaluated Westinghouse's 10 CFR Part 21 Program implementation by performing a review of information relating to identified concems in domestic reactor plants such as recent

. Ice condenser component problems.

b. Observations and Findinas Ice basket segments (typically 12 feet in length) are coupled together using sheet metal screws arranged in pairs at 60 degrees intervals around the circumference of the ice basket. Twelve screws on each basket segment are installed at the coupling joint (24 screws total per coupling

' joint). NRC inspection reports 50-390/97-04 and 50-391/97-04 documented that broken ice basket sheet metal screws had been found in the ice melt tank at Watts Bar. Missing or defective screws at the intermediate or lower coupling joints could allow an ice basket to separate and become a missile hazard under blowdown loads created in a design basis loss of coolant accident. Therefore, defects and deviations from the design specifications for these screws represent a potential condition adverse to safety.

b.1 Evaluation Not Performed For Nonconforming Ice Basket Sheet Metal Screvyn

- On June 8,1995, Westinghouse received a fated copy of Tennessee Valley Authority (TVA)

Central Laboratories Service Technical Report, ' Watts Bar Nuclear Plant, Ice Condenser Basket Screws," dated June 2,1995. This metallurgical report identified that failed and nonfailed ice basket sheet metal screws contained cracks at the screw thread roots. Among the potential causes for the failed screws identified in this report was that the pre-existing

" quench cracks" had been created during the heat treatment process step of manufacturing.

Additionally, this report identified measured core hardness values in excess of the 32 Rockwell C scale range identified in the original design specification 678956. A Watts Bar adverse condition report WBPER950246, closed on July 25,1995, listed corrective action No.

3 as completed, in which TVA requested Westinghouse to evaluate the report. However, the

. NRC inspection team identified that no' evaluation of the June 2,1995 report had been performed by Westinghouse. Because this report identified conditions potentially adverse to safety that deviated from the original design specification for the sheet metal screws, the team concluded that an evaluation of these deviations as described in 10 CFR 21 was required.

Failure to perform an evaluation of the deviations has been identified as a violation of 10 CFR 21.21(a)(1). (Identified as part of 99900404/98-02-01) b.2 . Incomplete Assessment of Broken and Missina Ice Basket Screws On June 16,1995, Westinghouse initiated Customer Blanket Order No. 94NNX-112361 pertaining to the disposition of a sheet metal screw issue at TVA Watts Bar. This order required Westinghouse to evaluate field deviation report WATM-10356, dated June 15,1995, which i

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identified that 162 screw heads and 32 screws had been found in the ice melt tank following ice basket loading and weighing operations completed in 1994. Westinghouse documented the assessment of this condition in a letter to TVA, dated June 22,1995.

This assessment concluded that the conditions identified with the screws did not affect the function or structural adequacy of the ice condenser containment at Watts Bar. The basis for this conclusion was: a statistical evaluation of the likelihood of one or more missing screws at a single coupling, a calculation of the minimum number of screws required at a basket coupling, ed an evaluation of the potential projectile / missile hazards associated with a failed ice basket segment.

However, the team identified that the serviceability of the remaining screws, with respect to .

cracking identified in the TVA metallurgical report of June 2,1995, and the potential

- consequences of failed ice basket segments impacting the air handling units within the ice condenser, potentially introducing glycol into the sump affecting emergency core cooling systems, was not reviewed. As a result, the team considered the assessment incomplete for the areas noted. Failure to perform design reviews to complete the assessment of these areas is identified as a nonconformance of 1C CFR 50 Appendix B, Criterion Ill. (99900404/98-02-02) b.3 Identification Of Conditions Adverse To Safetv On or about June 1,1998, Westinghouse received Gelles Laboratories report, " Metallurgical Evaluation of Screws From Cook Power Plant Unit 1 Ice Baskets," dated May 1998. The report identified a number of potential causes for failed ice basket sheet metal screws including pre-existing cracks and hydrogen induced cracking, potentially related to failure to adequately bake the screws prior to plating, or the excessive hardness values measured for the failed sheet metal screws. Further, Westinghouse had initiated testing under contract with the D.C. Cook nuclear plant to evaluate the acceptability of in-service screws with these types of deviations from the original Westinghouse design specification 678956.

The team determined that the potential conditions adverse to safety (the failed screws at the D.C. Cook nuclear plant) had not been entered into Westinghouse's formal review program, as required by Section 111 of ESBU procedure 21.0, nor had the condition been identified as a potential deviation in accordance with Westinghouse administrative instructions. Failure to follow ESBU 21.0 and enter this condition adverse to safety into the formal review process, is a identified as a nonconformance of 10 CFR 50, Appendix B, Criterion V. (99900404/98-02-04) b.4 Lack Of Material Specifications For Screws in Ice Basket Desian Drawinas Purchase Order 546-AZG-221520-BEP, " Ice Basket- Ice Condenser," Change Notice No. 2, dated July 21,1975, identified design specifications 678956 and 952745 used in the i

construction of ice baskets for domestic reactors with ice condenser containments.

! Westinghouse design specification 952745, dated January 10,1975, applied to all domestic l Westinghouse designed plants other than the D.C. Cook and Sequoyah nuclear plants. Section 4 of this design specification required the material specifications for the ice basket components to be listed on the applicable ice basket design drawings.

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. 1 The inspection team identified that " Ice Condenser Plant Ice Basket Detail and Assembly"

. drawings 1145E10 and 1191E57, Revision 7, dated February 1984, lacked material requirements, hardness values and strength levels for sheet metal screws required by original 3

design specification 678956. Specifically the screws, described as piece No. 9, referred to j

" Note Y" for material identification. However " Note Y" on the drawings lacked a description of the screw material required for fabrication, the required material hardness, and the required material strength levels.

This condition appeared to contribute to a May 30,1990, Licensee Event Report (LER),

submitted to the NRC by Duke Power Company (DPC), that identified installed ice condenser sheet metal screws of improper material at the McGuire Nuclear Station (LER 369/90-07). DPC Problem Investigation Report 1-M90-0107, dated August 24,1990, proposed resolving the problem by initiating an editorial change to PPC drawing MCM 1201.17-0596, Sheet 001 (Westinghouse drawing 1191E57, Revision D-8, dated December 1987, Sheet 1 of 3) to specify the material required for screws used in the ice basket couplings. The LER also concluded that the cause of the event was due to an installation deficiency because of improper material selection. The change consisted of revising " Note Y" on the drawing to specify the material specification for the screws which was omitted.

Drawing Revision D-9, dated June 1990, added "AISI 1022 carbon steel; minimum yield strength 130 ksi." Stainless steel screws of 10-16 type 410 material were installed in the bottom and intermediate couplings of the ice baskets instead of specially treated 10-32 AISI 1022 carbon steel screws, as specified by Westinghouse. The screws, installed in 1988, were identified in 1990 by DPC design engineering personnel to be a problem since the Westinghouse manufacturing drawing contained no reference as to the screw material;

~ therefore, the wrong material was chosen when replacing the missing coupling screws.

The team also reviewed certificates of compliance for the vendors (LAMCO and Great Lakes Manufacturing Co.) that fabricated the ice baskets and screws respectively for domestic ice condenser plants. However, the certificates of compliance stated that they met the requirements of drawings 1145E10 and 1191E57. At the conclusion of the inspection, documentation for domestic ice condenser plants (other than D.C. Cook and Sequoyah) could not be located by Westinghouse which identified that the screws had been fabricated with serviceable material of the required strength and hardness. On June 17,1998, Westinghouse issued Corrective Action Request (CAR) CAR 98-1157 to document the lack of design requirements for the screws in the applicable drawings. Further, Westinghouse staff reportedly intended to locate the records which would demonstrate that the screws were fabricated using  ;

the appropriate materials or notify licensees to implement testing to confirm that the appropriate material was used. Failure to include design material and strength requirements for the screws is a nonconformance of 10 CFR 50, Appendix B, Criterion lil. (99900404/98-02-03) b.5 - Post-Insoection Information Suoolied bv Westinghouse '

Foilowing the inspection, Westinghouse supplied additional information to the NRC staff regading the ice condenser sheet metal screw issues identified by the team during the inspection. The information was documented in Westinghouse letter NSD-NRC-98-5728, dated

-June 26,1998.

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Westinghouse is presently investigating the issues under its 10 CFR Part 21 program (Pl 98-016, dated June 17,1998), as delineated in Westinghouse procedure ESBU-21, and estimates completion by mid-August 1998. On June 24,1998, Westinghouse verbally notified all of its ice condenser containment customers of the issues identified during the inspection and the discussions between Westinghouse and the NRC staff. Westinghouse's action plan to address the issues will include the following:

1. Perform a metallurgical evaluation s' 200 sheet metal screws supplied by D.C. Cook.

2; Review screw design including coatings and confirm the failure mode and root cause.

3. Review available reports, including previous metallurgical testing performed by Duke Power Company and TVA, and prepare technical position.

4.. CAR 98-1170, dated June 17,1998, which will address the review of both 1995 TVA reports (June 2 and June 19) including the consequences of ejected ice baskets hitting air handling units. Based on the information obtained, a supplement to the June 22,1995 Westinghouse report will be prepared.

5. Conduct test program to establish load carrying capacity of screws, as applicabic.

Westinghouse stated that preliminary information to date has identified quench cracks in only about 2 percent of the screws tested from D.C. Cook wh!ch corresponds to industry manufacturing data for similar screws.

On June 17,1988, Westinghouse initiated CAR 98-1157 to review discrepancies between Westinghouse design specification 952745, dated January 10,1975, and drawings 1145E10 i and 1191E57, Revision 7, and determine if the correct screws were manufactured and supplied in accordance with Westinghouse specifications. The review will include a review of vendor purchase order documentation, a revision to the specification / drawing to include the required l mechanical properties for the screws, and the issuance of a report.

3.2 Inadeauste Part 21 Proaram Requirements During the January 1998 NRC inspection, the inspection team attempted to discuss an issue i

regarding a problem with a shunt trip device on a Westinghouse model DB-75 type circuit breaker instal!ed at the Ginna nuclear power station. The problem was that the plunger in the r, hunt trip did not retum to the full out position following operation. When the licensee tested l- the assemblies in stock in the warehouse, the plunger would not return to the full out position j on 2 of the 10 assemblies. Westinghouse staff stated they had not been contacted by Ginna i

and as a result, the inspectors did not pursue discussions regarding the adequacy of Westinghouse's actions regarding 10 CFR Part 21 and closed the issue.

However, subsequent to the issuance of the report, the NRC was informed by the licensee, Rochester Gas and Electric Corporation (RGE), that on November 24,1997, RGE had contacted Westinghouse to discuss the issue. Westinghouse reviewed their records and 8

determined that RGE had informed them of the issue and provided the inspection team with correspondence that had been sent to another circuit breaker engineer and also to Westinghouse Nuclear Safety Department (NSD) personnel. The NSD personnel coordinate the execution of ESBU's program developed to implement the provisions of 10 CFR Part 21 and its implementing procedure, ESBU 21.0, and AM F-3.0.

The team determined that another retired ESBU instruction Manual Procedure (ET-B-1.0) states that its purpose is () define the process and responsibilities for assuring that a safety _

concem identified by, or referred to, Westinghouse Engineering Technology personnel are controlled and properly dispositioned in accordance with the QA program and to the criteria of 10 CFR Part 50, Appendix B, and 10 CFR Part 21.

A review of the related records, procedures and discussions with the cognizant staff -

determined that RGE staff called the Westinghouse engineer to inform him of the problem that .

had been experienced, ask whether similar occurrences had happened, and request guidance.

As a result, the Westinghouse engineer discussed the issue with another engineer, a senior circuit breaker shop technician and NSD staff to ask whether they were aware of similar problems. The issue was not pursued further since they were not aware of any similar problems. Additionally, although the NSD staff received a copy of the potential deviation, they did not formally document the issue in the PD/PI process for disposition in accordance with Westinghouse procedures. This example of a potential deviation that was not adequately dispositioned or controlled by Westinghouse is a violation of 10 CFR Part 21. (identified as part of 99900404/98-02-01)

c. Conclusions The inspection team identified a violation and three nonconformances associated with failures to adequately implement 10 CFR 50, Appendix B, commitments and 10 CFR Part 21 requirements on issues pertaining to ice condenser sheet metal screws at the Watts Bar and D.C. Cook nuclear plants. The inspection team concluded that the issues represented an implementation breakdown of Westinghouse's QA and 10 CFR Part 21 program requiremen'.s.

3.3 i-ntrance and Exit Meetinas I An entrance meeting was held on June 15,1998, in which the scope of the inspection was discussed with Westinghouse management and staff. On June 17,1998, an exit meeting was

' held with Westinghouse management and staff to discuss the inspection findings.

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e 4 PERSONNEL CONTACTED l The following represents a partial list of persons contacted during the inspection:

Westinghouse Electric Comoany N. Liparulo, Manager G. Dillon, Manager J. Galembush, Senior Engineer l

C. Scrabis, Advanced Technical Engineer G. Rao, Senior Engineer R. Miller, Senior Engineer OPEN AND CLOSED ITEMS ITEMS OPENED i

99900404/98-02-01 (VIO)

Failure to evaluate a June 1995 metallurgical report from TVA Watts Bar which identified quench cracked sheet metal screws, and the failure to process a potential deviation related to ice basket sheet metal screws in accordance with Westinghouse's Part 21 program.

99900404/98-02-02 (NON)

Failure to perform design reviews to complete the assessment of missing and broken sheet metal screws at TVA Watts Bar.

99900404/98-02-03 (NON)

Failure to follow Westinghouse procedure ESBU 21.0 and enter conditions adverse to quality into the fermal review process.

99900404/98-02-04 (NON)

Failure to include on the design drawings design material and strength requirements for sheet metal screws.

ITEMS dLOSED 99900404/98-01-01 (URI)

Review to determine the deportability of problems in DTAs.

99900404/98-01-05 (URI)

DS-416 type circuit breaker problems.

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O ITEMS TO REMAIN OPEN 99900404/98 01-02 (URI)

Testing to be performed to substantiate seismic adequacy of circuit breaker functionality during design basis events.

99900404/98-01-03 (URI)

Potentialinadequate vendor manual narrative of assembly cautions.

99900404/98-01-04 (URI)

A selection of Pts and PDs must be reviewed to assess the effectiveness of Westinghouse's Part 21 program.

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