ML20205P625
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| ML20205P625 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| 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) | |
See also: IR 05000327/1986011
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S Etc UNITED STATES
, [S NUCLEAR REGULATORY COMMISSION
d' E f[ ^ REGION 11
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101 M ARIETTA ST RE ET, N.W.
ATL ANT A, GEORGI A 30323
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, APR 2 21986
Report Nos.: 50-327/86-11 and 50-328/86-11
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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
1 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)
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Approved by W&
T. E. Conlon, Section Chief
V '/-/Y-M
Date Signed
Engineering Branch
Division of Reactor Safety
SUMMARY
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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
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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 l
Trending Program for Reactor Trip Breakers, paragraph 10.
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REPORT DETAILS
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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
l *D. L. Cowart, Quality Surveillance Supervisor
*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
! *L. M. Nobles, Operating and Engineering Superintendent
; *R. W. Olson, Modification Manager
*B. Patterson, Maintenance Superintendent .
! *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
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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- .
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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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5. Background
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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) .
circuit breakers. l
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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.
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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.
; The licensee, Tennessee Valley Authority (TVA), responded to GL 83-28 in a
i letter dated November 7, 1983, with several supplemental responses to
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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-
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nent to performing post-trip reviews. The inspector reviewed the licensee's
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response, appropriate procedures, and interviewed responsible licensee
personnel to assess the adequacy of the licensee's program for post-trip
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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
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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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, 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.
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Following completion of the Reactor Trip Report, tne operations supervisor
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-
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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 2
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.
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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.
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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 l
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. I 1 I
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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
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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. l
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 i
for Equipment Classification, l
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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
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as required by procedures. The MRs reviewed and associated classifica-
tions assigned are identified below:
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MR No. Classification
100958 CSSC
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A-242400 CSSC
A-300157 CSSC
A-526430 CSSC
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A-300806 CSSC
A-298461 CSSC
A-298460 CSSC
The above records demonstrated that the licensee was properly classifying
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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
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Breaker Maintenance Instruction MI-10.9 revealed that one of the procedures
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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
4 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:
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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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c. Contract No. 83550 and 92795
Installation, Operating, and Service Instructions for Kumkle Safety and
Relief Valves by Kumkle Valve Co:
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The inspector noted no revision or new information issued to date.
d. Contract No. 91934, N2M-2-1
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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
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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
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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
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Transmitters by GE/MAC:
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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 i
selected replacement would be authorized for all GE/MAC transmit- ,
ters. With the cancellation of the DCR, the applicant can react l
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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c. Contract No. 54752
Electrical Distribution Equipment Factory Order 11799 by
i Arrow Hart, Inc.: 4
The inspector noted that the type of distribution equipment in
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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:
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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.
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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-
t filled the commitment made in the licensee's response of November 7, ,
1983, to NRC GL 83-28. The procedure covered the responsibilities for
the receipt and control of vendor manuals and delineated the duties of
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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
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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.
The inspector interviewed pertinent licensee personnel to determine the
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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
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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 7
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
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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.
(4) MI 10.9, Section 6.3.9.1 refers to Steps 5.3.9.2 through 5.3.9.5.2 i
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-
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nessed by the audit team on February 4,1986, was the first time
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 : [ i } (c) Position numbers are as follows for Unit 2, Panel 2-L-116: f
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i RTA 2 BKRC-099-KG/320T ! ! RTB 2 BKRC-099-KH/320T l 8 '
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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,
1
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 ;
i work orders in September 1985, to enscribe unique serial j
l numbers on each circuit breaker frame to alleviate this i
problem. The new Revision 11 of MI-10.9 will require that 7
these unique numbers be included on every page containing !
q work and test records. 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
1 the dropout voltage differs more than 5 volts from the reference
- voltage. The reference voltage is not stated in the text, and !
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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. ,
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! j (7) Section 6.3.20 calls for five UV trips prior to returning the ! ! reactor trip circuit breaker to service after maintenance. l 4 Westinghouse recommends ten UV trips in NSD-TB-83-02, Revision 1, 1
page 5. i
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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. ! 4 . i (10) Some figures in MI-10.9 are poorly reproduced (e.g., Appendix B, i i Figure IA) and are difficult to read. ! i
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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
2
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 errors.
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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:
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(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.
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(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 1
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fair representation of the performance of the reactor trip !
breaker. I
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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
1 depress Channel I." The oscillograph traces were acceptable. -
;
(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 :
1 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
1 drawings. Contrary to the above, the following discrepancies were ;
,
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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.
t (3) Test personnel were not familiar with the procedure and methods
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. ,
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g. Motor Operated Valves (MOV)
i The cognizant electrical engineer described the M0V maintenance system
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
1
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
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Operated Valve and Test System (M0 VATS) by a M0 VATS company
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" l
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 I
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'dsa 're collected in 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.
1
(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-
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ties, QA review, job safety analysis, work authorized by craftsmen
section, work complete, post-maintenance test complete, status
trading, maintenance history records, etc.).
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1 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
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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 l
l
* Cable Terminations, splicing, and repairing of damaged cables, j
Not examined by the inspector. '
** 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
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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,
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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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