IR 05000327/1986011

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Insp Repts 50-327/86-11 & 50-328/86-11 on 860203-07. Violation & Deviation Noted:Inadequacies in Development & Implementation of Maint Instruction MI-10.9 & Surveillance Instruction SI-227.1 Re Reactor Trip Breakers
ML20205P625
Person / Time
Site: Sequoyah  Tennessee Valley Authority icon.png
Issue date: 04/15/1986
From: Conlon T, Marlone Davis, Merriweather N
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20205P587 List:
References
RTR-NUREG-0737, RTR-NUREG-737 50-327-86-11, 50-328-86-11, GL-83-28, NUDOCS 8605210485
Download: ML20205P625 (24)


Text

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, APR 2 21986 ' Report Nos.: 50-327/86-11 and 50-328/86-11 i Licensee: Tennessee Valley Authority

6N 38A Lookout Piace j 1101 Market Street Chattanooga, TN 37402-2801 Docket Nos.: 50-327 and 50-328 License Nos.: DPR-77 and DPR-79 , Facility Name: Sequoyah 1 and 2 , Inspection Conducted: February 3-7, 1986 , Inspectors @ W/ // d2 8- /5 d

N. Merriweather, Team Leader Date Signed r??? FW /t- //- hr-M. J. Davis Date Signed ' ] Consultant: P. M. Chan, Lawrence Livermore National Laboratory J. Savage, Lawrence Livermore National Laboratory Accompanying Personnel-T. E. Conlon (February 6-7,1986) i Approved by W& V '/-/Y-M ' T. E. Conlon, Section Chief Date Signed Engineering Branch Division of Reactor Safety SUMMARY l t Scope: This special announced inspection involved 154 inspector-hours on site concerning licensee response to Generic Letter 83-28, Required Actions Based on

Generic Implications of Salem Anticipated Transient Without Scram (ATWS) Events.

' Areas inspected included: post-trip review; equipment classification; vendor interface and manual control; phst-maintenance testing; andfreactor trip system reliability.

Results: One violation and deviation was identified: Inadequacies in Develop-ment and Implementation of Maintenance Instruction MI-10.9 and ' Surveillance-Instruction SI-227.1, paragraphs 9c. and 9.f.; and Failure to Establish a Formalized Trending Program for Reactor Trip Breakers, paragraph 10.

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. REPORT DETAILS ' 1.

Persons Contacted Licensee Employees

  • H. Abercrombie, Site Director
  • P. R. Wallace, Plant Manager
  • R. C. Birchell, Mechanical Engineer
  • J. Blankenship, Information Officer
  • L. S. Bryant, Mechanical Maintenance Engineer Supervisor
  • C. R. Brimer, Manager, Site Services
  • R. L. Casteel, Nuclear Licensing Section

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  • D. L. Cowart, Quality Surveillance Supervisor l
  • E. A. Craigge, Industrial Safety Supervisor
  • J. T. Crittenden, Chief, Program Support Staff
  • H. D. Elkins, Jr., Group Supervisor Instrumentation
  • M. E. Frye, Compliance Engineer
  • M. R. Harding, Engineering Group Supervisor
  • G. B. Kirk, Compliance Supervisor
  • C. W. LaFever, Instrument Engineer Supervisor
  • R. Meadors, Nuclear Engineer

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  • L. L. McCormick, Regulatory Engineer

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  • L. M. Nobles, Operating and Engineering Superintendent
  • R. W. Olson, Modification Manager
  • B. Patterson, Maintenance Superintendent

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  • M. A. Purcell, Regulatory Engineer
  • M. R. Sedlacik, Electrical Modification Supervisor
  • M. A. Skarzinski, Electrical Maintenance Supervisor
  • T. Smith, Electrical Engineer
  • G. G. Wilson, Assistant Operations Group Supervisor H. R. Rogers, Shift Technical Adviser (STA), Compliance Engineer D. Reed, Records Storage Representative P. Wilson, Administrative Services Representative D. S. Richardson, Shift Engineer, Senior Reactor Operator (SR0)

F. H. Amburn, Modifications Group Engineer K. W. Vandergriff, Instrument Maintenance Representative Other licensee employees contacted included engineers, technicians, opera-tors, mechanics, security force members, and office personnel.

Other Organizations R. V. Matheison, Westinghouse Representative J. Turner, M0 VATS Representative , . , e

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NRC Resident Inspectors

  • K. Jenison, Senior Resident Inspector
  • L. Watson, Resident Inspector
  • Attended exit interview 2.

Exit Interview The inspection scope and findings were summarized on February 7, 1986, with l those persons indicated in paragraph 1 above.

The inspector described the areas inspected and discussed in detail the inspection findings.

No dis-senting comments were received from the licensee.

Inspector Followup Item 50-327, 328/86-11-01, Followup of the Licens-ee's Response to NRR for Post-Trip Review, paragraph 6.

. ' Inspector Followup Item 50-327, 328/86-11-02, Resolve Conflict Identi-fied in TVA's Response to GL 83-28 for Equipment Classification, paragraph 7.a.

Inspector Followup Item 50-327, 328/86-11-03, Review TVA's Methods for Revising CSSC List, paragraph 7.c Violation 50-327, 328/86-11-04, Inadequacies in Development and Imple-mentation of Maintenance Instruction MI-10.9 and Surveillance Instruction SI-227.1 paragraphs 9.c. and 9.f.

Deviation 50-327, 328/86-11-05, Failure to Establish a Formalized Trending Program for Reactor Trip Breakers, paragraph 10.

Unresolved Item 50-327, 328/86-11-06, Review TVA's evaluation of the RTB Shunt Trip Modification Using Actual Plant Parameters in Lieu of Nominal Values Specified by the Westinghouse Generic Design, paragraph 10.

The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspectors during this inspection.

3.

Licensee Action on Previous Enforcement Matters This subject was not addressed in the inspection.

4.

Unresolved Items Unresolved items are matters about which more information is required to determine whether they are acceptable or may involve violations or devia-tions. One new unresolved item was identified during this inspection and is discussed in paragraph 10.

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Background i In February 1983, the Salem Nuclear Power Station experienced two failures of the reactor trip system upon the receipt of trip signals. These failures were attributed to Westinghouse - Type DB-50 Reactor Trip System (RTS) . l circuit breakers.

The failures at Salem on February 22 and 25,1983, were believed to have been caused by a binding action within the undervoltage trip attachment ! (UVTA) located inside the breaker cubicle.

Due to problems identified with circuit breakers at Salem and at other nuclear plants, NRC issued Generic Letter (GL) 83-28, Required Actions Based ! on Generic Implications of Salem ATWS Events, dated July 8,1983.

J This letter required licensees of operating plants to respond to inter-mediate-tenn actions to ensure reliability of the RTS.

Actions to be performed included development of programs to provide for post-trip review, classification of equipment, vendor interface, post-maintenance testing, and RTS reliability improvement.

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The licensee, Tennessee Valley Authority (TVA), responded to GL 83-28 in a i letter dated November 7, 1983, with several supplemental responses to various sections of the GL.

- This inspection was performed to assess the adequacy of the licensee's current program, planned program improvements, and implementation of present ! procedures associated with post-trip review, equipment classification, vendor interface, post-maintenance testing, and reactor trip system reli-ability for Sequoyah Units 1 and 2.

The results of the inspection are discussed in detail in the paragraphs that follow.

6.

Post-Trip Review The licensee was requested in GL 83-28 to describe their program, proce-dures, and data collection capability to assure that the causes for unsched- , uled reactor shutdowns as well as the response of safety-related equipment, are fully understood prior to plant restart.

The licensee's response to GL 83-28 gives a detailed description of the program and procedures perti-nent to performing post-trip reviews. The inspector reviewed the licensee's , response, appropriate procedures, and interviewed responsible licensee . personnel to assess the adequacy of the licensee's program for post-trip ' reviews.

The results of this inspection are identified in the following . paragraphs.

By letter dated August 15, 1985, NRR provided the licensee with a prelimi-nary Technical Evaluation Report (TER) of the licensee's response to

Item 1.2 of GL 83-28, Post-Trip Review; Data and Information Capabilities for Sequoyah.

The TER identified three areas in which the licensee's ,

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response either failed to meet the review criteria of GL 83-28, or insuffi-cient information was provided to make a determination of the adequacy of the data and information capabilities.

The NRR letter requested a prompt response to the open issues described in the TER.

However, as of February 7,1986, the licensee has not provided this response (identified on the licensee's Management Action Tracking System as item number S0-420).

One problem area identified in the TER questioned whether data retention procedures ensured that post-trip review information packages were main-tained in an accessible manner for the life of the plant.

The inspector examined this item and confirmed that Sequoyah Administrative Instruction AI-7, " Recorder Charts and Quality Assurance Records" identifies in Enclosure 6 that AI-18, Package 18. " Reactor Trip Reports" are lifetime plant operations records. These Reactor Trip Report packages are maintained in the onsite permanent records storage vault.

Another problem area identified in the TER stated that all of the parameters specified in the TER for monitoring on the sequence of events and post-trip review reports were not monitored by these systems. The TER also questioned the performance characteristics of the plant process computer. The inspec-tor reviewed preliminary proposed software and hardware changes with Compli-ance Engineering personnel.

These proposed changes have not been submitted for approval to licensee management as of February 6, 1986.

Followup of the licensee's response to the TER and completion of the hardware and software changes identified in the submittal will be tracked as an Inspector Followup Item (IFI) 327, 328/86-11-01, Followup of the Licensee's Response to NRR for Post-Trip Review.

The post-trip review program is addressed and implemented by Sequoyah Administrative Instruction AI-18, Plant Reporting Requirements, Appendix B and File Package 18, Reactor Trip Report. The procedure states that in the event of a reactor trip or an unexplained power reduction, it is the respon-sibility of the Shift Engineer (SE) to analyze the cause and determine if operations can continue safely before returning the reactor to power.

The SE is a licensed senior reactor operator and supervises the operating shift.

Shift operations personnel initiate the Reactor Trip Report.

Section IV of the Reactor Trip Report requires that an evaluation of the event be per-formed in regard to Final Safety Analysis Report accident analysis by the SE or the Shift Technical Advisor.

At this time, an assessment for startup is made (Section V).

If the unit cannot be returned to service at this time, approval for startup must be granted by the Plant Manager, who may require Plant Operations Review Committee (PORC) to review the event and provide resolution prior to authorizing restart.

Administrative Instruction AI-2, " Authorities and Responsibilities for Safe Operation and Shutdown", requires permission from the plant manager, plant superintendent, or operations supervisor as a prerequisite to making the reactor critical.

This approval is documented in Section V of the reactor trip report by the SE.

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i Any unplanned manual or automatic reactor trips from power will require a , full reactor trip report and the assembly of an associated data package which includes control room strip charts, post-trip review report and sequence of events computer printouts.

! Following completion of the Reactor Trip Report, tne operations supervisor I j reviews the report and completes Section VII, Recommendations and Corrective

Action Followup, if actions are needed to be taken.

After review by the

Operations Supervisor, the trip report is sent to the PORC for an indepen-dent review of the adequacy of the trip review and corrective actions.

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i The Reactor Trip Report data package is retained in the onsite permanent

records storage vault for the life of the plant.

j ! The insp!ctor conducted a review of licensee procedures and verified that ] procedures were consistent with licensee responses to GL 83-28.

The inspector reviewed the Reactor Trip Report data packages generated for

seven reactor trips that occurred in 1985.

The packages were found to be j thorough and adequately documented the events.

The inspector also examined the licensee's post-trip review data collection

j capabilities and process computers.

Details of the inspection are as j follows: - i Sequoyah Units 1 and 2 have a Westinghouse PR00AC P250 process and alarm

j computer, which contains a sequence of events (SOE) program and a post-trip i review program.

The current system uses a half file disc which does not j have the data storage capacity to meet the TER guidelines for updating and i retaining post-trip information from approximately five minutes prior to the j trip until at least ten minutes after the trip. The licensee indicated that software and hardware changes are planned to provide a full file disc system

for increased data storage capability.

The plant computer is considered , ) non-Class IE, however, it is powered from an inverter.

! Each unit also has a DEC PDP 11/44 Technical Support Center computer system ! which provides the Safety Parameter Display System (SPOS) data collection ) and processing system associated with the NUREG-0737 upgrade of Emergency i Response facilities. This system enhances data collection capabilities.

Plant personnel preparing and/or reviewing the post-trip documentation , j appeared familiar with plant systems, equipment, and plant operation.

! ' Procedures provide for review of information from the trip and comparison with information derived from normal or expected operations and previous j shutdowns from similar situations.

Additionally, site procedures provide , for the identification of Reactor Trip Reports and accompanying data as

i Quality Assurance (QA) records and storage of the records in the permanent i i station records storage vault.

Within the areas inspected, no violations or deviations were identified.

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7.

Equipment Classification The licensee was requested in Section 2.1 of GL 83-28 to confirm that all components of the reactor trip system whose function is required to trip the reactor are identified as safety-related.

This identification was to be on documents, procedures, and information handling systems used in the plant to ' control safety-related activities including maintenance, work orders, and parts replacement.

In addition, the licensee was requested in Section 2.2 of GL 83-2f to describe their program for ensuring that all components of other saftty-related systems are also identified as safety-related on information handling systems used at the plant. The licensee's response to Sections 2.1 and 2.2 of GL 83-28 gives a detailed description of the program and procedures for safety-related equipment classification.

The licensee stated in their response dated November 7,1983, that TVA's Division of Nuclear Power identifies all components whose functioning is required to trip the reactor as safety-related.

Those components which include the reactor protection system, the solid state protection system, and all other components whose functions are defined as safety-related are now outlined in TVA's Operational Quality Assurance Manual (0QAM) as Criti-cal Systems, Structures or Components (CSSC).

The 0QAM is maintained as a corporate document.

This inspection was performed to verify that the licensee's program for equipment classification was adequate and consistent with the above re-sponse.

Interviews of licensee personnel and review of appropriate proce-dures and work documents revealed the following: a.

TVA's current program for maintenance of the CSSC list deviates from their submittal dated November 7,1983.

The 00AM has been replaced by the Nuclear Quality Assurance Manual and 9e CSSC lists for all TVA Nuclear Plants (BFN, SQN, WBN, and BLN) have been deleted and replaced with only criteria for preparation and maintenance of the CSSC list.

Thus, the CSSC list is no longer considered a corporate document as part of the 0QAM.

This change also makes each Site Director responsi-ble for assuring that the CSSC list is reviewed for completeness and accuracy and that a revision control system be established to review and verify each change to the CSSC list when it is made.

All of the above changes are contrary to the licensee's initial submit-tal to NRC in which the licensee stated that the CSSC list would be issued and controlled manually as part of the 00AM.

The inspector questioned whether the licensee had plans to revise their November 7, j 1983 response in which they described the 00AM as their CSSC list. No answer was provided.

The inspector informed the licensee that the Safety Evaluation Report for Items 2.1 and 2.2 has not been issued by NRR as of this date and that current information should be provided to NRR.

Therefore, this item is considered an IFI 50-327, 328/86-11-02, Resolve Conflict Identified Between TVA's in TVA's Response to GL 83-28 for Equipment Classification, . -

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b.

NQAM Appendix A, Part I identifies the criteria for inclusion of items on the CSSC list and responsibilities for maintenance and review of the list.

Part II of Appendix A identifies guidelines for inclusion of items on the CSSC list.

These requirements were used as.eference information in developing Sequoyah Nuclear Plant Administrative In-struction AI-39, Critical Structures, Systems, and Components.

Revision 2 dated January 9,1986, is the latest issue of this proce-dure.

In the body of this procedure, it references Standard Practice SQA-134 which contains the actual CSSC list and the Electrical Equip-ment Qualification List (10 CFR 50.49).

In reviewing procedures SQA-134 and AI-39, the inspector had a concern that SQA-134 was not a PORC reviewed procedure as is Administrative Instruction AI-39 and that AI-39 only references SQA-134 and not a specific revision of the procedure.

Subsequent discussions with QA revealed that a previous concern had been identified during an audit in which they recommended that certain Standard Practices be PORC reviewed.

In particular, SQA-134 had just recently been PORC approved on January 27, 1986.

However, because of a sign-off error, it had to be resubmitted on February 5, 1986.

c.

As required by the NQAM, the licensee has established a CSSC Committee to review and approve changes to the CSSC list. The CSSC Committee has met six times during the past year.

Minutes from these meetings indicate that several agenda items were closed prior to the specified actions being implemented and verified.

The inspector concluded that the licensee does not have a formalized system for tracking agenda items and for requesting changes to the CSSC list.

The licensee acknowledged this concern and indicated that they will consider devel-oping a more formal way to revise the CSSC list.

The inspector in-formed the licensee that this will be identified as an IFI 50-327, 328/86-11-03, Review TVA's Methods for Revising CSSC List.

d.

NQAM, Part V, Section 2.7 is the procedure which defines the responsi-bilities and requirements for the control and application of the Q-List. The Q-List identifies a list of features within the scope of TVA's quality assurance program.

This list is issued by the Office of Engineering as design drawings.

The Q-List for Sequoyah has been issued to the site, but is not being implemented.

Instead, the licen-see is using the CSSC list to classify work documents such as Work Requests and Work Plans.

The licensee indicated that the main reason for not implementing the Q-List is that it is not in a format that is very useful.

Hence, the Division of QA was tasked to develop a Q-List Specification to be used in developing a new Q-List. Discussions with responsible licensee personnel resulted in a commitment being made that three months after both Units (1 and 2) are returned to operation, the Q-List will be implemented. At this time, no schedule has been provid-ed for when Units 1 and 2 will return to operation status.

e.

The inspector randomly selected several Maintenance Requests (MRs) for the CSSC Reactor Protection System (System 99) for examination to verify that work activities were being classified as CSSC or non-CSSC e .. . . . . .. . . . .

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4 as required by procedures. The MRs reviewed and associated classifica- ' tions assigned are identified below: ' MR No.

Classification 100958 CSSC A-242400 CSSC , ' A-300157 CSSC A-526430 CSSC A-300806 CSSC < A-298461 CSSC A-298460 CSSC The above records demonstrated that the licensee was properly classifying MRs as CSSC or non-CSSC in accordance with SQA-134 and AI-39.

, ) Within the above area of equipment classification, no violations or devia-tions were identified.

G.

Vendor Interface and Manudi Control ! The inspector reviewed the licensee's response of November 7,1983, to GL 83-28 in regard to the comprehensive vendor interface program which the , licensee stated to be in place.

A cursory review of the Reactor Trip , Breaker Maintenance Instruction MI-10.9 revealed that one of the procedures J listed as references: Procedure No. TSO/5.0/5.0/2.14.0/1, titled, Reactor Trip Breakers, has incorporated the vendor's manual in its entirety.

] However, MI-10.9 itself failed to implement some of the vendor's l recommendations.

The inspector reviewed licensee's procedures in the area of assimilating other vendor information into the licensee's technical manuals. The inspec-

tor followed NRC Temporary Instruction TI 2515/64, Revision 1 and inspected . the following vendor manuals: a.

Contract No. 91934, N2M-2-29 ' Fuel Transfer System Technical Manual by Westinghouse:

The inspector verified that the eighth revision to the-subject vendor manual, known as "SNP Revision 8, October 29, 1985," is properly incorporated in the controlled copy located at the Vendor t Manuals Unit, as well as in the validated copy located in Electri-cal Maintenance.

b.

Contract No. 91934, N2M-2-6

Pressurizer by Westinghouse: The inspt:ctor verified that "SNP Revision 3, March 19,1984" is in the controlled copy.

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Contract No. 83550 and 92795 " Installation, Operating, and Service Instructions for Kumkle Safety and Relief Valves by Kumkle Valve Co: The inspector noted no revision or new information issued to date.

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Contract No. 91934, N2M-2-1 , Part Length Control Rod Drive Model 121J001 by Royal Industries: The inspector verified that Revisions 1 and 2 are in place.

The inspector also reviewed the following vendor manuals from vendors that had gone out of business or had relinquished certain product lines thus, leaving the applicant with no support on the hardware in question.

The inspector stayed within the guidelines of TI 2515/64 and reviewed only ' safety-related components. Specifically, the inspector reviewed the follow-ing vendor manuals who have gone out of business: a.

Contract No. 820498 Series 8800

Indicating Deviation Controllers by Beckman: The inspector noted that the applicant wrote a Design Change Request (DCR) #SQ-DCR-P-2236 to replace any of the subject equip-ment should it become necessary to replace or repair any of them.

A system exists in the licensee's design change process to evalu-ate alternatives ahead of anticipated need.

b.

Contract No. 92784 i Transmitters by GE/MAC:

The inspector noted that a previous DCR to replace transmitters by GE/MAC, who no longer manufactures transmitters, had been can-celled.

The applicant's rationale for the cancellation of the subject DCR was noted to be as follows: as per the DCR, a single selected replacement would be authorized for all GE/MAC transmit- , ters.

With the cancellation of the DCR, the applicant can react to each replacement on a case by case basis and order a suitable replacement for each specific application.

The inspector noted , that this would be a more flexible and rational approach as opposed to a blanket replacement policy.

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Contract No. 54752 Electrical Distribution Equipment Factory Order 11799 by i Arrow Hart, Inc.:

The inspector noted that the type of distribution equipment in question can be readily replaced by similar products of other '

vendors.

The inspector reviewed the Arrow Hart vendor manual in the Vendor Manual Unit, and noted that the controlled copy is kept up for ready references.

The inspector examined the following licensee's procedures on the vendor manual program and conferred with applicant's personnel to augment some findings: Administrative Instruction AI-23, " Vendor Manual Control," has gone - through its 21st revision as of November 1985.

The procedure defines the general techniques and responsibilities for the control and revi-sion of vendor manuals.

It covers vendor manual control from its receipt, to the dissemination of the manuals and its updates to the discipline cognizant engineer.

The inspector verified several manuals for conformance to procedure at the Drawing and Vendor Manual Unit at Sequoyah.

Procedure No. 1707.03.04, " Vendor Manual Program," Revision 0, dated - December 1984.

This procedure ensured the timely implementation of safety-related equipment ' vendor manuals, and their updates, and ful-filled the commitment made in the licensee's response of November 7, t , 1983, to NRC GL 83-28.

The procedure covered the responsibilities for the receipt and control of vendor manuals and delineated the duties of t the various licensee's organizations in regard to the revision and l utilization of safety-related vendor manuals.

i Within the areas inspected in vendor manual control, no violations or j deviations were identified.

9.

Post-Maintenance Testing The inspector reviewed the licensee's post-maintenance testing procedures ' and activities to verify that the requirements of GL 83-28 were being met and that the commitments in the licensee's response were being implemented at the Sequoyah Nuclear Plant.

The inspector examined procedures and completed maintenance records, witnessed a complete maintenance and post- , maintenance testing of a reactor trip breaker, and a timing test of a motor operated Volume Control Tank (VCT) Outlet Isolation Valve Level Control.

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adequacy of the licensee's post-maintenance test program.

The results of the inspection are as follows: a.

Review of MI-10.9 Reactor Trip Breaker Maintenance, Revision 10 (1) Several references mentioned on page 1 of the subject procedure were not known to the cognizant electrical engineer.

It was later + i determined that one of the references was in error.

, (2) A Revision 11 is now being written. This new revision will list a , i different combination of references which are intended to be current and applicable.

(3) Section 6.1 and 6.2 cover inspection of MG set circuit breakers.

Section 6.3 begins the inspection, lubrication, and testing of reactor trip and by-pass circuit breakers.

Revision 11 will clarify the application of MI-10.9 by sub-dividing the MI into two < parts: MI-10.9.1 for Reactor Protection System 99, and MI-10.9.2 for MG set Circuit Breakers System 85.

i (4) MI 10.9, Section 6.3.9.1 refers to Steps 5.3.9.2 through 5.3.9.5.2 which do not exist. Section 6.3.12.3.10 refers to Step 5.3.12.3.5 ~i which does not exist. The cognizant engineer subsequently identi-fied other similar errors (e.g., Section 6.3.16.2) and corrected ] them by a temporary change notice.

The maintenance being wit-nessed by the audit team on February 4,1986, was the first time l this typographical error had been noted and this raised the l concern on the accuracy of the procedure and the competence of the j craftsmen using the procedure since this same procedure had been completed four times in the past.

(5) Section 6.3.12 calls for replacement of the undervoltage trip attachment (UVTA) after 1250 cycles of operation. This cannot now , be done because the breakers do not have operation counters and no systematic record has been kept of UVTA operations by individual circuit breakers. The reasons are:

(a) The Westinghouse 08-50 reactor trip circuit breakers do not have unique serial numbers. The number stamped in the serial block on the circuit breaker nameplate is the Westinghouse ! manufacturing shop order number and several breakers have the ' same serial number (e.g., either 27-Y-1981B, 27-Y-1998B or < 27-Y-1998B1).

, (b) Maintendnce work has been identified by circuit breaker panel and cubicle position number.

This does not guarantee that ,

the individual positions have always contained the same breaker, because the position number does not follow the breaker.

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[ i } (c) Position numbers are as follows for Unit 2, Panel 2-L-116: f I i i RTA 2 BKRC-099-KG/320T ! ! RTB 2 BKRC-099-KH/320T l

BYA 2 BKRC-099-KG/319T i ' ' BYB 2 BKRC-099-KH/319T [ 4. (d) The inspector concluded that a unique TVA number was never f i assigned to the circuit breakers, but at the exit meeting,

the licensee stated that a unique numbering scheme would be ! j implemented to identify the reactor trip breakers.

The i ! inspector was advised that the cognizant engineer initiated

work orders in September 1985, to enscribe unique serial j i l numbers on each circuit breaker frame to alleviate this i problem.

The new Revision 11 of MI-10.9 will require that

these unique numbers be included on every page containing ! work and test records.

q j (6) Section 6.3.12.2.2 requires that the UVTA be replaced if the

dropout voltage test does not fall between 14.4 and 28.8 Vdc or if l

the dropout voltage differs more than 5 volts from the reference

voltage.

The reference voltage is not stated in the text, and ! there is no prerequisite in Section 3 that it be identified in '

Maintenance Request work instructions.

In the past, this step on f i page 39 of MI-10.9 has been left blank to be filled in later by the cognizant engineer, which is a violation of standard proce- ,

dure.

The inspector was advised that this deficiency would be , properly addressed in Revision 11 of MI-10.9.

! - j (7) Section 6.3.20 calls for five UV trips prior to returning the ! ! reactor trip circuit breaker to service after maintenance.

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Westinghouse recommends ten UV trips in NSD-TB-83-02, Revision 1,

page 5.

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I (8) NSD-TB-83-02, Revision 1 recommends that the Westinghouse 08-50 l circuit breakers be lubricated at intervals no greater than 200 ' cycles of the breaker. The inspector was unable to correlate this - l figure with the present six month maintenance period due to lack ! of breaker-by-breaker operation cycle records which include !

maintenance cycles, i ! ! i (9) Step 6.3.13.2, page 41 of Appendix A specifies 1/13" to 1/8" I clearance.

This was a typographical error, it was stated that it i ! should be 1/32" to 1/8" instead.

. i (10) Some figures in MI-10.9 are poorly reproduced (e.g., Appendix B, i i Figure IA) and are difficult to read.

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(11) Despite the above short comings (i.e.

Items 4, 5, 6, and 9 above) in Revision 10 of MI-10.9, this procedure was used and signed-off on several preventive maintenances in 1985 after the issue date, without Temporary Change Authorization (e.g., PM #1032-099 dated October 15, 1985, PM #1033-099 dated October 15, 1985, PM #1038-099 dated October 29, 1985, PM #1034-099 dated October 9, 1985, PM #1035-099 dated October 9,1985, PM #1036-099 dated October 29, 1985, and PM #0836-099 dated September 25,1985).

t (12) The 20 ounce calibrated weight used in Step 6.3.11.8 measured 19.2 ounces on the calibrated Chatillon scale.

However, the weight itself lacked any label, marking, or calibration sticker to identify it as a calibrated weigh + for the use in reactor trip breaker maintenance, b.

Witness of MI-10.9 Maintenance Activity on Unit 2 Reactor Trip Breaker A (27-Y-1981B-RT-3) Installed in Panel 2-L-116 The inspector witnessed an execution of MI-10.9 per PM #1035-099 on February 4, 1986, until the Work was stopped due to inability to complete Step 6.3.9.1 as a result of a typographical error in the step directing the craf tsmen to proceed to another step that does not exist in the procedure.

A temporary change request #86-268, dated February 5,1986 corrected the subject errors in the instruction and i the maintenance was completed on February 5, 1986.

Steps 6.3.1.2 and 6.3.1.3 could not be performed due to the control voltage being tagged out. Step: 7.2 and 7.5 were not performed because of a broken terminal board cover and the inability to complete Step 6.3.12.2.2.b because the reference voltage was neither known or given in the work order informa-tion. Some additional inspectors observations are as follows: (1) Step 6.3.12.3 has been designated N/A despite the fact that there is no way to verify the accumulated count of UVTA cycles.

(2) It was stated to the inspector that the UVTA coils have been replaced more often than 1250 cycles due to failure to pass the tests of Step 6.3.12.

The presently accepted values of accumu-lated cycle count have been subjectively estimated by the cogni-zant engineer and other personnel based on experience and judgement.

(3) Step 6.3.11.3 requires recording the force needed to trip the breakers, but there is no place in 6.3.11.3 on page 39 to record the measured force.

The step was signed off despite this omission.

(4) Discussion between the inspector and licensee personnel concerning the sources of some of the testing parameters revealed uncertainty regarding the sources for the TVA parameters.

The licensee stated

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that they are not necessarily committed to follow manufacturer's recommendations (e.g., replace UVTA coil after 1250 actions, five vs. ten post-maintenance electrical UVTA trips).

(5) The inspector understands that the five 08-50 spare breakers included shunt trips and counters.

The inspector feels some concern whether the equipment is fully qualified, or could be demonstrated to be fully qualified.

The Westinghouse on-site representative stated that Westinghouse could confirm if the equipment is qualified.

(6) The licensee stated that all spare breakers would be tested per MI-10.9 before use.

(7) The inspector understands that there is an in-progress modifica-tion to install shunt-trip controls to work in conjunction with the present UVTA controls.

This work is expected to be completed prior to start-up.

c.

Concluslum Drawn From Review of HI-10.9 10 CFR 50, Appendix B, Criterion V requires that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings of a type appropriate to the circumstancies dnd shall be accomplished in accordance with these instructions, procedures, or drawings.

In addition, those instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfac-torily accomplished.

Considering the above, in light of all the discrepancies identified with Maintenance Instruction MI-10.9, the inspectors concluded that this instruction is inadequate and is not being properly implemented.

A summary of the concerns is as follows: (1) The procedure did not include vendor recommendations as specified in Westinghouse Technical Bulletin NSD-TB-83-02, Revision 1 and the licensee could not provide adequate Justification for not implementing the recommendations.

(2) The technical review of the procedure appeared to be lacking due to the number of significant discrepancies found by the inspector, such items as: (a) The procedure contained poor quality drawings which were difficult to read.

(b) Typographical errors existed in the procedure, each by themselves enough to cause the technician to stop work on the procedure.

However, records indicate the procedure was performed several times with existing error _ _.

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, (c) The voltage used to energize the undervoltage coil during maintenance was greater than the actual voltage inside the reactor trip breaker cabinet.

(d) The procedure calls for replacement of the UVTA after 1250 . operations.

However, TVA does not have a systematic way to track the number of operations. Thus, the number of previous UVTA operations has always been unknown to the electrici technicians.

, (3) The licensee did not have unique identification numbers for the reactor trip breakers prior to October 1985.

Thus, the mainte-nance history on each breaker is not readily traceable.

(4) The importance of strictly adhering to procedures appeared to be

lacking by the electrical technicians.

d.

Review of SI-227.1, Post-Maintenance Testing of Reactor Trip Breakers l The inspector's review of Revision 2 of SI-227.1 resulted in the l following observations: ! (1) There was some uncertainty about the source and justification of < I the 0.2 second response time acceptance parameter stated on l page 1 Section 6.0.

. (2) The lack of a specific instruction to check and verify the ready ' state of the trip channels caused the failure of the first wit-nessed test because Channel II was locked out due to the shutdown state of the unit.

This prevented the completion of the test because the 48 volt (actual 43 vdc) signal supplied to the visi-i corder originated from Channel II.

! (3) The lack of a specific statement in Instruction 9 to " depress and j hold" the Channel I manual function test switch resulted in a failure to test during the second witnessed test.

i e.

Witness of SI 227.1 Timing Test Activity on Unit 2 RTB B The inspector's observations are as follows: (1) The witnessed test of February 5,1986, was not finished because the expected voltages did not appear.

It was noted that the test was done at a 43 Vdc level instead of the nominal 48 Vdc.

This ' raised the concern that the performance of the circuit breaker l maintenance as per MI-10.9, where 48 Vdc was used, may not be a

. fair representation of the performance of the reactor trip ! I breaker.

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' - ~..~ (2) The results of (1) above were later determined to be caused by 'a., locked out 48 Vdc channel, due to the station shutdown condition, and a misconnection to the wrong terminal in cabinet L-116.

The X18-1 (L-116) connection was made to a terminal block instead of

the proper relay coil terminal as intended.

This was determined later to be due to unclear labeling in the cabinet, and the inexperience of the instrument mechanic.

(3) It was later learned that the personnel assigned to the first witnessed test had little or no familiarity with the test proce-dure and prerequisites.

I (4) The inspector observed during the second witnessed test of February 6,1986, that the replacement crew of instrument mechan-ics seemed more knowledgeable of procedures and proceeded more ' efficiently than the first crew.

, ! (5) A temporary change #86-286 on February 6,.1986, changed' the hook-up details in Instruction 4.0 from Channel II test jack to ' Channel I test jack.

And, in Instructiun 9.0, interchanged the " depress" sequence of the Channel. I and II function test switches.

(6) The proper connections were made to the X1B-1 (L-116) relay coil.

(7) The first test run on February 6,1986, failed because the func-tion test switches were not operated properly due to unclear i Instruction 9.0.

I The first test run was not successful because the function test i switches were not depressed and held.

After some interpretation of the procedure by the mechanic, the test was successfully run, operated correctly as follows: " Depress and hold Channel II, then

1 depress Channel I."

The oscillograph traces were acceptable.

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(8) The recorded test time was 0.08 seconds which was within the t 0.2 second limit of acceptability.

f.

Conclusions Drawn from Witnessing Surveillance Instruction ! SI-227.1, Post-Maintenance Testing of Reactor Trip Breakers

10 CFR 50, Appendix R, Criterion V requires that activities affecting i quality shall be prescribed by documented instructions, procedures, or drawings of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or

drawings.

Contrary to the above, the following discrepancies were

, identified during the performance of Surveillance Instruction SI-227.1: , e t I (1) The procedure did not provide verification that appropriate ! initial conditions were established prior to performing the , procedure.

Consequently, the test could not be successfully performed on February 5, 1986.

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(2) Step 9.0 of the procedure. was vague in explaining what action was required by the instrument mechanic.

It required the mechanic to make an interpretation of what was required.

(3) Test personnel were not familiar with the procedure and methods t used to identify components inside reactor trip breaker cabinets.

This resulted in miswiring of test leads and misinterpretations of the procedure.

The concerns identified in paragraphs 9.c and 9.f above, constitute Violation 50-327, 328/86-11-04, Inadequacies in Development and ' Implementation of Maintenance Instruction MI-10.9 and Surveillance Instruction SI-227.1.

, g.

Motor Operated Valves (MOV) i ^ The cognizant electrical engineer described the M0V maintenance system i and exhibited applicable MI to the inspector. The inspector nuted that a total of about 300 environmentally qualified MOVs are distributed between Units 1 and 2 (i.e.,114 valves each), and that about 72 MOVs are common to both units.

A preventative maintenance program is now being worked up to establish for each valve a data base for future trending. At subsequent refuel-ing outages, a selected proportion of MOVs (say 25 percent) will be scheduled to be retested and the pertinent test data recorded.

The inspector's observations are as follows:

(1) Maintenance / test teams consist of a machinist and an electrician who have been trained in a TVA three-day Limitorque Valve Operator training class.

(2) All valve operators on the CSSC list will be tested by the Motor Operated Valve and Test System (M0 VATS) by a M0 VATS company i technician, after the valves have been maintained and setup for the test by TVA craftsmen per MI-10.43.

The M0 VATS test data includes the collection and storage of test data in a " Bubble" memory device including torque switch balancing, spring pack calibration, thrust valve setting and five graphic printcut charts.

(3) Following the M0VTAS test, a limit-to-limit operating time test (stroke time) will be performed by operation.

A person from the control room operating crew will observe and record the operating time, using a stop watch.

(4) A normal complete MOV maintenance schedule will span from one to five days, depending on the specific valve and plant operating needs.

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~ (5) A typical maintenance s,chedule is understood to include the

following: ) (a) A maintenance request (MR) is written and approved by the Electrical Maintenarice (EM) section.

(b) The Planning Section collects applicable instructions and

assembles them with other pertinent test documents.

(c) The QA,section reviews the MR.

(d) JAll N.plicable documents and instructio'ds 're collected in a a MR work package prior to performance by the craftsmen.

- (e') Each craftsman has a looseleaf refer 5nce book which contains , copic. of all instructions for all, procedures.

(6) All' steps in the maintenance'sc'enarios aYe guided by SQM-2,^ the extensive TVA Standard Practice Sequoyah Nuclear Plant Maintenance Management System.

This system covers all aspects of the mainte-nance process (e.g., Initiation of MR/WR documents, package contents, tag / card entries, planner entries, foreman responsibili-ties, QA review, job safety analysis, work authorized by craftsmen I ' section, work complete, post-maintenance test complete, status trading, maintenance history records, etc.). ~

h.

Witness of M0 VATS Test on VCT Outlet Isolation Valve Level Control 1-FCV-62-132-5403-121 Size SMB00 Order 347352 Ser, 120610 The inspector's observations are as follows: (1) The MOVATS tests were performed by a M0 VATS Co. Senior Technician using input transducers installed by the TVA crew.

Test ID is 020586-1-FCV-62-132.

(2) The test data was recorded on a memory device test instrument that displayed the input data versus time traces on a cathode ray tube screen and stored it in test device memory.

(3) The test data display was manipulated manually after recall from storage to select the~ desired pattern to place in the " Bubble" memory.

(4) After recording the. desired displays in the " Bubble" memory device, the memory. device wts given to the cognizant engineer for his permanent records.

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(5) The data recorded in the " Bubble" was later plotted out on five charts which represent: Out-of-seat motor current, and motor running current, and 97% full open position.

Closed-to-open stroke motor running current.

Spring-pack deflection and load cell curve (for calibration).

, Open-to-closed stroke motor running current and seating current.

Open-to-closed stroke motor current and torque switch and Limit Switch operation.

(6) The operating stroke time recorded by M0 VATS was 3.986 seconds.

(7) The operating stroke time observed by the control room operator with a stopwatch was approximately the same.

(8) There was some misunderstanding on the part of the control room operator as to whether the M0 VATS timing test was to be performed.

The matter was resolved by walkie-talkie when the inspector pointed out that the test was called-for in the MR. This incident raised the concern that the control room operator may not be familiar with this test procedure.

(9) The M0 VATS witnessed test was performed at the end of a regular MOV MR No. A-548687.

The working documents referenced during the completed MR work were: MI-10.43 Revision 2 MI-11.2 Revision 16 MI-10.46 Revision 0 MI-6.20 Revision 6

  • M9AI-7 Revision 6 MOVATS Data Sheet No Number MR Supplement Form TVA-64360 MR Supplement Form TVA-6446G
    • SI-166.6 Not Known (10) The MR work documents mentioned the following temporary changes which were not present in the MR packet given to the inspector.

TC 85-15, 85-1433, 85-1586, 85-1604, 85-1680, 86-084

  • Cable Terminations, splicing, and repairing of damaged cables, j

Not examined by the inspector.

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    • Stroke test.

Not examined by the inspector.

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1.

Miscellaneous Inspector General Observations (1) The inspector was advised of two differing philosophies: TVA does not utilize procedure walk / talk throughs because the written mis are considered to be sufficiently complete and accu-rate that any mechanic can follow them and execute accurate and correct work.

Some TVA sections have training goals which plan to aasign crafts-men to classes several days per year to enhance their capabilities and efficiencies.

(2) The value of recording trending data is acknowledged, but has not been uniformly and systematically collected.

In some cases, the effective use of data is hampered by missing or incomplete specif-ic equipment identification.

At the exit meeting, the applicant was reminded of the importance of trending of parameters such as: insulation resistance, response time, trip. torque and UVTA dropout voltage, as a means to predicting the degradation of the reactor trip breakers.

(3) The inspector was concerned over a frequent lack of specific knowledge concerning the sources and justifications for some acceptance parameter values used, and some manufacturers' recom-mendations were not implemented.

Within the area examined, two examples of one violation were identified.

10.

Reactor Trip System Reliability In a letter dated November 7, 1983, TVA committed to develop a program for trending of parameters to assess any possibility of performance degradation for the reactor trip breakers.

The licensee indicated that the trending program would consist of the following: - The compilation of all maintenance activity records into a history life.

The use of the Nuclear Plant Reliability Data System for breaker - failure data.

An MR system.

- I However, as of February 3-7, 1986, the licensee has failed to establish a j formalized method for trending of reactor trip breaker parameters.

Al-though, the licensee'_s cognizant electrical engineer was informally J

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recording the dropout voltage measurements for undervoltage trip attach-ments.

This is not considered a formalized trending program.

In addition, all parameters which could effect the breakers performance are not being recorded by the licensee.

Considering that the licensee committed to implement a trending program in their response dated November 7,1983, and failed to do so, the above concern constitutes a deviation from an NRC commitment and is identified as Deviation 50-327, 328/86-11-05, Failure to Establish a Formalized Trending Program for Reactor Trip Breakers.

GL 83-28, Item 4.3 required licensee's of Westinghouse reactors to modify their plants by providing automatic reactor trip system actuation of the breaker shunt trip attachments. TVA responded to this item by committing to implement the WOG generic design package at Sequoyah.

The licensee has developed a Work Package to implement the WOG generic design for the automatic shunt trip.

The inspector requested the design evaluations / calculations to confirm that components selected were based on actual plant conditions, such as the measured voltages on the UVTA inside the breaker cubicles.

The licensee indicated that actual plant parameters were not considered and that an evaluation would be performed to verify the adequacy of the design.

This concern is identified as unresolved item 50-327, 328/86-11-07, Review TVA's Evaluation of the RTB Shunt Trip Modifi-cation Using Actual Plant Parameters in Lieu of Nominal Values Specified by the Westinghouse Generic Design.

Within the areas examined, one deviation was identified.

11. Procedures Reviewed SQNP Administrative Instruction AI-27, " Shift Technical Advisor" - Rev. 7, September 24, 1985 SQNP Administrative Instruction AI-18, " Plant Reporting Requirements" - Rev. 42, November 26, 1985 Appendix A, File Package 18 " Notification and Licensee Event Report (LER)" i Appendix B, File Package 18 " Reactor Trip Report" SQNP Administrative Instruction AI-2, " Authorities and Responsibilities for Safe Operation and Shutdown", Rev. 25, November 7, 1985 SQNP Administrative Instruction AI-4, " Plant Instructions - Document con-trol", Rev. 52, December 24, 1985 SQNP Administrative Instruction AI-7, " Recorder Charts and Quality Assurance Records", Rev. 36, April 19, 1985 -

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SQNP Standard Practice SQA-21, "0nsite Independent Review (Plant Operations Review Committee)", Rev. 10, November 25, 1985 SQNP Standard Practice SQA-84, " Reportable Occurrences", Rev. 4, January 28, 1984 SQNP Standard Practice SQN-146, " Shift Technical Advisor (STA)", Rev. O, November 3, 1983 ' SQNP Operations Group OSLT-1, " Nuclear Generating Plant Operator Training Programs", April 25, 1985 Procedure No. 1707.03.04, " Vendor Manual Program," dated December 21, 1984 Procedure No. 0601.01, " Review, Reporting, and Feedback of Operating Experi-ence Items," dated June 4, 1985 Procedure No. NQAM, Part II, Section 2.1, " Plant Maintenance," dated October 12, 1984 Maintenance Instruction MI-10.9, " Removal, Inspection, Lubrication, and Replacement of Control Rod Drive MG Set, Reactor Trip, and Reactor Trip Bypass Circuit Breakers, Six Months, Units 1 and 2," Revision 7, dated May 11, 1984 Surveillance Instruction SI-227.1, " Post-Maintenance Response Time Test of Reactor Trip Breakers RTA and RTB," Revision 2.

Administrative Procedure AI-23, " Vendor Manual Control," Revision 21.

Administrative Procedure AI-25, " Drawing Control After Unit Licensing," Revision 12.

Standard Practice SQA-125, " Controlled Documents", Revision 5, dated November 15, 1985.

Administrative Procedure AI-19, Part III, " Plant Modifications," Revision 12.

Surveillance Instruction SI-268, " Verification of P.4 Interlock," Revision 5.

Standard Practice SQM-2, " Maintenance Management System," Revision 16.

Standard Practice SQA-134, " Critical Structure, Systems, and Components (CSSC) List," Revision 7.

Maintenance Instruction MI-10.43, " Procedure for Testing of Motor Operated Valves Using the M0 VATS-2000 System," Revision 2.

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Maintenance Instruction MI-11.2, " Motor Operated Valve Adjustment Guidelines Units 1 and 2," Revision 16.

Maintenance Instruction MI-10.46, "Limitorque, Motor Operated / Control Valve," Revision 0.

Maintenance Inspection MI-6.20, " Configuration Control During Maintenance Activity," Revision 6.

Special Maintenance Instruction SMI-0-317-22, " Field Verification of Revised Limitorque Electric Motor Operator Data," Rcvision 0.

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