ML20244C391
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| ML20244C391 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 04/04/1989 |
| From: | Brady J, Elrod S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20244C347 | List: |
| References | |
| 50-327-88-50, 50-328-88-50, NUDOCS 8904200179 | |
| Download: ML20244C391 (61) | |
See also: IR 05000327/1988050
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, UNITED STATES
NUCLEAR REGULATORY COMMISSION
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hl. ;.f ;y -n REGION 11 - p ,j 101 MARIETTA STREET.N.W. E 5" L '2 - ATLANTA, GEORGIA 30323
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-Report Nos. 50-327/88-50,' 50-328/88-50
' Tennessee' Valley Authority
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' Licensee:
.6N 38A Lookout Place
1101 Market Street
Chattanooga, TN 37402-2801
: Docket Nos. 50-327 and 50-328 License Nos.: DPR-77 and DPR-79
Facility Name: Seq'oyah
u Units 1 and 2
Inspection Conducted: December 12-16, 1988 and January 9-26, 1989
Lead Inspector: } 8LA - />1 h Mn 4/f/89
Date Signed
f.;A.JElro Tdm L(ader
Team Members:
W. C. Bearden; Resident Inspector.
B. R.-Bonser, Project Engineer
J. N. Donohew, Senior Project Manager
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'G. E. Gears, Senior. Project Manager
R. D. Gibbs, Reactor Inspector
G. T. Hubbard, Branch Chief
D. P. Loveless,' Resident Inspector
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'G. A. Walton, Senior Resident Inspector
' Approved by: M[ag[ 9M[87
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@f B. Brady, ddting Chief, Date Signed
TVA Projects Section 1
TVA Projects Division
Summary
Scope: This routine, announced inspection involved inspection onsite in the
areas of. pla'nt operations; surveillance control, system outage
control, and work control prccesses; the corrective action program;
maintenance activities; plant modifications process; implementation
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- and verification of commitments made to the NRC; QA, QC, and QM
activities; the qualified reviewer process; and followup of events.
The inspection reviewed quality and quality verification in relation
H to' the ability of line management to get workers to accept
responsibility for doinc quality work, the ability of the quality
assurance and quality monitoring organizations to verify quality by
audits and surveillance, and the ability of the organization as a
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8904200179 89041o ,
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~g whole' to' _verifyf and , accept" quality at the interfaces between groups
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.when deliverables.are transferred.
Results: Onei violation was identified for. failure to take adequate corrective
caction to preclude repetition of previous violation 327, 328/87-30-01
. involving lack of control over ' plant . evolutions', and system and
equipment status 'in the radioactive waste area, . paragraph- 5 (VIO
!327,~328/88-50-01).
.Four unresolved items * were identified.
1. Trending within ACPs and the appropriate . thresholds for entering
.the CAQR; process, paragraph 7 (URI 327, 328/88-50-02).
2L Failure to include vendor torque requirements. in maintenance
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' instructions,. paragraph 8 (URI 327, 328/88-50-04).
3. . Completion of workplan review and the reporting of repairs and
replacements made under the- requirements of ASME Code
Section XI, paragraph 10 (URI 327, 328/88-50-05).
4. Engineering evaluations of vendor manuals, paragraph 5 (URI 327,
328/88-50-07).
It was concluded that site line management was strongly' dedicated to
quality and was convincing workers that quality work was what was
expected. One exception was found in the radwaste processing area.
This was revealed by the resin transfer problem discussed herein.
The events indicated that management attention had been lacking and
that overall site procedure upgrades had not had an affect on
upgrading quality in . this area. SQN has not yet completed the
indoctrination of personnel regarding accepting the responsibility of
doing quality work; however, thi s inspection found substantial
improvements, compared to past years, in the quality awareness of
personnel.
The function of the quality monitoring organization was to assist
site management in meeting quality objectives by iden ti fyi ng
conditions adverse to quality on a real-time basis before they
impacted on nuclear safety, reliability, or component operability.
This inspection concluded that the quality monitoring organization
was a well qualified, adequately staffed organization which was
performing its function well.
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' * Unresolved Items are matters about which more information is required
to determine whether they are acceptable or may involve violations or
deviations.
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The use of interfaces between groups and by the organization as a
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whole, to verify -and accept quality when deliverables were trans-
ferred ' was not emphasized- as 'a quality verification tool. For
example,..the Maintenance Department was using an interface
organization between the shops 'and QA to ensure thatLcompleted
surveillance tests represented quality work prior to their transfer
to - QA for review, however, some of the problems that were being
identified for correction had resulted because procedure changes had.
not bee'n adequately, communicated to the shop organization responsible
for performing them. An interface problem was also identified
between engineering and the plant in relation to vendora manuals
having conflicting data ' and resulted from a lack of communication
between the two organizations. These examples are discussed
L further in the. report details. Although interface problems between
engineering and the plant were identified by the NRC staff in an
earlier . inspection report (327, 328/87-52),. interfaces were still
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not . actively 'used by site or corporate management for quality
verification purposes.
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In order to reduce the enormous amount of intense upper management
effort necessary to make the CAQR system work, the licensee developed
a change to the CAQR process and implemented it in September 1988,
c immediately prior to the restart of Unit 1. The change provided
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several ACPs- to act as corrective action screening processes.. Those -
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issues not meeting the acceptance criteria for being a CAQR stayed in
the ACPs for resolution. This inspection concluded that the changes
were adequately implemented and strongly supported by senior line
management. The changes appeared to have the desired effect of
forcing insignificant and less significant issues down to the proper
level for resolution, while keeping safety significant items at the
senior management level.
The inspectors found SQN's process of plant surveillance control,
system outage control and work control to be well established,
controlled by procedures, and working. The process provided for the
identification of work needing to be performed, establishment of
procedures to do the work, scheduling of'the work, performance of the
work, and tracking of the work to completion,
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REPORT DETAILS
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1. Persons Contacted
Licensee Employees 4
*J. LaPoint, Site Director j
S. Smith, Plant Manager i
J. Anthony, Operations Group Supervisor, POTC
K. Allen, Periodic Test Coordinator
*T. Arney, Quality Assurance Manager
*W. Aslinger, Asst. Site Rep. Employee Concern Program
*R. Beecken, Maintenance Superintendent
H. Birch, Unit 1 Work Control Supervisor
*J. Blackburn, Special Projects Engineer
G. Boles, Manager, Maintenance Planning and Technical
*E. Boyles, Site Rep. Employee Concern Program
L. Bryant, Program Support Manager
S. Chapman, Supervisor, Document Closure
*M. Cooper, Compliance Licensing Manager
D. Craven, Plant Support Superintendent !
I. DiBase, Environmental Qualification and Preventive
Maintenance Program Supervisor
H. Elkins, Instrument Maintenance Group Manager
R. Fortenberry, Technical Support Supervisor
H. Gammage, Site Procedures Manager
J. Hamilton, Quality Engineering Manager
*R. Hays, Radioactive Waste Processing Supervisor
*J. Holland, Corrective Action Program Manager
T. Howard, Quality Assurance Surveillance Supervisor
S. Johnson, Quality Assurance Technical Support Supervisor
D. Jones, Shift Outage Manager
N. Kazanas, Vice President, Nuclear Quality Assurance
*D. Kelley, Waste Water Processing Group Manager
*J. Klein, Maintenance Trending and NPRDS Supervisor
R. Lewis, Modifications Engineer
C. Lonas, System Evaluator
*J. Maddox, Engineering Assurance Lead Engineer
*L. Martin, Site Quality Manager
T. Matthews, Licensing Project Manager
*D. Michlink, Coordinating Project Engineer
R. Miles, Modifications Manager
L. Parscale, Manager of Licensing Performance
J. Patrick, Operations Group Manager
*J. Petty, Shift Technical Advisor Supervisor
R. Pierce, Mechanical Maintenance Supervisor
G. Putt, Work Control / Outage Group Superintendent
M. Ray, Site Licensing Staff Manager
*J. Robinson, Modifications Manager, Engineering
F. Roemer, Nuclear Quality Assurance Engineer
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*R. Rogers, Plant Support Superintendent
R. Shell, Compliance and Services Manager 1
*S. Spencer, Licensing Engineer
T. Spink,l Replacement Items Program Manager
M..Sullivan, Radiological Controls Superintendent {
D. Thomas,' Impact Evaluator l
*P. Wallace, Site Programs Manager
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*J. Walker, Manager of Operations . Support.and Procedures {
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;*J. Ward, Outage Scheduling Supervisor
J.. Wilder, Senior Engineering Specialist.
'*L. Wheeler, Materials Manager
C. Whittemore, Licensing Engineer
*A."Wilkey,'Sequoyah Quality Audit Group Manager
NRC Employees
*F. McCoy, Assistant Director for TVA Inspection-Programs
*L. Watson, Section Chief
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K. Jenison, Senior Resident InspectorL
P. Harmon, Senior _ Resident Inspector
L 'P. Humphrey, Resident Inspector
*J. Brady, Project Engineer
* Attended exit interview
' NOTE: Acronyms and.initialisms used in this report are listed in the last
paragraph.
2. Introduction
- This special, announced NRC team inspection at the Sequoyah Nuclear
Plant was performed to evaluate the acceptability of the line and QV.
organizations' activities and . management's support of these activities.
This inspection was performed primarily under the guidance of NRC
' Inspection Manual Chapter 35702, " Inspection of Quality Verification
Function". The inspection consisted of personnel interviews, direct
observation'of in progress activities, and review of work documents.
Line organizations must be aware of quality requirements and must be
confident of their ability to perform to the required level of quality.
If the QV organizations are technically credible, they can and should
help define identified deficiencies, provide insight into the root cause
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of deficiencies, and approve and confirm the resolution of deficiencies
in 'a ; technically meaningful way. The inspection also assessed line
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management's ability to ensure that identified deficiencies are dealt
with promptly and completely.
Quality Verification Function Inspections are not intended to verify
licensee compliance with administrative controls; they are intended to
verify'the technical adequacy of safety-related activities. However, if
deficiencies are found in these activities, the underlying procedures and
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, administrative controls 'are reviewed.' The. intent of these inspections is-
> to . improve plant operational safety through inspection-processes that' are
r : focused on activities'that. affect plant safety.and reliability.
_ The QVFI at Sequoyah primarily focused ~ on plant operations; corrective'and.
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preventive. maintenance 'of . plant systems and components; and the . control.
, system 'for' surveillance, system outages, and other work. A ' secondary
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' focus.was -the completion of and verification of commitments :made to the - ;
LNRC. -Several program areas, such as ' the corrective action program, .were '
Jalso': inspected. . The inspectors reviewed selected documented examples- in .
these and closely associated areas to _ identify safety-significant problems
(if any) ~ to .be used as ' vehicles for evaluating the - effectiveness - of-
qualityJachievement, _ self-verification, and other verification. The
results'of,this review.are discussed in'this report.
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L3. :In spection .0bjec'tives':
The primary. objectives of this inspection were:
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Assess ~the effectiveness of the licensee's various line organizations-
'in achieving and self-verifying quality in their. functions.
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Assess ' the . " quality verification effectiveness" of the 'other .QV
organizations: in identifying,- resolving, and preventing
safety-significant ' problems and deficiencies 'in various functional
areas.
I? For this assessment, " quality verification effectiveness" is defined
as the . ability of the ~ 1icensee to verify quality and to identify,
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correct, and prevent problems It is not limited tc the licensee's
-Quality Assurance Organization, but is the aggregate of all efforts
to verify quality results and take corrective action when a quality
result was not obtained. It specifically includes the line
organizations.
- Assess the effectiveness of the _line management in ensuring that
safety-significant problems and deficiencies are dealt with promptly
and. completely, in response to input from the QV organizations.
- Assess the effectiveness of communication of generic issues within
the QV organizations and line management.
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Assess the effectiveness of communication of lessons learned from
other TVA sites.
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Assess ef fecti /eness of engineering support to correct
deficiencies. {
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A secondary objective, closely related to site performance and verifica- )
tion, was to assess the performance and verification effectiveness related 1
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toEthe commitments made to. the NRC which were prerequisites for ' allowing
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the return to operation.
4. Inspection Scope:
The inspection was divided into the following functional areas:
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Plant Operations page 5
-- Surveillance Control / System Outage Control / Work Control Processes -
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Corrective-Action Program page 21
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Maintenance' Activities page 26-
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Safety Information Management. System / Corporate Commitment Tracking
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Plant Modification Process page 38
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-QA Routine? Audits / QC Activities / and Special Surveillance (or
monitoring: activities) in Support of Operations page 47
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Independent-Qualified Reviewer Process page 50
l5. Plant Operations
At the beginning of the operations inspection, both reactor units were
operating' at power. To better grasp the workings of the Operations
Department. and the QA/QM programs related to operations, the inspectors
met briefly with Operations and QA management.
To detect previous problems and identify potential repetitive problems or-
problem trends,1988 documentation from various problem tracking systems
was reviewed. The documentation review included TROI printouts - both
open and closed items, PRO printouts, selected QM reports, and the NRC
Open Items List. The documentation review revealed there had been a
signi f.i cant number of configuration control and operational status
problems prior to and duriag the restart of Unit 2 in late 1987 and the
first half of 1988. The problems had the following characteristics:
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These problems centered around maintaining cognizance of the opera-
tional. status of critical systems, structures, and components. This
included status records of valve power supplies, instrumentation,
penetrations, and structural components.
- The problems had occurred during normal plant operation, during the
performance of surveillance instructions, and during maintenance.
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. .Some:of the'more significant problems found, which the licensee indicated
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Then-current: procedures. and system checklists did not_ always match
existing as-configured prints and existing system, configurations.
J. Procedures which' called for closing valves with the plant in Mode 4
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would' prevent the centrifugal charging _ pumps.from having their normal
cold leg : injection path to ' the .RCS. (See CAQR SQP 880151 discussed
'below).
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Some of' the ERCW- valves. maintained by- SI-682 were be'ing moved from
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their required positions without 'any documented reason. (See CAQR:
SQP 880213 discussed below)'
c- : Conflicting' valve positions between' an SI for system 67 and a GOI,
. coupled with J the valves being found out of positioni during SI-
performance.'(See CAQR SQP 880490 discussed below)
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' Valves-_ _ being omitted- from a valve- . check 11st could cause an:
inappropriate valve alignment.
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. Not- correctly maintaining configuration control logs in the control
room.
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. Failures -of the system ' alignment corrective action program to
eliminate SOI checklist inadequacies prior to restarting system
alignments.
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. A check of NRC inspection reports showed that some of ~ the above-listed
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problems were identified as a result of NRC findings:
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.NRC inspectors had reviewed the licensee's control of system
configuration status by auditing logs and recently-completed SOI-
checklists and walking down' the. RHR and UHI systems. Four examples
of failure to properly implement procedures' associated with
controlling plant configuration were identified. (IR 327,328/88-26)
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Spills of primary coolant water occurred as a result of misconfigured
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systems. (IR 327, 328/87-24 and 87-30)
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Examples of components being out of position. (IR 327, 328/87-66 and
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88-06) (See CAQR: SQP 880245 discussed below)
L Specific CAQRs were examined more closely for details and corrective l actions: p. L
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SQP 880151 - - Procedures GOI-1 and G01-3 had BIT isolation valves
LL 63-25, '-26, -39, & -40 isolated while the plant was in Mode 4 - ! preventing a (requi~ed for Mode 4) flow path from the centrifugal
charging pumps to the normal, cold-leg, injection path to the RCS.
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The corrective actfons were checked and appeared to be sufficient and
complete.
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SQP 830213 - Many performances of SI-682 over the past several years
l~ had shown unexplained deficiencies. ERCW FCVs were being moved from l
required positions. The corrective actions given in the CAQR were
checked and appeared to be complete and satisfactory.
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SQP 880245 - This CAQR was written in response to NRC violation 327,
328/88-06-02. The CAQR involved a failure of the system alignment
corrective action program to eliminate SOI ' checklist inadequacies
prior to restarting the system alignment process. ' Corrective actions
for some.of the specific examples cited were checked. The corrective
actions appeared to..be sufficient and complete.
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SQP 880414 - This CAQR. involved.the use of status boards to maintain
configuration control . The corrective action included revision of
AI-58 to delete the use of stat 0s boards for configuration control
and require the use of a configura} ion log.
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SQP 880490: - This CAQR - involved conflicts between procedures
pertaining to the positions for certain ERCW valves. The corrective
action was checked and appeared to be sufficient and complete.
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SQP 880504 - This CAQR was written because the waste disposal system
(liquid) checklist . did . not contain certain system valves. This
review;found thev1isted valvesnto inow be included in the system
checklist;
The review of these - CAQRs indicated to the inspector that previous
configuration control problems had received significant appropriate
attention and that the corrective actions given in the CAQRs were
sufficient and had been completed. A review of recent licensee quality
reports- in the configuration / status' control area ' identified no
significant problems.
Having completed the past problem review for background 5 information,
current activities were reviewed.
As part of . the. review of this area, the NRC inspectors reviewed the
conduct of operations activities. NRC inspectors spent some time in most
plant locations including the control room. This effort included normal
working hoursp backshift, and shift turnover. This time was primarily
spent directly observing operational activities during power operation on
both units and included routine shift activities and the performance of
surveillance testing. Since configuration control had been a previous
problem, this review was sensitive to that subject.
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Control room activities were generally conducted in an efficient and
professional mannere Formal and clear communications were observed in
most cases. Personnel traffic and. noise levels in the control room were
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controlled such that 'they did not interfere with routine shift' activities.
. Operators were attentive, aware of, plant conditions, and remained in their.
-designated areas. ' Shift manning was in accordance with TS requirements
and proper relief.was observed prior to shift ~ personnel being discharged
from their duties. Major scheduled activities appeared to' receive the
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proper amount of. management review and planning prior to being included in
[ the daily scheduled list of work activities. l' ' The 111censee. had; established ' systems to maintain cognizance of. system .
status and maintain configuration control. The effectiveness - of the
present systems and their conformance to the appropriate Als was reviewed.
A System Status File was ' set up in the control room for Units 1 and 2.
Critical systems were aligned as required by appropriate valve and power
availability' checklists following the last outage. The existence and
. maintenance of these files was verified. As evidenced by their corrective-
action- in the .CAQRs, the licensee had expended significant effort in
assuring that the checklists included all valves and electrical
components. However, some problems still existed -as evidenced by .the
finding, while transferring spent resin, of a waste disposal- system
interface valve that was not listed in a checklist' er shown on approved
drawings.
.When conditions required deviating from the normal system ' alignment, the
deviations were to be entered in a Configuration Log. The Configuration
Log and the System Status File should reflect the current status of
critical systems. The Units 1 and 2 control room Configuration Logs were
checked. No deficiencies were noted. As described in AI-58, Maintaining
Cognizance of Operational Status - Configuration Status Control, the log
was indexed by system numbers corresponding to the numbers given in the
- AI, which was. maintained in the unit horseshoe area. The log entries
appeared to conform to the AI.
The Test Awareness Log was reviewed. The Test Awareness Log is used
for ongoing activities controlled by approved procedures that provide both
configuration changes and return-to-normal within the procedure. These
activities were not entered in the Configuration Control Log. A review of
the entries in the Unit 1 Test Awareness ' Log showed one deficiency.
Maintenance Instruction MI-20.12, Calibration of Bourdon Tube Pressure
Indicators, had been started on January 15, 1989, at 7:55 am, had not been
closed out. The Unit 1 SOS indicated it was an error and the log entry
should have been closed out.
The. Critical Valve Summary Checklists in the Unit 1 and 2 control room
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areas were reviewed. The list is utilized to provide verification of
proper valve alignment each shif t on safety related systems to ensure
system operability in the applicable modes. The inspector's review of the
checklist revealed no deficiencies.
The Unit 1 UD log was reviewed for configuration control practices. One
minor deficiency was noted on the 2300 - 0700 log for January 16-17, 1989.
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, An LCOLwhich was entered and logged.for changeout of'a radiation monitor
= filter.was not logged when;the changeout was complete, however,' completion
of the. activity.was-logged. Elapsed time from beginning to end.was about'
eight' minutes. .
.The Unit 1 LCO Action' Log was reviewed. The LCO. Ac' tion Log was not
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treated as' a .QA record but was treated as an operator. aid. The UO log
discussed'above provides the QA record. No deficiencies were noted during
this review.
The inspector's. overal1Li.mpression of the configuration / status-control
systems established by Lthe licensee was. good. However, even in the small-
sample size reviewed by the . inspector, minor deficiencies were noted.
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.The111censee. was : developing a pilot program .for use during the Unit. 2,
l Cycle 3 refueling outage. The program will utilize a TS Condition Report'-
to track out-of-service TS: components or systems when the plant is:in an
operational mode where that TS would not be presently applicable or where
g - the. TS .would ' become more conservative later in a different- mode of
operation.. The. program effectiveness in ensuring TS compliance will be
evaluated during future inspection activities.
Tours of .the Unit 1 and'2 Auxiliary, Control, and Turbine Buildings were
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conducted to observe plant conditions', including general- cleanliness /
housekeeping, potential fire hazards, leaks, and adverse equipment
operation. The NRC inspector noted that cleanliness and housekeeping were
excellent. No significant material or equipment problems were noted,
however, one concern was noted in that all six'AFW pumps appeared to have
excessive packing leakage while the pumps were idle. This concern was
discussed with licensee management who stated that no acceptance criteria
for leakage exists and :the packing glands were intentionally maintained
loose to increase packing life and decrease shaft sleeve wear DCR 2565
has been written to replace these pumps' packings with mechanical seal
assemblies. Since the first pump is scheduled to be modified during the
upcoming Unit 2, Cycle 3 outage, the NRC inspector believes that the
licensee has adequately addressed this concern- l
In' addition to the routine system configuration control set-up, control
must be maintained during the performance of sis which involve
manipulation of the plant. To prevent recurrence of mispositioneo valves,
the licensee formed a dedic.ated operations SI team. The formation of this
team limited the number of persons performing operations department SI
tests, increased each person's exposure to the sis and enhanced the
internal communications within the group. The inspector specifically
observed the functioning of the SI team during surveillance SI-5 and
interviewed team members. The SI team concept appeared to be effective in
improving efficiency and control.
The NRC inspectors observed portions of the following surveillance tests:
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SI-2, Operations Shift' Log.
SI-2, Operations Shift Log
SI-3,. Operations. Daily Log
SI-3, Operations Daily Log.
.' SI-5, : AFW Valve Position Verification
" SI-37.1, Containment Spray Pump 1B-B
Quarterly Operating-Test
SI-166;1, UHI Isolation Valve Full St'oking- r
SI-744, Monitoring of UHI Isolation Valve Accumulator Pressure
During the conduct of the surveillance. tests, ' operations ~ personnel
responsible' for conducting the tests, whether from the SI . team or: not,
appeared- to have an excellent knowledge of the sis and the associated
plant systems. The instructions reviewed contained an . adequate level: of
information and technical detail' to allow proper performance. All'
personnel involved in the ' performance of the testing attended a'
. pre-evolution briefing and all SI prerequisites were satisfied prior to
starting each test. ' The testing was properly coordinated and applicable
steps were followed verbatim during actual performance of the testing.
The NRC inspector reviewed numerous completed QM reports covering-
operations activities that had been performed since the creation of the
site QM organization. QM reports . reviewed' included: configuration
control, temporary alterations, TS surveillance, . operator logs, shift
- relief and turnover, shift compliment, and plant staff overtime.
Additionally, the.NRC inspectors reviewed completed quality audit reports
SQA 88-808 and SQA 88-815 which covered a variety of operations-related
activities.
The review of QM reports and quality audit reports identified no
significant problem areas. However, past control of temporary alterations
and the accuracy and detail contained in past operator logs were
identified repeatedly by QM and audit reports as weaknesses that warranted
greater management attention. The NRC inspector noted that operator logs
have since improved and that the control of temporary alterations has
recently received improved management attention.
An operator journal, or log, should contain a narrative of the plant's
status and all events required to provide an accurate history of plant
operations as stated in AI-6, Log Entries and Review. Although
improvement has been noted in the detail included in the various logs
i~ maintained by the operating shifts, a reviewer has not always been able to
determine the full history associated with a given event from the 00 log
or SOS log. Often it has been necessary to refer to several other logs
and charts to reconstruct the actual conditions that led up to a
particular event.
The NRC inspector believes that an increased emphasis by management and
the QM Group would result in continued improvement. Although log entries
'
were not the subject of inspection findings identified in previous NRC
inspection reports 327, 328/88-17, 88-35, and 88-39, the lack of good
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detailed. log entries contributed to the failure to perform an adequate
post trip review in previous Violation 88-35-01 ~and problems associated j
with operators unknowingly ' entering. LCOs in previous URI 88-17-02 and <
Violation 88-39-02.
i
The licensee has had a long term problem with the control of temporary
alterations. Abuse of the TACF system in the past had resulted in ]
many modifications being controlled under AI-9, Control of Temporary .
Alterations, long after the short time period intended for conversion had
elapsed. This resulted in a large backlog of TACFs and had the potential
to -strain administrative control effectiveness. This problem had been
identified' in NRC Inspection Reports 85-46, 86-20, 86-27,.87-08, 88-47,
and 88-51. The. problem was also discussed in two recent QM reports and
the Febuary, 1988, ISEG Monthly Report, 88-02-SQN-I.
As .of' December 1986, approximately 200 TACFs still remained active for
both units. - Although a management action tracking system required routine
review for the purpose of justifying continued need, as of May 1988, 138
"
active TACFs'still existed, and many TACFs issued prior to 1984 continued
to remain in place. The licensee had committed to INP0 to clear all
temporary alterations that were in place on January 1, 1984, before Unit I
startup following Cycle 4 completion. In spite of this, the problem
appeared to receive inadequate ~ management attention until the later half
of 1988.
During July - August 1988, two monitoring reports were issued which
identified various administrative errors such as incorrect entries in the
TACF index and missing TACF forms. Additionally, one of the monitoring
reports identified that as of July 18, 1988, 90 TACF reviews were
delinquent. To ensure TACFs are properly reviewed and closed in a timely
manner, AI-9, Section 7.2.2, requires that a review by the responsible
section supervisor be documented on Attachment 6A. . Forty-four of the 90
deliquent reviews were on CSSC Systems. This deficiency was documented
in CAQR SQQ 880420. The NRC inspector reviewed the corrective action for
this CAQR and determined that it was adequate. The recent revision to
AI-9, if followed, should preclude recurrance of this problem.
Recent improvement in licensee management attention in this area resulted
in further reduction of the backlog. AI-9 recently enderwent a .inajor
revision. Changes included a decision to reassign responsibility as
implementing organization from Planning and Scheduling to the Technical
Support Group Superintendent - who will perform periodic reviews of all
installed temporary alterations. Additionally, all proposed TACFs and
extensions for existing TACFs must be approved by the Plant Manager. PORC
o. review will be required prior to installation of any TACF on CSSC
equipment or for extension of a TACF installed on CSSC equipment. As of
L January '18, 1989, 94 active TACFs exist and the licensee has developed a
work-off schedule for the remaining TACFs to meet the INPO commitment.
This- schedule includes budgeting, for FY89 and FY90, for DNE and the
onsite modifications group to close all but six of the remaining TACFs by
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ym. . I f [ September 1990. . These six TACFs are associated' with' future UHI removal
and temporary thermocouple in Unit 2 steam valve' vaults.
Alth'ough:the licensee appears to be on .the path toward' correcting this'
l j deficiency, a large amount' of. time .was allowed to pass without adequate
- management' attention in this' area.
On January 11, 1989, the licensee attempted to' transfer spent resin from
"
- the .2ACVCS mixed-bed demineralized to the' SRST -using SOI-77.3, Rev 15',
l Section'B'- Waste Processing - Spent Resin Storage Tank. The. resin' failed
,
'
to transfer.. completely. Radiation dose rates in: the' vacinity of the
transfer lines. varied but ranged as high as 90 Rem pe'r hour near one
, valve. It was believed'.that a commonly-used diaphragm. valve in the
transfer line might have failed with a ruptured diaphragm, blocking the
line. Because. ~ of 'the unusual result ,of the transfer attempt, NRC
inspectors reviewed controlling' literature and licensee recovery
activities to determine the extent of quality attainment and quality.
: verification Literature reviewed included:
. -
SOI-77.3,LRev.15, Waste Processing,. Sections B and C
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MI-11.7.1, Rev.6, Hand Operated Grinnell or Saunders Type Diaphragm-
Valve Rebuilding.for All Systems
Aproved vendor manual SQN-VTM-I207-0010, Vendor Technical Manual for
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. In'sta11 ati on and Maintenance Instructions for Nuclear Diaphragm
Valves.
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Numerous PM requirement sheets for diaphram valves
SOI-77.3B was found to be unuseable as a category."A" verbatim compliance
procedure ICF 89-0035 was written under detailed management direction,
field verified, and approved prior to use' in the recovery. S0I-77.3C was
also found to be unuseable and not followed - resulting in a personnel
contamination, which is discussed in more detail below. CAQR SQN 890016
was issued for the inedequate procedure.
The vcendor manual had two conflicting sections addrest.ing the same
valves. Items in confl?ct included body-to-bonnet bolt tcrque valts and
whether or not to use torque values or just tighten the bolts l
finger-ttght. CAQR SQN 8900026 was issued to resolve this issue.
t . Although the PMs and MI-11.7.1 appeared to be sound, having been written i
with informal vendor consultation, t. hey did not implement all of the
requirements of the above VTM. DNE had not documented evaluations for
these deviations. CAQR SQN 890026 also addressed this condition.
At the end of the inspection, DNE representatives indicated that it may be
acceptable to issue conflicting information and let the maintenance staff
pick the version to use. This is an indication of an interface problem
between DNE and the plant staff in relation to vendor technical manuals.
Interface problems between DNE and the plant had been previously
identified in IR 327,328/87-52 pertaining to the use of compensatory !
measures. I
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l Based on the above-described findings, the status of vendor manuals as the
basis for plant technical instructions is unresolved pending licensee, and
subsequent NRC, review. This is URI 327, 328/88-50-07.
l On - January 16, 1989 an AVO was contaminated during the performance of l SOI-77.3C, Waste Processing - Transfer of Spent Resin from SRST to l Shipping Liner. Numerous procedural errors, failures to follow
procedures, and operations outsic'e of procedures were common in the
S01-77.3C performance. The review of the incident highlighted the
following problems:
-
Instruction step "B" required the performer to initiate a WR to
install a jumper on contacts SCG-3 and SCG-2 in junction box 3017 to
allow 0-FCV-77-225 to remain open during resin sluicing. This step
was- performed without using a WR. The design of 0-FCV-77-225 was to
automatically close upon a high level indication signal from
0-LE-77-225. This level element was not in current use at the plant.
A temporary level indicator was used in the shipping liner instead.
Previously this indicator was jumpered out via a TACF. This TACF was
terminated and step "B" added to control the subject jumpers. Step
"G" later re-verified that this was accomplished by requiring the
performer to review the procedure prior to loading the shipping liner
and to verify that the jumper had been installed by the WR initiated
in procedure step "B". Lack of a WR or design change to control
these jumpers is an example of inadequate configuration control.
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Step "E" required the performer to obtain a SRST level reading on
0-LI-77-48. The SRST water level indication was poorly designed such
that it was unreliable at higher tank pressures. Therefore, the
operators would open the tank vent, 0-FCV-77-51, to reduce the
pressure prior to taking true tank water level readings. This
knowledge was not discussed in the procedure nor did the procedure
allow tha opening of the vent valve for this purpose.
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Step "F" required the performer to open 0-FCV-77-45, fill the SRST to :
approximately 10-15's above the resin level, then close 0-FCV-77-45. ;
Step "F" was not performed because there was no indication of SRST
2
resin level. Additionally, as addressed above, the water level
indication for the SRST was unreliable. The operators explained that
the step was accomplished Dy first draining the tank watar through
0-FCV-77-698, then slowly filling the SRST while alternately opening
nitrogen valva 0-FCV-77-46 and venting via 0-FCV-77-51. When the
point was reached that the tant quickly pressurized, they knew that
the void space was steall and, therefore, that the tank was almost
full. This, they claimed, provided the basis for the assumption *
that the resin was covered with water. The licensee later determined
that the SRST was completely full and overflowing with the resin.
Appropriate performance of this step would have determined this
problem. This is an example of inadequate instrumentation to
accomplish the procedure requirements.
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Step "H" required the performer to open valves 0-77-953 and 0-77-845.
Valve 0-77-845 did not exist in the plant. Operators simply crossed
out the valve number in SOI-77.3C and-wrote in those valves they
wanted to open. This was accomplished without approval or knowledge
of the control room operators. This is an example of inadequate
i configuration control, inadequate procedures, and failure to follow
procedures.
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Step "J" required the performer to test the shipping liner level
transmitter after setting up to load the liner. This evolution
involved multiple valve manipulations not addressed in the procedure.
This is an example of an inadequate procedure.
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Steps "N" and "0" required the performer to perform the following
manipulations: !
N. Pressurize SRST to 80 psig, then open in order.
(1) 0-FCV-77-300 PW to flush
(2) 0-FCV-77-225 Spent resin tank to liner
(3) 0-FCV-77-49 Spent resin tank outlet
O. When back flow is noted:
Open 0-FVC-77-226
Close 0-FCV-77-300 (Liner is now filling)
NOTE: (1) Maintain Nitrogen pressure at 80 psig during
transfer
Procedure steps "N" and "0" wvre adequate to perform the evolution by
pressurizing the tank, back flushing the lines, and then reversing
flow in the lines to sluice the resin to the liner. However, the
operators in the radioactive waste crganizatior, did not understand
the procecure's approach. '1 hey, therefore, took steps to pressurize
the tank after aligning for the sluice. This it an example of
fnilure to follow procedures.
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Step "P" required the performer to watch the liner level transmitter.
When all four lights are on, the liner is full Then perform steps
Q, R, and S in rapid succession. !
During this evolutien, the operator only transferred a small quantity
of resin to the liner and did not attempt to fill tne liner. The
operator knew from his training that the resin had to be sampled
prior to filling the liner. The movement of a small quantity of
resin, the sampling process and the analysis of that sample were not
proceduralized. This is an example of an inadequate procedure.
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Step "S"'. required the performer to close 0-FCV-77-226, Liner Fill
Valve, when water flow to the liner was observed. As'an alternative, j
the operators 'and HP technician determined when the dose rates in the
line dropped, indicating that the : resin had' passed, because. the
temporary'. liner level indication was not sensitive enough to. indicate
the onset of' flow by a changing level. .
This is an example of an
inadequate' procedure.
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Step'"T" required the performer. to close 0-FCV-77-225, Spent Resin '1
Tank -Isolation Valve, when flow to the SRST was observed. The
operator stated that 0-FCV-77-225 would actually be closed shortly
after 0-FCV-77-226 was closed because flow indication to the SRST was .
unreliable. .This is an example of an inadequate design. Following
.the perfo'rmance.of step "T", S01-77.3C assumed the lines to be clear
of resin.-
-Actually, following the flush, the HP technician detected a hot
spot of approximately 15 Rem per hour on the backside of valve
0-FCV-77-400. The operator told the inspector that this usually
occurred during resin transfers from the SRST to the shipping liner.
The operator .then disconnected the dewatering pump discharge line
from the ~ fitting at 0-FCV-77-401 (to the tritiated drain tank)
and~ connected it to the fitting upstream of 0-FCV-77-400 (resin
dispensing header-to auxiliary contract [ equipment]). He then opened
valves 0-FCV-77-400, 0-FCV-77-225, and 0-FCV-77-226; started the
dewatering pump; and began to recirculate the water in the liner.
This' was accomplished by taking a suction on the dewatering vanes
inside the liner, discharging into the auxiliary contract header
through valves 400, 225 and 226 and then back into the liner. This
evolution flushed the resins from the hot spot and into the liner.
This entire evolution was performed outside of an approved procedure
and is an example of an inadequate procedure and a design deficiency.
Following this flushing evolution, the operator, again without a
procedure, isolated the valves acd disconnected the discharge
hose from the fitting. This breaching of the system was not first
discussed with the continuous-coverage HP technicians. Upon
disconnecting the hose, residual pressure in the lire caused water
to spray on the walls and floor, and c' contaminated the: operator's
face. The operator was subsequently decontaminated.
'
This is an example of equipment and personnel hazards generated from
performing evolutions without approved procedures. This contemina-
tion would likely have besn prevented had appropriate procedures been
in place.
The inspector noted that during the performance of 501-77.3C. the
operator had not signed several steps as being complete prior to
being contaminated. The operator's replacement, noting that he had
accomplished these steps, signed them off. This is a violation of
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AI-4, Preparation, Review, Approval and Use of Site Procedures /
Instructions. This is an example of failure to follow procedure.
The licensee management replaced the supervisor of the radioactive waste
organization and temporarily revised 501-77.3 to enable them to move the
spent resin and flush the hot spots. The inspector observed licensee
activities during this recovery. Several times, the operators from the
waste organization suggested that evolutions be performed that would
violate the procedure. These evolutions were not performed because the
new group management determined them to be unacceptable. This suggested
to the inspector that, prior to the management changes, the entire group
lacked adequate training, supervision, and self discipline to safely
perform the operations.
Violation 327, 328/87-30-01 was written in July 1987, to address a lack
of control over plant evolutions and the status of systems and equipme'nt.
The NRC issued the Notice of Violation to obtain the management attention
necessary to resolve the underlying problem.
This previous citation noted specifically that:
a. The unit operator did not use the formal change process of AI-4 to
revise SI-166.3. Instead, a system realignment was improvised
without written or formally-approved instructions. This example is
similar to many of the current examples noted above, including
steps E, F, H, J, N, 0, P, S, and T.
b. Five valves were shut in an attempt to isolate the SG maintenance
area from the RWST without entry in the configuration log as
required, and vent valve 1-HCV-68-594 was opened without this
deviation from the normal valve alignment being entered in the
configuration log. In the current examples, valves were re-aligned
during these previous evolutions without procedure or configuration
control, as noted in steps E, F, H, J, N, 0, S, and T.
,c. On December 2, 1966, December 14, 1986, and May 22, 1987, SI 4S.1
was performed without appropriate instrumentation to verify the
j flow rate. In tne current examples, level indications in the solid
radioactive waste (spent rer.in) system were inadequate and flow 1
indications did not exi st to perform the steps of S01-77.3C as
written, at noted in steps E, F, S, and T.
The licensee's response to the previcus Notice of Violation committed,
in part, tc the following actions:
'
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a. Perform procedural adherence training that emphasized the require-
ments and reasons for procedural compliance to over 900 employees.
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b .' Discuss'the.importance of configuration control and failure to comply
"
with procedures regarding these [ previous] events during the
Operations Lessons' Learned Sessions.
l' NRC Inspection Report 327,328/87-50 re-opened NRC review of ' violations. !' resulting from a 1984. thimble tube ejection event,- including violation
84-24-01. Violation 84-24-01 addressed failureLto establish and implement-
procedures for ~ the conduct of equipment control, procedure review and
.approval , per formanc e of maintenance, radiation -work permit access con-
trol, and access to containment. The NRC was concerned that perhaps the
corrective actions taken for violation -84-24-01 had been inadequate to
i preclude subsequent violation 87-30-01 discussed above.
In a letter dated November 8, 1987, the licensee submitted a revised
response to Violation 84-24-01. In this response, the' licensee stated
othat procedurals adherence was now required by AI-4 and that additional
-
training of plant personnel had been- performed on procedural adherence.
Additionally the licensee stated that this training corrected problems
with individuals not obtaining procedure changes prior to performing work
not specifically addressed ~in the procedure.
10CFR 50, Appendix B, Criterion XVI, Corrective Action, requires that, for
significant conditions. adverse to quality, corrective measures determine
the cause and preclude repetition. Prior to January 16, 1989, the
' licensee failed to preclude repetition of Violation 327, 328/87-30-01
. af ter having ' completed corrective action. Lack of control over plant-
evolutions, system status and equipment status was still evident in the
,
radioactive waste area. As a result:
a. Multiple evolutions were performed outside of approved plant
procedures during spent resin transfers from January 11-19, 1989.
b. Known inadequate drawings were utilized affecting procedural controls
of the temporary resin interface valve.
'
c. Known design deficiencies were not corrected causing recurring
radiation hot spots in excess of one Rem per hour to be created.
The conditions existing, cembined with evolutions beir.g performed outside
of approved procedures, allowed an unwarranted personnel contamination.
This is identified as Violation 327, 328/88-50-01.
.' E h', 6. Surveillance Control / System Outage Control / Work Control Processes I
To understand how work at SQN was controlled and quality verification in
the line organizations was accomplished, the inspectors interviewed SQN 1
personnel who were involved in the preparation, scheduling, and control of J
I
plant work. In addition to the interviews, the inspectors observed
samples of their in progress work.
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At SQN, surveillance. control, system outage control, and work control were
three: interrelated ~ functions performed by- the Work Control / Outage
;
V . organization, the Outage Scheduling organization, and the Maintenance
Planning and Technical organization. These. organizations coordinated -
their; respective activities to ensure that required work within the plant,.
i.e. , sis, WRs, and _ PMs were identified, scheduled, and performed as ,
req'ui red. '
a. Work Control /0utage' Control Organization'n
The: Work ' Control / Outage Control organization . was responsible for
scheduling when the work was- to be performed in . the plant and
assuring that the work was accomplished. These functions were
accomplished. utilizing System Evaluators (SE), Work Schedulers, Shift
Outage Managers, and Impact Evaluators. ,
,
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.SEs were' assigned to specific systems in a specific Unit (Unit 1 or
2) and were . responsible for maintaining cognizance of . work being
performed on tneir assigned systems. Duties of the SEs included:
review of WRs,- processing of WRs to the required organizations. .for
'
, planning, determination of when work required removal of equipment
from plant -service, . and when equipment . tag outs were required,
coordination of work on assigned systems with work schedulers and the
craft organizations performing the work, and ' ensuring the work was
completed.
-Work schedulers were responsible for the actual scheduling of the
work start and duration. The schedulers utilized a computer listing
of work available to track the work required to be completed. Using
this information, the schedulers, through coordination with craft
organizations and the SEs, established schedules for work
performance.
For periodic work having pre-established procedures for performance,
e.g., sis, scheduling was controlled by the " Periodic Test
Coordinator" organization. This organization was responsible for
ensuring that periodic tests were. includec in the PDWL, which was
differem f or each shift of each day and was irsued three times each
day. The PDWL also included other significant information relative
to plant activities for each shift including identification of major
work activities of the previous shift and major work activities for
the onccming shift. In addition to ensuring that periodie work 1
activities were scheduled, the periodic test coordinator organization
was responsible for ensuring that the periodic work was indeed ]
completed as required. ;
;
Shift outage managers kept track of the work scheduled to be
performed each shift and the work actually accomplished.. The shift ,
outage managers utilized shift turnover meetings to update oncoming
shifts of the work previously accomplished and the work scheduled to )
be accomplished during the oncoming shift. During these turnover
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meetings, the shift managers provided copies of the PDWL. . As
discussed above,'these lists included. scheduled sis as well.as major
work-' activities -completed during the: previous shift and those
scheduled for the oncoming shift.
Impact evaluators were responsible for reviewing-work which was' to be
performed in the plant to evaluate the work's effect on the. plant
status. While- the SEs were concerned with the effects of work on
their assigned systems, the itpact evaluators, who were SR0s, were
concerned with whether work on one . system would adversely affect
another system, or systems, .resulting in unplanned changes in the
plant status, e'.g., a reactor trip.
.
The Work: Control / Outage. organization and other groups also utilized
several data bases in addition to the PDWL to track work and ensure
~
, ,
it was. completed. The" data bases included the " Prime" and "P/2" data
bases. for. tracking 'and following work to completion. Some of the
lists utilized were the Preliminary DWL, the DWL, the Site POD, the
Site Weekly WR Summary, and the STORM, which was used to track outage
work,
b. Outage Scheduling Organization
The Outage Scheduling organization was responsible for the tracking
'of sis, PMs, and WRs; providing advance notice to the Work
Control / Outage organization when work-should be scheduled; and either
providing work packages or ensuring that approved work packages had
been provided to the organizations performing the work. In addition,
this organization was- responsible for outage schedule development.
Another function of the organization was to maintain a history of
maintenance activities that had been performed on plant equipment.
The NRC inspectors' interviews of personnel and in progress work
observation focused on work in the support of sis, PMs, and WRs.
The scheduling organization utilized PM coordinators, WR
coordinators, and SI schedulers to keep data bases up-to-date
regardir.g the plant work to be performed in their respective areas of
re spon sibil ity. In addition to keeping the data bases current, these
personnel either put together the work packages, e.g., PMs, for the
craft people to follow in perforr.41ng the work or ensured that the
packages, e.g. , WRs, had -received the required processirg by others
, before being provided to the craft people for use.
'
c. Maintenance Planning and Technical Organization
This organization was responsible for writing working procedures for
plant work when specific procedures had not been previously written.
-
The planners were required to be knowledgeable of how to perform work
in their area of expertise, e.g., mechanical or electrical equipment,
the equipment configuration in the plant, and the plant design and
technical specification requirements. If they were uncertain of the
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' required work, they had access.to the latest approved plant drawings
.and other approved plant requirements. In 'some cases, the planners
may have lto go to the worksite to. see the required work. in order to
write adequate procedures'.
In addition. to writing- work procedures, the planners worked with .
the maintenance craft persons reviewing the work procedures to
ensure that they understood-the work and apeed that it- could be -
,
accomplished as: written in the - procedures. . Aod'tionally, if ? the
'
craft people were ' to . run -into .significant proble.Ls during the
performance of a' procedure, the craft may have to stop work and have i
the planner revise - the work procedures. - Any revised procedures of -
-this type ' would still be required to undergo strict. approval
processing' before' . implementation of the work. ' If minor procedure.
problems were to. be identified during the performance ofa' work
. package, the. craft may modify the activities, document the change's on
.an ICF, and proceed with the work. These ICFs would be then- routed
back to. the planners so that the problem can be. corrected in future
activities,
d. LDocument Closure
g In additional- to the organizations discussed above, the inspectors
reviewed the operation of the Maintenance ' Department's " Document
Closure Group", where completed SI packages are reviewed for adequacy
and completion prior to being transferred to the QA Department for
review and . storage as QA records. This group's function was to
monitor the interface between the Maintenance Department and QA
Department to ensure that quality products are transferred. This
review identified that approximately 50% of the packages were being
l' ' rejected by the Document Closure Group as not being completed
properly; Most of the rejections were considered minor or of an
administrative nature. Although the packages were being corrected
prior to becoming QA recoros, the inspectors were concerned with this
high rate of rejection, aspecially since PMs and WRs were also
scheduled to begin sienilar review by the: Document Closure Grcup in
early' 1989. This group was considered a.c an example of a positive
use of interfaces .for quelity verification.
L Tne inspectors discussed this problem with mainter.ance managers who
assured the inspectors that the Maintenance Department was also
L concerned with the high rejection rate and was working to resche the h problem. It developed that one of the root causes was that the group p responsible for the administrative instruction had changed it to L require speciff e data on the coversheet turned in with sis, but had '
not changed . over 1200 SIS - each with their own coversheet - to
conform. 'The maintenance - shops were then be'i ng blamed for the
inability to perform procedures. This was a management control l
problem and indicated that interfaces between maintenance groups
'
within the Maintenance Department were not being effectively used as
a quality verification tool . SQN appeared to be taking action to
resolve it and no specific NRC follow-up is planned.
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e. Site Procedures Staff
During the inspection, the inspectors learned that.SQN had. initiated
a new. program to improve site procedures, including sis and PMs. The
l inspectors ' reviewed the program as it existed during the inspection L and the plans for the future. Th'e program included standardization
- and computerization 'of procedures as well as an extensive database
L development ' effort based on existing database information. This
: program was part of an .overall TVA effort to standardize procedures-
throughout TVA and to make them more user friendly. The program at
SQN was scheduled to be completed in 1991'.
f. Summary
The NRC inspectors. performed the .following tasks in evaluating- the
control.of work within the SQN plant.
'
-
Reviewed documents and procedures associated with the work
Li control activities. discussed above (including SI-1, Surveillance
Program . the controlling document for the execution of the
-plant surveillance program).
-
Interviewed personnel in the organizations discussed above
concerning their job activities and observed the performance of
-job activities by some of the personnel interviewed.
-
Observed demonstrations ' of SQN data bases related to work
control and the information contained in the data bases.
-
' Witnessed the performance of SI-37.2, " Containment Spray Pump
1B-B Quarterly Operability Test," following installation of a
new pump rotor.
The inspectors found SQN's' process of plant surveillance
control / system outage control / work control to be well established,
contro*tled by procedures, and working. The process provided for the
identification of work needin;; to be performed, establishment of
procedurer to do the work, scheduling of the work, performance of the
work, and tracking of the work to completion.
.
The inspector found that, within the organizations discussed above,
each organization faad estabibbed internal controls to verify the
. H s >
quality of it's own vork. Tho SQN work control process also required
extensive coordinat' ion between organizations -
which provided
additional checks to verify the quality of the work being performed.
T The SQN process provided for corrections to problems as the work was
being performed. In fact, while the inspectors witnessed the
performance of SI-37.2, workers noticed a packing leak on a valve
used during the SI and documented the condition on a WR for
correction.
o
. - - _ - - . _ _ _ - _ . - _ _ . _ _ _ - - _ _ - - - - . _ _ _ - _ - . _ _ - . _ _ _ . _ . _ - _ - _ . - - . _ _ - _ _ - _ _ - _ _ _ - - _ - - - - - . - _ . _ . - - _ _ _ _ _ _ _ . - _ - _ _ . - - - - _ _ _ _ _ _ _ _ _ . _ . - _ _ _ - - _ - _ _ . - _ _ _ . _ _ _ . _ . - _ . - _ - - _ .-._ _ _ . - . - - . _ _ _
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In addition to the quality verification activities of the organi-
-Iations discussed above, the quality assurance organization was
involved in the approval of certain specific sis, WRs, and PMs, e.g.,
work being performed on TS equipment or systems. As further verifi-
cation of the quality of the work, QA performed audits of the various
organizations to verify the quality: of their work. At the time of
the inspection, QA had performed audits of the Outage Scheduling and
Maintenance Planning and Technical organizations and had an audit
scheduled for the Work Control / Outage organization in 1989. The
inspectors did not review the audits which had been performed.
With the exception of the observation relative to document closure
(paragraph 6.d), the inspectors did not identify any additional
concerns regarding the control of work at SQN.
7. Corrective Action Program
The inspectors reviewed the corrective action program as revised in
September,1988, and documented in the corporate NQAM Part I, Section
2.16, Corrective Action, and SQN procedure AI-12, Part III, Corrective
Action. This program was designed to comply with 10 CFR 50, Appendix B,
Criterion XVI. It consisted of the CAQR program and a number of
administrative control programs. TVA management was responsible for
evaluating adverse conditions ar<d documenting the condition in - the
appropriate corrective action program. This review generally emphasized
the following attributes:
-
Management Review Process
-
DNE Support
-
Involvement of QA/QV in the detection and resolution of problems.
-
Operability Determination
--
Root Cause Evaluation
-
Recurrence Control
-
Distribution of CAQRs/PRDs
-
Evaluation of ACP items and trends to detect CAQRs/PRDs
a. Conditions Adverse To Quality Report (CAQR) Program
As part of the inspectors' evaluation of the CAQR program at SQN,
they discussed the history of the program with TVA personnel.
- Prior to February 1987, TVA had a program for controlling
conditions adverse to quality (CAQs) which focused more on
design problems than non-design CAQs. This program identified
CAQ processing documents under various names such as problem
_ - _ _ _ _ _ _ _ _ _ -
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-
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22
i
identification reports or significant condition reports. At
the ctime of this inspection, SQN was still dealing with approx-
' imately-180 items from this earlier program.
--
LIn- February .1987, TVA/SQN ' implemented the revised corrective -
action program .to identify all .CAQs as CAQRs. .This program
superseded and combined the various earlier programs. While
this program was an improvement to. the old system, problems-
istill existed relative to the identification of CAQs and
obtaining adequate and timely corrective action. In March 1988,
:: TVA/SQN set up a. management review. committee to help process
CAQRs and ensure adequate and timely resolution of identified
'
problems. At. the time of -this inspection, there were
approximately .400 CAQRs (since February,1987) being tracked at
SQN.~
-
Since this program was being overloaded with quality concerns
which were considered of minor. significance compared to. other
' concerns, TVA/SQN, in September, 1988, revised the procedures to .
take credit for administrative control procedures which could
control the less significant concerns. These ACPs addressed,
but were .not limited- to, such things as MRs, drawing
discrepancies, LERs, and responses to external audit reports.
-
-This latest ~ revision provided for the documentation of a problem
(CAQ) on a CAQR-PRD form. Once.the problem was documented on
the CAQR-PRD form, it was evaluated by a management reviewer
from the organization of the individual who documented the
problem. If the management reviewer could not determine whether
the problem .was a CAQR or PRD, the reviewer could take the
problem 'to the MRC for classification. At the time of this
inspection, SQN was taking all submitted CAQR-PRD forms to the
MRC to ensure that an appropriate call was being made on
identified problems. Even though PRDs were considered to be
less safety significant than CAQRs, SQN also tracked the PRDs to
resolution to ensure adequate resolution.
C ' The inspectors' review determined that both TVA corporate and SQN
procedures provided for escalation of CAQs to the highest levels of
management, under certain conditions defined by procedures, if CAQ
classification and resolution were not considered acceptable by the
initiator.
The SQN MRC, which consisted of the management reviewers from onsite
organizations, met every morning (Monday-Friday) to review and
discuss CAQR-PRDs. The inspectors attended one MRC meeting on
December 13, 1988. The inspectors observed the discussions relative
..
to about a dozen different CAQR-PRDs and found the actions taken by
the MRC to be conservative and appropriate to the safety significance
l
of the CAQR-PRDs. SQN personnel pointed out to the inspectors that
l all CAQR-PRDs resolved as " accept as is" or " repair" would be sent to
the PORC and plant manager for approval.
- - - _- _ - - . _ - - _ _ _ _ _ _ _ - _ - _ _ - - _ _ _
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p As - part off the CAQR program: review, ' the inspectors selected- the i
- below-listed CAQRs from a computer 11 sting of- CAQRs. These CAQRs 'I
~
' were reviewed in detail for . appropriate classification and adequacy
l .offresolution. The.CAQRs reviewed were:
p ?SQP 880532 SQP 880533
S0P 880575 SQP 880592
SQP 880570 SQP 880578
SQE 880557901 SQE 880560901
SQA'880576902 SQA 880577902
SQP 880496
Of note, CAQRs SQA- 880576902 and SQA 880577902 identified,. respec-
>
tively, TVA. audit concerns with the disposition of five CAQRs as
" accept as is" without proper- PORC and plant manager approval . and
enine:CAQRs which were classified as PRDs when the TVA auditors.f'elt
theyl should have been CAQRs. The' inspectors reviewed.the resolution-
or pboposed resolution of these two CAQRs, and the other CAQRs listed
'
'above, and. found the resolutions to be conservative and appropriate
to the identified concern.
.The inspectors reviewed the following two periodic reports which SQN
used to maintain an overview of the CAQR-PRD process:
Sequoyah Nuclear Plant - Condition. Adverse to Quality Weekly
'
-
;. Report
-
Sequoyah Nuclear Plant - Quality Assurance Monthly Trend Report
This review.found these reports-to provide the recipients (including
corporate,.SQN, and other site personnel) a good overview of how the.
CAQR-PRD process was working. . The reports provided trending
'information on old ' program CAQs, new program. CAQRs, and PRDs;
information concerning areas, e.g. , design, where the problems were
occurring; as well as other information deemed significant to
controlling the CAQR-PRD process.
The inspectors also reviewed ECP Investigation Report -
ECP-88-CH-185-01, dated December 6,1988, from the Employee Concern
. '
Program. This report raised significant questions concerning "the
ability of employees to process conditions adverse to quality in the
corrective action program." This investigation, which was conducted
from September 14 to November 28, 1988, was initiated because of
employee concerns received at ECP of fices at Knoxville, SQN, and BFN.
;
Since these locations were the only locations with specific concerns, S
the investigation concentrated on implementation of the CAQR-PRD f
program at these locations.
As a result of the investigation, the concerns with the corrective
M action process were partially substantiated. One CAQR (CHS880070) 1
" was written citing several examples of a CAQR-PRD not being issued
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.because~the:effect on quality;could not be determined at the time of
the proposed initiation of the CAQR-PRD. Additionally, the report
b~ ^
made 12 recommendations for the TVA corrective action program. The
report ' documented that a majority of the concerns in the
investigation occurred before- the latest revision of the corrective
. action process in September 1988.
The'~ inspectors held several discussions with ECP personnel, SQN site
R QA' personnel, and corporate NQA personnel regarding the meaning of
the investigation findings and 'the ' corrective actions planned or
already taken to resolve the : concerns. .The inspectors reviewed a-
' draft of the response .to the ' report from the Vice President, NQA to
the Manager, ECP, Based on the discussions and review- of. documents
associated with ' resolving the investigation report concerns, the
inspectors found that TVA and - SQN. were taking the concerns and
recommendations of the report seriously and were actively-working to
resolve' theiissues to everyone's satisfaction. The inspectors.noted
that some of the concerns identif.ied in the report had been corrected
by the September .1988 revision to - the . corrective action program
summa ry.
Based on reviews and discussions with SQN personnel concerning the
~
CAQR-PRD program, the inspectors determined that, while the TVA/SQN
. program was experiencing some implementation problems, the program
was working. The inspectors found that TVA corporate and SQN staffs
i were taking actions to identify problems-- and to address and resolve
: the' problems as they were identified. That SQN had identified
: problems in the program and was working to correct them was evidence
'
that-SQN had QV organizations and methods in place and functioning.
Methods of quality verification included, . but were not limited to,
. trending reports, engineering assurance audits, quality assurance
c audits, and ECP investigations.
The inspectors had no specific new findings regarding the CAQR-PRD
process. In view of the TVA-identified problems being addressed at
the time of the inspection, future NRC inspections will continue to
monitor the TVA and SQN corrective action program. The inspectors
dia ' identi fy problems with the ACPs. These are discussed in the
following paragraphs.
b. Administrative Control Programs
The ACP review consisted of two phases:
-
Review of documents to determine that appropriate licensee
document reviews had been made to ensure that CAQs were not
present. The review found no single items which required CAQRs
to be written. It was concluded that the individual items in
the ACPs had received adequate reviews.
= __
_ _ _ _ _ _ _ _ _ _ _ _
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-
Review .'of. tracking and trending programs to determine ' that
appropriate: licensee trend reviews had been made to ensure that-
'
CAQs were not present.
: The ~ inspector 'revie'wed the ACP documents listed in Appendix' A,
paragraph 2, to determine that appropriate licensee document reviews
.had been made to~ ensure that CAQs were not present. The. review found
no. items whic.h required- CAQRs to be written. It wts concluded that:
the individual items in the ACPs had received adequate. reviews.
The inspector reviewed the tracking and trending programs within the.
ACPs.'
-
Each tracking program was documented and appeared to be adequate
to follow corrective actions to completion. - No examples of
inadequate tracking were found.
-
Regarding the ACP trending programs, the ACPs were. determined to
be part of the CAQR process in September 1988. Therefore, . very
little was: available to trend in some programs. Several ACP
. trending programs were not established until - January 1,1989.
The inspector noted QA involvement in the development and-
standardization of the trending programs. These trending
programs will .be reviewed further during future -inspections, as
part of URI 327,328/88-50-02 discussed below, when more material
~
is available-for review.
The inspector reviewed, in depth, the trending program for the WR
process. TVA utilized the NPRDS computer program and an in-house
program called EQIS to trend WRs. Equipment data and failure
analysis from the WR were entered into. the programs' data files.
~ Periodically the programs were asked to look for previously-
established trigger points intended to indicate potential trends,
e.g., any 8 failures of items from a single vendor in a given year.
These trigger point results were then reviewed by appropriate
engineers to determine if an actual trend existed or if the trigger
represented some other condition, perhaps a procedural inadequacy.
Once a trend was noted, the engineer would determine appropriate
corrective action and assure that it was implemented.
At this point, the maintenance organization was not entering these
trends into the CAQR process. AI-12, Part III, Section 2.1.2.E
stated that CAQRs would be written for " confirmed adverse trends in
activities identified by trend analysis." The maintenance management
stated that the above requirement only dictated that they review
these trends for meeting other criteria of the CAQR process. At the
end of this -inspection, corporate QA and the maintenance department
were working together to resolve this possible procedure violation
and to determine an adequate threshold for entering the CAQR process.
i
4
,.
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- < ...
a-
26
This item is unresolved pending the results of the corporate QA
review and will be reviewed, along with recently-started trending
_
programs discussed.above, during future inspections. This is URI
327, 328/88-50-02.
The June 30, 1988, Semiannual Component Failure Trending Report was
reviewed. It identified 36 components- listed in Appendix A,
paragraph 2, as having trends and documented the corrective actions
taken. Also reviewed were approximately 150 WRs on the
above-discussed and similar components written from July through
December 1988. No examples of continuing trends were noted and the
corrective actions taken were seen as a positive influence on the
system.
8. Maintenance Activities
During this inspection, an in-depth review of the maintenance program was
conducted. This review included corrective maintenance, the predictive
analysis program, and preventive maintenance. The inspection included
observation of work in progress and review of the associated work
documentation. The inspection also included a detailed review of
completed work order packages - including applicable maintenance and
calibration procedures, the vendor manual for each component, and
associated documentation for the completed work. Completed work order
packages were selected based on the importance of the component to plant
safety and to provide a cross-sectional overview of the various types of
maintenance activities. All work reviewed had been completed within the
past two years. The primary focus of this review was to determine the
technical adequacy of the work performed. Specific areas reviewed are
addressed in the following paragraphs:
a. Review of Maintenance in Progress and Completed Corrective
Maintenance.
As previously discussed, inspection sampling was designed to provide
a cross section of maintenance practices. The review of completed
work requests focused on the Unit 2 AFW System. Work reviewed ranged
from a simple gauge replacement to the overhaul of pumps and
motor-operated valves. In general, the following attributes were
reviewed:
-
The work instructions, including the applicable maintenance
procedures, for each WR were reviewed for technical adequacy,
clarity and inclusion of appropriate acceptance criteria.
Adherence to the work instructions and proper sign-offs were
also verified.
-
The vendor manual for each component was compared to the WR and
maintenance instruction to assure that vendor recommendations
had been properly included. This review also included a
____
a .-
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.
27
I
comparison of vendor manual recommendations for preventive
maintenance to the site PM for the equipment.
-
Torquing requirements for system closure fasteners were
verified.
-
The control of calibrated equipment esed to perform the work was
reviewed. This review included verification that the range of
the equipment was appropriate, the equipment was within the
current calibration cycle when used, and that the individual
calibration records showed that the accuracy of the instrument, ,
when calibrated, was within the stated acceptance criteria. l
-
The purchasing documentation for installed materials was
reviewed for the inclusion of required regulatory requirements.
-
The receipt inspection records for installed materials were
reviewed to assure that the materials were certified as
ready-for-installation.
-
Traceability of installed materials from the WR to the purchase
order and receipt inspection records was verified.
-
Performance of the proper PMT was verified. In many cases this
check also included a detailed review of the results of the
testing.
-
The component history was reviewed for component failure trends.
The following specific WRs were included in this portion of the
inspection. A brief description of each work item is included and
any specific problem areas are discussed so that appropriate
corrective actions can be initiated:
-
WR B295059,- AFW pump 2A-A. This work request replaced the i
inboard pump bearing due to a degraded condition found during
'
vibration analysis.
l
No documentation was present to verify that PMT had been accom- i
plished following the bearing replacement. Work instruction
MI-10.4.1 required performance of SI-130.2 as the PMT but para- ;
graph 6.5 of MI-10.4.1, for SI performance verification, had
been marked "N/A". Licensee investigation determined that
SI-130.2 had not been performed, however, a vibration analysis l
in accordance with TI-96 had been performed - but no records of
this test had been included in the WR. Further investigation
determined that the personnel who had marked the PMT
requirements "N/A" had failed to follow paragraph 11.3.6 of SQM
2, Maintenance Management System, Rev. 33, which required: "If
the section responsible for the PMT makes a determination that
the test is not required; a responsible person. . . shall "N/A"
o-- - _ - - - - - - - - - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ - - - _ _-.---.--------------_______----.____________________a
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28
L l 1 '
the PMT, provide - explanatory remarks, . sign, and date in the
appropriate block."
Note: Site CAQR SQP890013 was issued to correct this
: deficiency.
This deficiency is an example of non-willful failure to follow
maintenance procedures. It did not result in degradation of
equipment. It occurred in early 1988 along with other .similar
problems which- tha licensee recognized as requiring generic
corrective action. NRC Violation 327, 328/88-28-01 was later
issued concerning a similar situation in May 1988. LER 2-88022
reported that situation. Widespread training was conducted
under CAQR SQP 880349. Recent NRC reviews have not found this
situation to be common. This violation meets the criteria
specified- in Section V of the NRC Enforcement Policy for not
issuing a Notice of Violation and is not cited.
This issue -is identified as LIV 50-327,328/88-50-03, example 1
of 2, for tracking purposes and is considered closed.
-
WR B262462, Steam Generator #4 Level Indicator 2-LT-003-107F.
This WR'was issued to correct the condition where this level
indicator read greater than 6% higher than the other steam
generator level indicators. The work involved tightening the
fittings on the low pressure sensing line, back filling the
transmitter, and performing an operability check.
During the review of this WR, it was noted that paragraph 5.4.6
of MI-19.1.5 required the maintenance technician to record the
positions of bistables LS-548A, LS-548B, and LS548C as either
" normal" or " tripped". The bistables were recorded as being.in
the " tripped" condition. Later, paragraph 5.4.13 required the
bistables to be returned to the original position found in
paragraph 5.4.6, in this case, " tripped". In lieu of returning
the bistables to the " tripped" position or obtaining a pro- .
cedure change to allow the bistables to be put in the " normal" I
position, the maintenance technician placed the bistables in the
" normal" position in violation of the procedure - then continued
with the remainder of the procedure.
Note: Site CAQR SQP890017 was issued to correct this
deficiency.
This deficiency is an example of non-willful failure to follow I
maintenance procedures. It did not result in degradation of
equipment. It occurred in early 1983 along with other similar
problems which the licensee recognized as requiring generic
corrective action. NRC Violation 327, 328/88-28-01 was later
issued concerning a similar situation in May 1988. LER 2-88022
reported that situation. Widespread training was conducted
_ _ _ _ -
7-
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29
'
+ under CAQR SQP 880349. Recent NRC reviews have not found this
situation to be' common. This violation meets the criteria
~
specified in LSection V of; the NRC Enforcement- Policy ' for not
issuing'a Notice of~ Violation and is not cited.
This issue is identified 'as: LIV 50-327,328/88-50-03, example 2.
'
.of 2, for tracking purposes and is considered to be closed.
_
-
WR B237011, Steam Generator #2 Flow Transmitter 2-FT-003-155.
This WR 'was ~ issued ' to correct an indication problem.'under
3 no-flow conditions.
-
WR B267211, Turbine Driven AFW Pump 2A-S. This WR replaced the
. Unit 2 turbine' driven AFW pump rotating element.'
-
WR B234108, Feedwater Isolation Valve 2-FCV-003-87. This 'WR
corrected this valve's out-of-specification stroke time.
-" WR B239447, AFW Pump 2B-8. This WR replaced the pump thrust
' bearing'due to a degraded condition being indicated by vibration
analysis.
-
WR B267403, AFW Level Control Valve 2-LCV-003-164A. . This WR
L corrected a problem with leak-through when the valve was closed.
-
'WR AS48867, Temperature Indicator on the Turbine Driven AFW Pump
Outboard' Bearing Housing. This WR replaced the subject
temperature indicator due to the temperature readings being high
when the. pump was not running.
-
WR B104839, AFW Piping Welds AFD F25AA, AFD F25BB, and AFD F38.
This.WR was issued to correct unsatisfactory weld defects found
by nondestructive testing of the welds.
!
-
WR B261181, AFW Level Control Valve 2-LCV-003-172A. This WR
corrected a problem with leak-through when the valve was closed.
-
WR B789967, AFW Pump 2B-B Inlet Pressure Indicator. This WR )
recalibrates the indicator to correct inaccurate readings when 4
the pump was not . running, i.e., the indicator indicated 29.5 I
psig vice 16 psig for the condensate storage tank head pressure.
-
WR A298541, Steam Generator #3 AFW Flow Indicator 2-FI-003-147C.
This WR was issued to replace and calibrate the subject flow
indicator. 1
j
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WR B267237, AFW Steam Trap Drain Valve 2-VLV-003-C18S. This WR l
lapped the valve seats.
l- l
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_ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ - _ _ _ - _ - _ _ _ _ _ _ _ _ _ - _ - _ - _ - _ _ _ _ _ __ a
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-
- WR' B218701. Thiss WRf was . issued to correct -aEleak at the-
downstream' flange of the cavitating venturi in the' discharge
-
'
U line from AFW Pump 2A-A. m -
WR B117221, Turbine Driven AFW Pump 2A-S. This WR realigned the
H subject pump. p
-
'WR B128783, AFW Check Valve. 2-VLV-003-873B. This ~ - WR
L. disassembled and. inspected the check 1 valve to sati sfy : the
requirements of IE Notice 86-09.
Note: .The vendor manual for this valve was. not avaliable on
site.
-
. WR B234138, AFW Level Control Valve 2-LCV-003-156A. This WR
corrected this valve's failure to meet its required stroke time
and involved disassembly, cleaning, inspection and lubrication
of the valve operator.
-
WR B218703, Current Meter 2-EI-003-1198 for AFW Pump 2A-A. This
WR calibrated the meter.
-
WR B219827, Motor Driven Suction Valve 2-FCV-003-116A from ERCW
to AFW Pump 2A-A. This WR disassembled the valve for internal
pipe-weld inspection.
Note: The vendor manual for this valve was not available on
site.
-
WR B267624, AFW Level Control Valve' 2-LCV-003-1738. This WR
replaced the valve positioner.
-
WR B784804,- Containment Spray Pump IB-B. This WR replaced the
pump's rotating assembly due to a degrading condition identified
by vibration analysis.
Note: A portion of this work was observed by the inspectors.
'
Two particular concerns were identified during a review of the
pump coupling vendor manual, i.e., K0P-Flex Inc. Service Manual
1900-01 of June,1987, as follows:
a. The coupling vendor manual, in the section entitled
" Dynamically Balanced Couplings", states that each bolt,
J nut and lockwasher must be maintained as a set to assure
proper coupling balance. Additionally, this same section j
requires that the coupling be assembled observing the i
coupling match marks. These requirements were not included !
in site procedure MI-6.8, Containment Spray Pump :
Inspection, Revision 6, the MI for disassembly and
reassembly of the containment spray pump. Investigation 3
.
_ _ _ __ ___ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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-31
determined that the requirements were covered in craft
training,. however, it is believed that including these
requirements ,in the' applicable site procedures would-
'
enhance the procedures and. possibly prevent rework'in the
,
future.
Note: Several .Sequoyah pumps utilize this type of
'- coupling.
b. TheLvendor coupling manual coupling assembly section, sheet
2, required torquing of the ' coupling fasteners and the
lubrication plugs to specified torque values. These torque
values had not been specified in MI-6.8, Rev. 6, as being
applicable to the work. on WR B784804. Further
investigation. determined that. no engineering evaluation of - '
these torque' values had been performed by the site.
The lack of inclusion of these torque values in site
procedures: and the lack- of an engineering evaluation .,
'
justi fying the_ exclusion of these torque values is
unresolved. pending engineering- evaluation -(URI
327,328/88-50-04).
'During a tour of the Auxiliary Building, the inspector. observed'that
a limit. swit'ch was ' misaligned with the actuator arm and 'would not
operate l as required on the containment spray pump room A cooler
supply control- valve 1-FCV-67-184. -A .second limit switch
2-FCV-67-176 was also misaligned.
The licensee provided the following information relative to this
issue:
- The function of the limit switch was to indicate valve status by
operating lights on panels OM-27A.
-
The consequences of the limit switch not activating would be l
'
that the valve position would have to be visually verified.
-
The valves are checked every 92 days by performing SI-166, which
full -strokes certain valves during operation. Valves FCV-67-176
and FCV-67-184 are required to be stroked on a 92 day period.
-
The SI would identify misalignment problems because the lights
would not operate properly.
-
On December 14, 1988, WRs B283176 and B283172 were issued to
realign the limit switches.
Numerous other limit switches were observed during various tours of
the facility and no other discrepancies were observed. The inspector
concluded that the misalignment probably occurred as a result of the
- _ - _ _ _ _ _ _ _ _ _ _ _ 1
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i
switch being inadvertently bumped:during' work activities in the area.-
,'l The surveillance program appeared.to be-adequate and this. appeared to
be an~ isolated occurrence.
,
b. Predictive Analysis
.The predictive' analysis program was also reviewed during -this
inspection. ' The" licensee. primarily. used three Ltypes' of- predictive
. analysis to anticipate component failures. . Lubricating oil analysis
and vibration analysis were_ used to predict failures in : rotating.
equipment. M0 VATS testing was used to predict motor-operated valve
' failures.
- The vibration ' analysis program has played a. very active role in
predicting component failures at Sequoyah. This was evidenced by
three of the- corrective maintenance WRs on .the AFW pumps and the-
containment ' spray. pumo, reviewed at ~ random by the inspector, being -
1 the result of degrading pump performance.being found during vibration.
analysis. ' Vibration analysis was' an integral part of the site's ASME
Code Section XI pump testing program and was also an integral element
of.the PMT~following any maintenance which would. change-the vibration
. signature -of the equipment. The program was started in 1979 and
computerized data trending became fully developed by-about 1983.
I
MOVATS motor operated valve testing also played a very active part-in
the' Sequoyah maintenance program. The program:was started about 3
1/2 years ago -and has expanded' to coverEnearly 375 safety-related
valves. Site procedures required MOVATS testing any time maintenance
:affected MOV operability. Additionally, MOVATS testing was
accomplished by the PM program. A sampling of Unit 2 AFW system
valve data determined that 13 valves were in the program:and most of
,4 these valves had, in fact, 'been tested twice. ' Seventy' five valves-
per unit require mandatory- outage testing due. to NRC commitments made
<
because the valve motor thermal overloads were by passed.
1
At the time of the inspection, enh one M0 VATS engineer was on' site.
Site management representatives indicated that they intended to
provide additional personnel to accomplish the required testing
during the impending Unit 2 outage.
j:
The lubricating oil analysis program, on the other hand, had not yet
been developed to the extent that the vibration analysis and MOVATS
'
i l programs had been. Though lubricating oil analysis was being
accomplished as required by site PMs, the site had not yet
accomplished the close integration of the vibration analysis program,
the lubricating oil analysis program, and' the periodic oil change
outs specified by various PMs. The PM program manager indicated ,
plans to take future actions to strengthen this area. j
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c. - Preventive Maintenance
During this inspection, ' the. licensee's PM program was reviewed to
.- o _ determine the extent of PM actually being performed. This portion of
- the inspection also focused primarily on AFW system components, with
r
additional emphasis on personnel airlocks. The inspection was
,
. accomplished- by comparing vendor manual PM - requirements /
recommendations to - site PM instructions, reviewing adherence to
. licensee ' and vendor-established PM intervals, reviewing the PM
,
upgrade program, and discussion of the program with licensee
[ - personnel.
The .SQN' PM ~ program was in a :significant period of change. In
October, 1987,' the -site had developed approximately 2400 PMs. In
January,1989, there were over 3900 PMs, i.e. , a jump of over 1500
- PMs in just 15. months. This significant increase was attributed' to
the upgrade - program efforts by both TVA and contractor personnel .
The upgrade program was being accomplished through a very detailed
review of various data bases; e.g., NPRDS data, Sequoyah Failure-
History,'IE Notices and Bulletins, vendor manuals, etc.; to determine
l .
PM.l requirements for generic plant equipment items - followed by .
-
development of PM instructions to accomplish the requirements. The
licensee's effort in this area is noteworthy since, once completely
L implemented, Sequoyah will have a very strong PM program.
However, two areas of concern were noted:
-
The ' by-weekly . listing of delinquent PM's, dated January 11,
1989, (A delinquent PM is one which has exceeded its 25 percent
grace period) listed 330 delinquent 'PMs - 146 of the 330
involved safety-related equipment.
-
Of 55 required PMs on the Unit 2 AFW system, 41 had never been
accomplished on the equipment. Also, of the 55 PMs, about 30,
or over half -of the PMs, had been developed in the last 15
months. If these statistics hold true for the other plant
systems, then the PM workload indicated by the total number of
delinquencies may be grossly understated since many of the new
PMs have not been scheduled. Additionally, although the
licensee has been very aggressive in developing PMs, the
equipment in the plant has not been subjected to the PM, thus
the benefit to plant equipment has not yet been realized.
In December 1988 the licensee recognized the delinquent PM problem ;
- and initiated the following actions to reverse the previous year's
increasing trend. These corrective actions have not been implemented
long enough to' determine their effectiveness:
-
A PM analyst staff of 3 engineers was established.
-
Maintenance discipline managers were made accountable for
performance of PMs by tying PM performance to the personal
performance evaluation.
.
hm__au_m______-m_.__m_m.-m___________..___m_-___.___ _ _ _ _ _ _ - . _ _ _ _ _ _ _ _ - _ _ _ _ . _ . _ - _ _ . _ _ _ _ _ _ . ___. _ - _ _ _ _
.
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Electrical maintenance was provided an act' ion. plan'to reduce PM
delinquencies.
-
:The PM administrative program was revised toistrengthen the
control of.PM performance.
y
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- A feedback' system was established to improve PM tasks and .
correct problems with' proper identification of PMs as outage or
.non-outage.
-
An. integrated -schedule we.s being developed - which -should
. minimize delay.
-
. Efforts'were ongoing to level the PM workload throughout the
.
year.
-
.The PM work flow path had been improved to minimize
administrative delays.
Specific PM activities reviewed also included those discussed below:
-
PM' Instruction 1897, Upper Head Injection Level Switches. This
PM accomplished a set point check of the UHI valve closure
switches.
Note: .A portion of this PM was observed by the inspectors.
.This PM ;was accomplished- under a TS one-hour LCO. Personnel
. performing the PM were very knowledgeable and the PM was
performed in accordance.with the procedural. instructions. Three
problems did occur during this work and these problems were
properly dispositioned by PR0s, as discussed below:
-
PRO 2-89-005 reported, as required by.the PM instructions,
that switch 2-LS-87-21 was out of the range specified by
TS.
-
PRO 2-89-006 reported that the one-hour LCO time limit for
performance of this PM had been exceeded by six minutes.
-
PRO 2-89-007 reported an actuation of valve 2-FCV-87-23,
l which was caused by valving-in switch 2-LS-87-21 too '
rapidly during the return of the switch to service.
L -
The PMs for the personnel airlocks were reviewed. Vendors Manu-
l al SQN-VTM-C310-0030 " Personnel Airlock, Contract 92120 Fabric- l- ator, Chicago Bridge and Iron Company" contained maintenance l
requirements in the following areas: 1
-
Shaft Seals - The shaft seals are teflon packing and
require no maintenance.
i
Q_--_-- __- . _ - _ - _ _ _ _ _ - - _ _ _ _ - _ _ _ _ - - _ _ _ _ - _ _ - _ - _ _ _ _ _ . __ ___- ____ . _ _ _ _ _ - - _ _ _ _ - . _ _-_ _ _
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Bearings and Flange Block Lubrication
-
. General - lubrication of gears,. sprockets', and chains and
bushings
The -inspector reviewed the PM performance data for both Unit 1-
and 2 reactor building airlock doors, as listed below: ,.
-
Unit 1 Lower- Airlock: 2-19-87, 8-15-87, 2-07-88, 8-17-88 _
-
- Unit 1. Upper Airlock: 2-16-87, 8-16-87, 2-08-88, 8-17-88:
-
Unit 2 Lower Ai_rlock: 2-15-87,,8-17-87, 2-01-88, 8-29-88-
:- Unit 2 Upper Airlock: 2-17-87, 8-17-87, 2-01-88, 8-29-88
The PMs' referenced above were performed as required and complied
with the vendor manual.
L. 9. Safety. Information Management System -/ Corporate Commitment Tracking
System
a. . Safety'Information Management-System
During this . inspection period, the inspectors reviewed the accuracy
of licensee input. to the SIMS - system and 'the licensee's use of the -
_
CCTS system to. manage commitments made'to the NRC, SIMS is an NRC
-
- database system which provides NRC. management with a single source of
reliable information- on NRC's management of generic and
plant-specific s afety issues. It .contains, in part, the
' licensee-supplied data -that shows the status of - the licensee's
implementation' of safety ' issues which are ' resolved by safety
evaluations issued by the NRC staff. For Sequoyah, the SIMS database-
contains data on the following issues:
-
Licensee employee concern element reports.
L
-
NRC Multiplant Actions - including the TMI action plan in
NUREG-0737.
-
License amendments, which are primarily amendments to the Units
1 and 2 TS.
-
The remaining plant-specific issues.
SIMS was not a restart issue for the Unit 1 and 2 restarts from their
recent extended outages. The last update of the SIMS for SQN on the
licensee's implementation of safety issues was with data in the I
licensee's letter dated February 17, 1988. The licensee will be
providing the NRC another SIMS update by a letter scheduled to be ,
submitted by February 15, 1989. j
l
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During this inspection ~ period, the inspectors reviewed the SIMS
database update data planned to be provided by the licensee in the
above letter.
The inspectors reviewed files relating to the employee concern
element reports for Sequoyah, the amendments to the TS, and the TMI
-~
'
action plan for Sequoyah. The files relating to the employee concern
element reports for Sequoyah were reviewed at the TVA offices in
Chattanooga. The other files were reviewed at the Sequoyah site.
The data to be submitted to the NRC in February, 1989, is to be
compiled from the files that were reviewed. The inspector concluded
from this review that the licensee is providing accurate input data
for the SIMS database.
b. Corporate Commitment Tracking System
The CCTS is the licensee's system for tracking commitments to the
NRC. This inspection reviewed the CCTS system for the following
attributes:
-
The process to add commitments.
-
Tracking of commitment completion status.
-
Late responses.
-
Verification of commitment completion.
-
Commitment closure documentation.
-
Closure document errors.
-
QA or QM surveillance and audits.
CCTS discussions were held at the corporate headquarters with
individuals who control CCTS and at the Sequoyah site with site
licensing coordinators and site CCTS coordinators who use the system.
Procedures reviewed included:
-
ONP Standard ONP-STD-6.1.1, Rev. 0-C, Managing and Tracking NRC
commitments, dated September 21, 1988.
-
Procedure 0605.01, Commitment Management and Tracking, Rev. O,
dated January 13, 1987.
-
Sequoyah Standard Practice SQA-135, Commitment Management,
Tracking and Closure, Rev. 8, dated December 22, 1988.
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l. L
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Sequoyah Site Licensing Staff Section Instruction Letter
SLS-SIL-03, Rev.1, Status Management, Escalation, Trending 'and
Distribution of NRC Commitments, dated January 20, 1989.
Thel 'following ' CCTS closure packages, completed by TVAL during the
period from July I to September 30, 1988, were reviewed:
.NCO-85-0102-001 NCO-85-0522-001
NCO-85-0527-001- NCO-86-0156-159
NCO-86-0282-006 NCO-86-0471-007
NC0-87-0226-001 NCO-87-0232-001
NC0-87-0260-002 NC0-87-0356-005
NCO-88-0102-001 NCO-88-0119-005
NCO-88-0127-005 NCO-88-0136-001
~ Commitments are made through letters, through LERs from the.licen'see
to the NRC, and during licensee /NRC meetings or telephone conference:
: calls. The licensee had procedures to formalize these commitments.
and input. them into CCTS. . The licensee's processito formalize a
commitment to the NRC included developing an action plan with the -
estimated cost and manpower to complete the commitment. Licensee
personnel stated that, because the commitment action plan put the
cost of completing the commitment into the budget, there was an ~
incentive for 'the licensee to identify all commitments made to the
NRC. The inspectors concluded that the procedures were acceptable to
ensure that commitments made by the licensee to the NRC would be
tracked by CCTS.
'Sequoyah site licensing personnel explained the format of the CCTS
computer printout and the different reports made at Sequoyah to track
the status of the CCTS items, to assure that the commitments would be
met on the agreed-upon schedule, and to identify late commitments.
The inspectors concluded that the licensee has been acceptably
tracking the status of CCTS items to assure timely completion.
The selected completed CCTS commitment packages listed above were
reviewed, and site licensing and CCTS coordination individuals were
interviewed concerning CCTS commitment closure methods. Examples of ;
'
QA audits of CCTS closure packages were also reviewed. Ten percent
of the CCTS closure packages were routinely sampled by QA, in
addition to other closure packages specifically identified by Site
Licensing for QA verification. QA trending did identi fy some
problems with the commitment closure process. However,
recently-developed TVA Section Instruction Letter SLS-SIL-03, Rev.1, l
'
of January 20, 1989, provides more explicit guidance to appropriate
staff personnel to further streamline the CCTS review and
verification process. Continued training using these newly revised
guidelines, plus the QA trending audits, should provide the licensee
with continuing confidence that proper CCTS closure is occurring.
I
_ _ - _ - - _ _ _ _ _ . - _ . _ _ _ _ _ _ _ _ _ _ __ _Y
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Within the scope of this inspection, the inspectors concluded that
"
the-licensee has been acceptably addressing commitments made .to the
NRC.
10. Plant Modification Process
This portion of the inspection focused on modification workplans,
observation of field work, and QM activities in the. modification area.
a. Modification Workplans
Discussions with Modification personnel revealed there were
approximately 1300 workplans remaining . open, some dating back to
1980. All required physical work was completed on these workplans
. prior to plant startup, however, the work plans were left open for
the various reasons indicated below:
-
255 Workplans were status WI "Workplan Implementation". The
workplan was being implemented, and the cognizant engineer had
.the workplan.
-- 106 Workplans were status HPC " Hold - Partial Complete". Work
had been started but not completed. -The document coordinator
had the workplan in the hold file.
-
~ 15 Workplans were status HM " Hold for Material". The document
coordinator had the workplan in the hold file.
-
14 Workplans were status HPM - " Hold for Manpower". The.
document coordinator had the workplan in the. hold file.
-
. 390 Workplans were status- FC "Workplan in Field - Complete".
The cognizant engineer had-39 days to complete the documentation
and turn in the workplan.
-
489 Workplans were status DC "Workplan Sent to Drawing Control
Center for Drawing Update".
-
3 Workplans were status DCP "Workplan Sent to Drawing Control
for a Partial Update".
-
33 Workplans were status ANI " Authorized Nuclear Inspector".
The ANII had the completed workplan for review and sign-off.
-
206 Workplans were status WDR -
"Workplan Documentation
Resolution". The completed workplan had been returned to the
cognizant engineer for documentation corrections.
The inspector selected six workplans for review:
-
Workplan 7346-01 - Issued in April 1988
. _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ - _ _ _ _ _ - _ - _ - _ - _ _ - _ _ _ - _ _ _ _ _ - - _ _ _ - _ _ _ _ _ - _ _ _ - - - _ -____
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Workplans 9557, 9560, 8652 - Issued in 1981
-
Workplan 11788 - Issued in 1985
-
Workplan 12618 - Issued in August 1987
l
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Workplans 9557 and 9560 required the changing of carbon. steel
lines to stainless steel on the ERCW 2-inch diameter and smaller
lines. They were issued in 1981 and some work was accomplished
through 1984. Other piping change-outs described on these
workplans had never been scheduled for work. The inspector's
review did not identify any safety issues caused by these
workplans being left open. However, some items that required
licensee action were incomplete. For example, of 28 weld data
sheets reviewed, 26 were not confirmed by QC, as required, to ,
assure that the welders were properly qualified. l
1
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Workplan 7346-01 was issued to replace existing normal feeder j
cables-for the turbine driven AFW pump and to replace cables for -
the turbine driven AFW pump room DC vent motor and starter. The
field work was completed on September 27, 1988. At that time,
the cognizant engineer sent the secondar.y drawings to drawing
control for revision. The workplan will remain open until the
drawing revisions are completed. Procedures SQEP-30 and SQEP-42
required that secondary drawings be updated (revised) within 90
days. The drawing ' revisions should have been completed by
December 27, 1988. On January 24, 1989, the completion notice
had not been returned to modifications.
The licensee advised that CAQR No. SQP880126, Rev. 1, issued
February 16, 1988, identified a problem in the drawing update
program. The corrective action, issued January 4,1989, stated
that: ,
1) Primary Drawings: The Sequoyah Engineering Project will
develop an overall revision plan to encompass resolution of
all [the] primary drawing backlog in order to prevent
recurrence of redlines on the primary drawings.
2) The Sequoyah Engineering Project shall meet the 90 day
timeframe for secondary drawing revisions as designated in
site procedures SQEP-30 and SQEP-42 for all work received
on or later than December 1, 1988.
Scheduled completion date: March 1, 1989.
3) The Sequoyah Engineering Project will develop an overall
secondary drawing revision backlog plan (work received
prior to December 1, 1988). Currently this backlog is
approximately 5,100 revisions. This plan shall be
established based on budget and schedule for ECN/DCN
I
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closure. A scheduled completion date cannot be determined
at> this ' time due to pending budget / sche'dule' approvals . by
'the. senior vice president of nuclear power in April 1989.
A schedule. shall be prepared within 60 days of the'..
' established budget,- noted above, .to encompass all of the
,
. secondary drawing backlog.
f. Based on the above, it could not be determined when this
workplan will.be completed and closed. Some line items remained
< open. ' The engineer advised that they would be addressed when
-
the workplan was reviewed for closure.
The inspector's review of the workplan did not identify any
safety' issues.-
-
Workplan 12618 was issued August 14, 1987 to ' replace the.
.non qualified discharge pressure gauges for AFW pump A-A and B-B
with qualified seismic -1(L) pressure gauges. The work was
completed and the workplan closed on October 14', 1988. The
1nspector's-workplan ~ review included seismic qualification
report SQEP-C2-L7156. " Seismic Qualification of Auxiliary
Feedwater Pressure Gauges." This document justified the seismic
adequacy'of the new gauges. All items reviewed on this workplan
were found acceptable.
,
,
--
Workplan 8652 was issued May 22, 1980 to replace the ice
condenser air handling unit three-way glycol valves with
two position solenoid valves. This work was completed shortly
-after the work plan was issued. However, the workplan required a
.PMT be performed per procedure MTI-11 " Ice Condenser Air
Handling Unit Performance". This PMT was . never. performed.
Further,' MTI-11 has since become an obsolete procedure. The
workplan remains open with no planned completion date.
-
Workplan 11788 was issued to move two pressure transmitters and
one level transmitter from inside the crane wall to inside of
the incore instrumentation room. The pressure transmitters were
identified as 1-PT-68-322 and 1-PT-68-323. The level
transmitter was identified as 1-LT-68-320. Approximately 91
welds were completed on this worKplan in 1985. In each case,
the workp'lan required a QC post work review to verify that the
welders were properly qualified to perform the applicable weld
"
procedure. This verification review had yet to be performed in I
1989.
After QC completes the welder qualification verification, the
cognizant engineer is to review and sign the weld data sheets.
This had not been done.
-_-___=___-_:____- . _ - - _ _ _ . _ _. _ . - _ _ - _ _ _ _ _ _ _ _. _ _ _ _ _ _ !
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- As : indicated in ' th'e workplan on August 29, 1985, the ANII
" required a review of the complete package. This - had not been
.
done.
- On September 21, 1988, a Partial Modification Completion Form
was' completed for the subject' workplan. It identified the work-
remainingL to be a leak check of new piping 'and tubing and
: calibration of the instruments. .The partial modification
! package was approved on September 25, 1988.
The inspector met with the onsite ANII to discuss the
requirements for completing the reviews he had specified on the
: workpl ar.s .
1
-
AS!!E Code Section XI contains the ANII's authority and
-jurisdictional boundaries. The code does not specify that
the ANII review the workplan prior to placing the unit in
-service.
-
ASME Code Section XI does specify the following:
Paragraph IWA 6250 requires that records and reports of
repairs and replacements shall -be prepared . . _ and filed
with the . . . regulatory authorities having- jurisdiction at
the plant site. Paragraph IWA-6220 requires that the
inservice inspection reports shall be filed with the
regulatory authorities ... within 90 days after completion
of the inservice' inspection.
Discussions with -the ANII and mechanical test engineers show a
large . number of outstanding workplans which could involve ASME
Code Section _XI, Class 1 or 2 activities, and require ANII
review before submittal to the NRC. These reviews had to occur
within 90 days after completion of the inservice inspection (end.
of the refueling outage) so that the report could be submitted
as required. The licensee advised that a nonconforming
condition, discussed below, was being issued in this area. This
concern was identified by the licensee.
CAQR SQP880607 was prepared January 13, 1989 for Unit 1 and was
in the review cycle prior to issuance at the end of the
inspection. It identifies the following deficiencies regarding
compliance with ASME Section XI:
-
All summary reports of repairs and replacements to ASME
Code Class 1 & 2 pressure-retaining components and their
supports and pressure retaining parts of components have
not been submitted within 30 days after Unit i refueling
i outage. Note: The 30 day requirement is an internal l requirement so that the 90 day reporting requirements are
met.
1
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Several ASME work requests / maintenance requests have been
microfilmed as completed QA records without obtaining ANII
final review of the package.
The CAQR identifies that approximately 49 percent of the ' ASME k
Code Section X7 work packages have not been reviewed or approved
by the ANII or QA, as requided for package closure. The CAQR
indicates that February 5,1989, is the required date for the
Unit I refueling outage summary reports to ba submitted to the.
NRC, but only about 51 percent of the reports will be completed.
The remaining 49 percent of the reports will be submitted 90
days later on May 6, 1989.
The CAQR apparent cause analysis identified that, due to the
large number of workplans and WRs that were worked to support
Unit 1 and Unit 2 restarts and the priorities of field wo'rk
completion that were set, not all workplans and WRs were
administratively and procedurally closed out. Therefore, the
reviews, corrections, approvals and summary reports were not
accomplished as the field work was completed - which caused the
delays and an extensive backlog.
Additionally, a memorandum dated December 19, 1988 (RIMS S08
881219 805) identified the following conditions relative to this
issue:
$
-
No tracking system existed for ASME Code Section XI WPs, ;
WRs, or MRs that required ANIl review and involvement.
-
Lack of defined responsibilities in the preparation of
documentation for repairs and replacement.
-
No compiled list of ASME Code Section XI work during the :
Unit 1 Cycle 3 (April 1984 - November 10,1988) timeframe.
-
Different [various] implementing procedures. I
Though this inspector's review has not identified any issues
which would directly affect safety, the inspector did determine,
based on this review and extensive discussions with engineers
regarding the large backlog of uncompleted workplans and
uncompleted ASME Code Section XI reviews, that the fcilowing
management control oversights exist concerning workplans, WRs, "
and MRs:
i
-
The large backlog of workplans exists primarily due to the -l
large volume of modifications at the site. However, better 6
administrative controls should have been in place to l
systematically complete the verifications and paperwork
after the field work was completed. For example, welds
that were completed in 1985 should have received
a
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43
confirmation that the' welders were qualified shortly'after
completion of the welding. Completion of the-items at thes
time they occurred would have prevented some of - the
backlog.
--
Failure to implement status contro.ls for inservice
inspection of repairs and replacement, along with the
failure to complete workplan processing thru the ANII and
QA review cycle, could result in a failure to comply with
10 CFR 50.55(a.)- requirements if prompt action. were - not
taken.
This is identified as URI 327,328/88-50-05 " Completion of
Workplan Review and Reporting of. Section XI- Repair and
Replacement" pending the licensee's resolution of these issues.
b .- Observation of Field Work
The inspector conducted several tours of the Auxiliary Building and
' reviewed work being performed by the licensee's modification group.
The. modification activity reviewed involved welding of- the -16-inch
diameter stainless steel- piping serving two new - CCW HXs. The
~
. licensee:was replacing one' existing CCW HX with .two smaller units.
The ' welding activity' in progress at the ; time of the inspection was
the fitup-and tack welding of the 16-inch diameter, 0.375. inch wall-
thickness, stainless' steel pipe. Tack welding was physica'lly . in
progress.
The inspector interviewed the welder and welder foreman.regarding the
welding. paramaters required- for welding. stainless steel . The welder
advised that'the following parameters-were applicable for the weld in
progress: preheat was +60 F,. interpass temperature was controlled at
500 F maximum, weld rod was 1/8" diameter, and weld material was type
E316. The welder also advised that -the welding procedure was GT '
88-0-3.
Review of the welding parameters established in the qualification of
GT 88-0-3 showed that the maximum interpass temperature was
established at' 350 F and not at 500 F as stated by the welder. All
of-the other stated parameters were correct.
-Interviews of the welder, welder foreman, modification engineer, QC
inspector, and QA surveillance inspector regarding interpass
temperature controls established the following:
-
The welder stated he did not use Tempilstiks to check interpass
L temperatures.
-
The digital thermometer found at the weld station had a range of
0-300 F and was only effective for measuring preheat
temperature.
)
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.
QC inspectors did not verify interpass temperatures.
'
-
-
QM monitors did not perform interpass temperature checks,
i
'
-
Early QM monitoring records did not indicate that stainless
steel interpass temperatures were verified in the past. Recent
reports for monitoring performed after the issue was identified
showed interpass -temperature controls to be in effect and so
noted in the report.
-
The modification group contained a weld engineering section
which had been performing in process monitoring of welding using
a draft procedure.
The inspector reviewed numerous reports titled " Welding - Activity
Verification Report" which had been conducted by the weld enginee' ring
section in accordance with draft procedure DNC-GCP 8.1.4-02
" Verification of Procedural and Specification Requirements for DNC
Welding Activities." At the time of the inspector's review, this
document had not received final approval and was not issued. Prior
to the exit meeting, the licensee advised that the procedure would
not be issued for 2-3 months and, during the interim period, an SQA
procedure would be issued - but the SQA procedure planned issue date
was also unknown at the time. A review of past reports found that
the weld. engineers had not specifically indicated on the report that
interpass temperatures had been checked. However, numerous reports
issued after the concern was identified by the inspector specifically
addressed interpass temperature controls.
The licenses immediately implemented a training program on January
11-12, 1989 for all boilermakers, steam fitters, and foremen involved
with welding activities on site. The training was titled " Welding -
Interpass Temperature". A total of 51 employees received the
training.
Additionally, the licensee was conducting further training of welders
and foreman for understanding welding symbols and the details of
welding procedures. This training was scheduled for completion in
Jan ua ry , 1989.
On January 11, 1989, the QC Supervisor directed that effective
January 17, 1989, the Welding QC Section would implement an informal
weld surveillance program. The welding surveillance program
included, among other things, an ettribute for verifying interpass
temperature controls during welding. ;
The inspectors concern regarding the need to control interpass
temperatures was that elevated temperature tends to increase the
sensitization of austenitic stainless steel, and if combined with
causative conditions, intergranular stress corrosion cracking might
occur. The licensee provided the inspector with a report from the
- - _ _ _ _ _ _ . - _ _ . _ - .
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' licensee's metallurgical engineering ~ group that had conducted
metallurgical analysis of _ weld samples that were welded to the
parameters of welding procedure GT 88-0-3 except that- the interpass
temperatures were intentionally raised- to greater' than 800 F. The
tests .had been performed -to' address an NRC Notice of Violation at
Watts Bar involving similar issues. The welds had been subjected to
a ' corrosion test in accordance with ASTM A262, Practice A. The
samples were then sectioned, polished, and etched using the ASTM A262'
.
procedure. The - results for butt, welds were reported ' as followsi
"Neither weldment shows complete sensitization as evidenced by grains
"
completely surrounded by ditches. ;Both specimens show partial
ditching of 'the grain boundary, which 'is an acceptable-microstructure
under the conditions of A262, Practice A." The inspector's review of
these tests concluded that some sensitization- occurred at increased
elevated temperatures, which was undesirable, however, failure would
not. occur as a result of the increased temperatures.
The inspector concluded, based on the interviews and . records
reviewed,'that although~interpass temperature controls were not being
assured in the past, adequate controls were being implemented by the
~
end of the inspection. Further, the ASTM A262 test determined that
-the welds would not' fail when ' welded at elevated interpass
temperatures.
Failure to follow weld procedure GT 88-0-3 for control of interpass
temperature is a severity level V violation. This review has found
that it was not willful, not similar to prior violations for which
. corrective actions have not been sufficient to prevent recurrence,
and that the licensee has taken extensive corrective action prior to
-the end of the inspection. This . violation meets the criteria
specified in Section V of'the NRC Enforcement Policy for not issuing
a Notice of Violation and is not cited. This issue will be. tracked
as LIV 327, 328/88-50-06 and is considered to be closed.
The inspector audited the activities associated with a welder
perforaing qualification test welding in accordance with ASME Code
Section IX. The weld procedure and procedure qualification test
requirements were at the test station. The weld engineering section
personnel administering the test were very knowledgeable of the
requirements for qualifying welders. Additionally the inspector
witnessed the sample preparation and the performance of the required
side bend test.
<
.The areas reviewed by the inspector were found acceptable.
I
c. Quality Monitoring Activities - Modifications
The inspector reviewed the licensee's QM program in the area of
modifications, including interviews with the quality manager and
l quality supervisor. Additionally, the inspector accompanied a QM '
monitor on a field monitoring activity of in process welder
i
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46
qualification. The QM monitor was knowledgeable of the area of
welding and ' welder qualification requirements. The items specified
on the checklist were followed and appropriately marked. Areas not
applicable were also marked on the checklist as required.
The inspector also reviewed 10 completed QM reports on mechanical
activities. All areas reviewed were found acceptable.
The review found the QM' program to be a very effective means for the
licensee to self monitor and implement corrective actions, when
needed, on areas previewed as potential problem areas. Additionally,
it was found to be a very effective tool for the licensee to verify
that acceptable areas' continue that way. The monitoring activities
reviewed were well planned, and were being adequately implemented.
They appeared.to be an excellent management tool.
d. RHR Sump Valve Room
The inspector reviewed selected listed items relative to the
licensee's basis for determining that an adequate margin of safety
exists in the RHR sump valve rooms as presently constructed. The
rooms were originally intended to meet containment building post-LOCA
conditions. As presently constructed, the rooms, adjacent to the
containment, are not fully able to withstand containment post-LOCA
conditions because electrical and mechanical penetrations into the
room from the auxiliary building are not leaktight. The licensee had
changed the FSAR in 1988. The licensee reviewed the issue and
established the position that catastrophic sump line or sump valve
failure, which could result in large leakage into this room, were not
credible based on the following:
-
Remote possibility of line failure
-
Weld integrity adequacy
-
Valve body failure probability being very remote
-
Postulated leakage from minor stem packing leakage being <
determined to be virtually no leakage
1
The following documents relative to this issue were reviewed: l
1
-
Ultrasonic examination reports for welds RHR-1-5 R1, RHR-1-6. !
(Weld RHR-1-6 was rejected for lack of fusion, repaired and
reinspected, and found to be acceptable.)
-
Detail Weld Procedure GT18-0-1, Rev. 6
I
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Weld History Records for Welds RHR-2-4, RHR-1-3, RHR-1-4, i
RHR-2-3
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Process Specification '3.M.7.1(2) " Specification For. Ultrasonic
h Examination of Weld Joints"
All areas reviewed by the-inspector were found acceptable.
.' 11. QAlRcutineAudits/QCActivities/andSpecialSurveillances(orMonitoring
Activities) in Support of Operations
a. ' Quality Assurance Audit Organization
The- NRC inspectors reviewed the following site . Quality Assurance
audits which were either complete orlin progress:
SQA.87-0003,. Mechanical Maintenance
SQA 87-0015, Electrical Maintenance
SQA:87-0018, Plant Modifications and Design Control
SQA 88-804, Independent Qualified Review Process
SQA 88-808, Compliance'with Technical Specifications
SQA 88-811, Instrument Maintenance
SQA 88-815, Ongoing audit on various programmatic areas
associated with the CVCS system.
CMAP Audit, performed in June 1988, covered the Quality Audit
and Monitoring Programs
The NRC-inspector noted that the Quality Audit reports appeared-to be
thorough, containing an adequate level of detail'to stand alone as QA
records, and clearly stated any findings - whether recommendations,
concerns, or conditions ~ adverse to quality.
The inspector. selected the names of two of the licensee's lead
auditors for personnel qualification review. They were lead auditors ;
on two of five audits then in progress at Sequoyah~ . Their audit
areas were: )
-
Inspection Audit of Quality Control
-
Experience Review Audit
This review compared the individual's certification to the 'following
documents:
-
QM1 317 Rev. 2 dated June 14, 1988 " Auditor Training and
Certification For NQA & EB Personnel"
s
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-
QMI . 329 Rev. O, dated June 12, 1987 (Addendum 1, dated
January 19,'1988; Addendum. 2, dated February 12, 1988) titled-
" Periodic Evaluation of Auditor Performance."
The review of the documents completed in accordance with QMI 317,
,
Attachments 1 and 2, determined that both individuals met the
required elements in the areas of education, experience, professional
accomplishment, and prior management experience. They had attended
the- required training courses and had participated in an adequate
number of previous audits. The records reviewed indicated they were
well qualified as QA lead auditors.
The CMAP audit discussed below also concluded that auditors were
properly trained and certified. In a CAQR that identified five
deficiencies associated with the NDE Level III qualification for one
of the lead auditors, the deficiencies were administrative in nat'ure
and did not affect the actual qualification of the lead auditor.
During June .13 - 24,1988, a CMAP audit was conducted of the DNQA.
This special audit of Segouyah quality-assurance-related activities
was conducted by a team of experienced personnel from three outside
nuclear utilities to satisfy commitments made in TVA Topical Report
TVA-TR75-1A, Sections 17.1.2.5 and 17.2.2.5.
The CMAP audit team concluded that both the quality audit and quality
monitoring programs were effective, that both programs were conducted
to the appropriate depth to identify program and hardware problems in
a timely manner and that problems were clearly and accurately
described on CAQRs.
This audit resulted in several recommendations associated with more
efficient utilization of personnel and planning / scheduling of quality
audit and monitoring activities. The NRC inspector noted that these
recommendations had been adequately addressed by the licensee with
some minor program changes.
,
Two CAQRs were identified during the CMAP audit:
-
CAQR CHS 880047 identified five deficiencies associated with the
NDE Level III qualification for one of the lead auditors. The
deficiencies were administrative in nature and did not affect
the actual qualification of the lead auditor.
l l
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CAQR SQA 880432 identified the failure to adhere to procedures l
'
,
to adequately document the COTS and provide the required
information for trending.
l
b. Quality Monitoring Organization
i
The licensee's onsite Quality Monitoring Organization was recently
'
created to replace the older Quality Surveillance Organization.
1
.
1
_ _ _ _ _ _ - _ _ _ - _ _ _ _ _
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r.: 4- ,
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p 49 k Quality Surveillance and Monitoring activities were verification R techniques intended by the' licensee to assist the site management in- L meeting quality ' objectives by providing continuing evaluation of
performance and identifying conditions adverse to quality before they
impact nuclear. safety . reliability or component' operability. Unlike
.
Quality Audit activities which .are generally. programmatic .in. nature,
y are generally ; part of a structured program designed to satisfy l regulatory requirements, and usually consist.primarily of the review m of documents or records, the QM group featured inspections on a real
. time basis - including observation of activities'in progress.
u The Quality Monitoring Program was designed to provide for a " quick. l look" by an experienced QE within each of the disipline areas. This
- was accomplished by a group of 12.QEs and 3 group supervisors. Each
QE' was expected to complete an average of 8 quality monitoring
reports ~per month for a group total of approximately 1,000 monitoring-
reports per year. The group was still able to perform surveillance
on . a - case-by-case basi s , however, these were, more detailed and
required a' greater. amount of. time than monitoring reports.
L An NRC inspector reviewed the. organizational structure of the QM
' -
Group along with selected resumes for QEs and group supervisors.
Interviews'were conducted with several QEs and:two group supervisors.
Additionally, the licensee provided a matrix which outlined each
individual's background. Although the group was a relatively new
organization, it contained a well qualified blend of personnel with a
good mix of education, training, and work related experience within 3
-the various technical. discipline areas. i
One of the recommendations of the CMAP audit discussed above was for
improvement of the monitoring schedule. It was viewed by the CMAp
team as too structured and not - allowing sufficient time for
independent inspection effort by the QEs. In response, the licensee
modified their program to allow approximately 50% of a QE's time to
remain free which provided sufficient flexibility to perform special
and immediate monitoring while assuring specific areas were being
monitored. To accomplish this, QEs and group supervisors were
required to attend routine plant meetings such as the operations
nift turnover briefings and daily plant status meetings, and to make
frequent plant tours to maintain cognizance of ongoing plant status.
They were then expected to schedule meaningful "real time" activities
for monitoring.
NRC inspectors accompanied QEs during portions of selected ongoing
monitoring activities. In each case the individual attributes that
were scheduled to be checked were verified and denoted as such on the
checklist used by the QE. The NRC inspector then reviewed several of
the completed monitoring reports and discussed the results with the
respective QE. The results section of each QM report was found to be
complete, including subjective comments concerning the area
- _ _ _ - - _ - _ - _ _ _ _ _ - _ . ___ __. _- -- -_- - _ _ _ - _ - _ - _ _ _ - _ _ _ __ -____- _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - - - _ - _ - _ _ -
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monitored. The QE also orally communicated the results to a licensee
management representative for the respective discipline.
The NRC inspector determined that the QM Group was an experienced and
motivated group of capable individuals that should contribute.to the
management effectiveness at SQN.
12. Independent. Qualified Reviewer Process
The IQR process was a new technical review and control process implemented
prior to-Unit 2 startup that supported the review and approval of changes
to procedures and changes or modifications to safety-related structures,
systems', or components, as -allowed by TS 6.5.1A. The intent of the new
process was to reduce the administrative burden on the PORC and allow more
available time for PORC _ review of significant issues essential to the
-
operation of the plant,'thereby improving PORC effectiveness.
'
IQRs were required to perform a detailed technical review of procedures,
changes to procedures and' safety evaluations for the following:
-
Procedures required'by TS 6.8.1
-
0ther procedures which affect nuclear safety
- P1' ant modifications to safety-related structures, systems, or
components
These ' reviews must be performed by reviewers trained within their
respective area of responsibility or expertise. Reviews were to be
performed in accordance. with AI-43, Independent Qualified Review, by
personnel- designated on Appendix A of AI-43. The reviewer could not be
the preparer of the procedure change or the plant modification. Reviews
must include a USQD screening review, i.e., a determination of whether or
not an unreviewed safety -question per 10 CFR 50.59 was involved.
Screening reviews would be performed in accordance with SQEP-128,
10 CFR 50.59 Qualified Safety Evaluations. Additionally, each review must
include a determination of whether or not a cross-disciplinary review is
necessary.
The NRC inspector reviewed completed Quality Audit Report SQA 88-804,
which was performed January 19 -
February 18, 1988, concerning SQN
programs for control of the USQC and procedure change processes. The
audit concluded that both programs, including the IQR process, were
effective in achieving the desired results. Although, no CAQRs directly
relating to the IQR process were identified, several concerns and
recommendations related to the other processes were identified. The
licensee had generally addressed each of these concerns by redsion of
an associated procedure or' otherwi se implementing the recommended
improvement. Subsequent NRC Notice of Violation 327, 328/88-43-01
(September 1988) addressed the 10 CFR 50.59 review process, both from
the TVA generic and SQN-specific aspects.
i
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51
Various administrative, surveillance and operating procedures were
selected for review by the NRC inspector to verify the adequacy of the
licensee's determination of cross-disciphnary-review necessity. In all
cases where the NRC inspector believed that a review by another section
might be required, the associated procedure routing page reflected that a
cross-disciplinary review had been performed by the appropriate section.
In Audit Report SQA 88-804, a concern was identified about the failure of
TS 6.5.1A to require TACFs to have an IQR review. In reply to the audit
concern, the licensee stated that all TACFs on CSSC components receive a
safety evaluation, screening review for USQ, and Plant Manager /PORC review
and approval prior to implementation. Since PORC review and approval in
this manner satisfies the intent of TS 6.5.1A, the licensee has adequately
addressed the audit concern.
Per AI-43, to qualify as reviewers, individuals must meet the following
experience requirements:
-
Possess an academic degree in engineering or related field and have
3 years nuclear related experience or
-
Hold an active Reactor Operator (RO) or Senior Reactor Operator
(SRO) License
Individuals who do not possess the formal education or license specified
above are evaluated for consideration on a case-by-case basis.
Additionally, each reviewer must undergo four hours of formal training
covering the duties of Independent Qualified Reviewers and eight hours of
formal training covering USQD evaluations. The NRC inspector selected for
review eleven Independent Qualified Reviewers from the approved list. The
qualifications for those reviewers were then checked to determine the
quality of the selection process. The licensee documents the management
review of the individual's education, experience and training on
Attachment 3 to AI-43. For those Attachment 3s reviewed, the level and
type of experience appeared to be adequate and consistent with the area of
discipline that the reviewer was assigned. Additionally, any personnel
that required special case-by-case approval due to lack of a degree or
license had a reasonable level of other experience to justify their use as
a reviewer.
The licensee's IQR process, as implemented, appeared to provide indepen-
dent technical and screening reviews for those procedures, changes or
plant modifications specified above. Additionally, this program also
reduces the administrative burden on the PORC.
13. Exit Interview (30703)
l The inspection scope and findings were summarized on January 26, 1989,
with those persons indicated in paragraph 1. During the inspection
period, frequent discussions were held with the Site Director, Plant
Manager and other managers concerning inspection findings. The inspectors
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described thel areas inspected and discussed in detail the inspection
findings listed below. The licensee' acknowledged the inspection findings
and did not' identify as proprietary any of the- material reviewed by the
l- inspectors during the inspection.
. Item Number Description and Reference
[ 327, 328/88-50-01- Violation - Failure to_Take' Adequate '
Corrective-Action'en Previous Violation 327,
328/87-30-01, paragraph 5.
327, 328/88-50-02 URI - Trending Within ACPs and the
, Appropriate Thresholds for Entering the CAQR
Process, paragraph 6.
327, 328/88-50-03 ' LIV - Uncited level V Violation for fail'ure
to Follow Maintenance Procedures, Two-
Examples, paragraph 7.
URI'- Inclusion of Vendor Manual Torque
'
;327, 328/88-50 04
Requirements in Maintenance Instructions,
paragraph 7.
~
327, 328/88-50-05 URI - Completion of Workplan Review and
Reporting of ASME Code Section XI Repair and
Replacement, paragraph 10.
,
327, 328/88-50-06 LIV - Uncited Level V Violation for
Failure to Follow Welding Procedures,
paragraph 10.
327, 328/88-50-07 URI - Engineering Evaluations of Vendur
Manuals, paragraph 5.
'
14. List ~of Abbreviations
ABGTS- Auxiliary Building Gas Treatment System
ABSCE- Auxiliary Building Secondary Containment Enclosure
ACP - Administrative Control Program
AFW -
Auxiliary Feedwater (system)
AI -
Administrative Instruction
ANII - Authorized Nuclear Inservice Inspector
Abnormal Operating Instruction
'
A01 -
ASME Code-American Society of Mechanical Engineers Boiler and Pressure
Vessel Code
AVO -
Auxiliary Unit Operator
ASOS - Assistant Shift Operating Supervisor
BFN - Browns Ferry Nuclear Plant
BIT' - Boron Injection Tank
C&A -
Control and Auxiliary Buildings
___ _ __-___--_- _
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c. l [ ,
*
CAQR -
CCP -
Condition Adverse to Quality. Report
' Centrifugal Charging Pump
b. CCTS - Corporate Coinmitment Tracking System
CCW HX- Component Cooling _ Water Heat Exchanger
CMAP -- Cooperative Management Audit Program
COPS - Cold Overpressure Protection System
COTS - ' Correct on the Spot
'CSSC .
Critical Structures', Systems and Components
CVI - Containment Ventilation Isolation
DC . -' Design Change
DCN - Design Change Notice-
DCR
'
Design Change. Request
DNE - Division. of Nuclear Engineering
DNQA - ' Division of Nuclear Quality Assurance
DWL ' - . Daily Work List-
EA .-~ Engineering Assurance
.ECCS -
; Emergency Core Cooling System-
ECP - Employee Concern ~ Program
~EDG -
-Emergency Diesel Generator
EI -
Emergency Instructions-
ENS -: Emergency Notification System
EQIS - Equipment Qualification Information System
ERCW - Essentia1' Raw Cooling Water
ESF. -
Engineered Safety Feature
FCV - . Flow Control Valve
FI -
Flow Indicator
FSAR - Final Safety Analysis Report
GDC - General Design Criteria
GL '-
Generic Letter
HCV -
Hydraulic Control Valve
HIC - Hand-operated Indicating Controller
H0 ' - Hold Order
HP '- Health Physics
i to p- Current to. Pneumatic (converter)
ICF - Instruction Change Form
IFI - Inspector. Followup Item
IM -
Instrument Maintenance
IMI -
Instrument Maintenance Instruction
IN -
NRC Information Notice
INPO - Institute of Nuclear Power Operations
IQR
-
Independent Qualified Reviewer
'
IR -
Inspection Report
ISEG - Independent Safety Engineering Group
KVA -
Kilovolt-Amp
KW -
Kilowatt j
KV
-
Kilovolt ]
LE -
Level-Element
LER -
Licensee Event Report #
LC0 -
Limiting Condition for Operation
LOCA - Loss of Coolant Accident
i
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LT -
Level Transmitter
MI -
' Maintenance Instruction
-M&TE - Measuring:and Test Equipment.-
~MOVATS- . Motor-Operated Valve Actuator' Test System
MCR -
. Main Control Room -
MRC - Management Review Co'mmittee
MTI - Mechanical Test Instruction
.N2 -
-Nitrogen Gas
.N/A: -
Not Applicable
' NDE - -Non-Destructive Examination
>
'NOV -
Notice of Violation
NPRDS - Nuclear. Plant _ Reliability Data System
NQA - - Nuclear. Quality' Assurance
NQAM'- Nuclear Quality Assurance Manual
NRC. - Nuclear Regulatory. Commission
-ONP - Office of Nuclear Power.(TVA)
'0SLA-- -Operations.Section Letter Administrative
OSLT - Operations Section Letter..- Training ,
OSP_ - Office _of Special Projects' i
PDWL - Priority Daily Work List
PIR - Problem Identification Report . _
Preventive' Maintenance (action, instruction, etc.)
'
,PM -
PMT. - Post Modification Test
PNL. - Panel-
POD '_- Plan of the Day
PORC - Plant Operations Review. Committee
.PORS - Plant Operation Review Staff
POTC - Power. Operations Training Center
PRD -
-Problem Reporting Document
PRO -
Potentially Reportable Occurrence
PT --
Pressure Transmitter
'QA -
Quality Assurance
QC
-
-Quality. Control
.QE
-
Quality Evaluator
QM
-
Quality Monitor (ing)
QV-
-
Quality Verification
.QVFI - -Quality Verification Function Inspection
,
RCS -
Reactor Coolant System
'
Rem - Roentgen-Equivalent-Man
Rev. - Revision
RG .
Regulatory Guide
? RM -
Radiation Monitor
l' RHR - Residual Heat Removal
-RTD - Resistance Temperature Detector
^
RWP - Radiation Work Permit
RWST - Reactor Water Storage Tank
SCR - Significant Condition Report
SE -
System Evaluator.
SER -
Safety Evaluation Report
.SG -
Steam Generator
SI -
Surveillance Instruction i
..
-
-
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Op ~ [. '.
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. t
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,
- ;.
. , ESIMS - Safety,Information Management System :
1 : SIS -
1 Safety Injection-System -
; SOI -
System'0perating Instructions )
-SOS: =
. Shift Operating Supervisor 1
SQM -- Sequoyah Standard Practice Maintenance I
SQN' : Sequoyah Nuclear Plant.-
SR~ -
-Surveillance Requiremen;s
- SRO' - Senior Reactor Operator
'SRST - Spent. Resin Storage Tank:
STI -- .Special Test Instruction
' STORM- Shift Turnover
1 Restart Meetingn(report)
~
TACF - ' Temporary Alteration Control Form
.
-TMI - Three Mile. Island. Nuclear = Plant y
TROI - . Tracking .of Open > Items (system) !
,. 'TS: -
Technical: Specifications
L .TSCR -- -Technical ~ Specification. Condition Report
' Tennessee Valley Authority
'
TVA .
,
'
UHI -
, Upper Head Injection (system).
URI -
Unresolved' Item
3 ' 00; -
Unit,0perator_'_ _
.
USQD - 'Unreviewed Safety Question Determination
.VIO ~- . Violation'(of NRC requirements) i'
VLV - Valve
--Work' Control Group
'
-WCG: -
WP -
: Work Plan
' Work Request
'
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:WR
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56 f
APPENDIX A
Administrative Control Program Items Reviewed
1. The inspector reviewed the following documents from the below-listed ACPs
to determine that appropriate licensee document reviews had been made to
ensure that CAQs were not present:
-
Work Request (WR) - SQM-2, Maintenance Management System
"
-
B 753029 - Main Steam Safety Valve Leaks Thru.
-
0 283836 - SIS Accumulator Tank 2 Pressure Indicator
Failed Low.
-
B 769983 --RHR Pump 2A-A Discharge Line Flow Transmitter
Low Side Fitting Is Gaulded.
-
B 769234 - Negative Rate Trip Light Will Not Clear.
l
B 283669 - Lower Radiation Monitor Is Not Indicating That
'
-
It Has Power.
-
B 283663 - Auxiliary Feedwater Pump 2A-A, Add Oil To
Inboard Bearing.
-
Potential Reportable Occurrence (PRO) - SQA-84, Potential
Reportable Occurrences.
-
1-88-263 - Waste Gas Analyzer declared inoperable.
-
1-88-274 - Seismic Recorder 0-XR-52-7SB was found
inoperable when investigating WR B753263. j
-
1-88-293 - Entered LC0 for Unit I when condenser vacuum
exhaust flow rate was declared inoperable.
-
1-88-307 - Snubber clamp for 1-FLV-70-87 was not in place
when mode 4 was entered.
-
1-88-320 - Fire Alarm was initiated due to fire on the
Unit 1 turbine head end.
-
Drawing Discrepancies - AI25 Part II, Drawing Deviations.
-
88DD4114 - Revise drawing number 47W 0810-1
-
88DD4128 - Revise drawing number 47W 0805-2
-
s,< , .
'
,' N
z,;. .
57:
-
' 88DD4164;-_ Revise drawing number 47W'0865-1
~
-
- 88004171.- Revise drawing ' umber n 47W.0834-2
?.
- : 88DD4173 -LRevise drawing number.47W 0865-2-
-
Radiological Incident Report - RCI_1, Radiological Control
. Program.
.
'
- -
. RIR-88-30 - Individua1Eentered containment in violation of
'RWP.
sc -
RIR-88-31 - Employee failed to sign-in on RWP.
-
, RIR-88-32 - Worker failed to process previous RIR's within
t. 1 hour, violation of RCI-1.
-
- RIR-88-39 Worker performed personal decontamination-
without notifying Radiation Control.
'
~- - Housekeeping Deficiencies - SOA-66, Plant Housekeeping.
-
Routine monthly report November 1988. Nine items in waste
packaging railroad bay.
-
Routine' monthly report September 1988. Two items in-
additional. equipment building.
> -
Routine monthly report November 1988. Seven items in the
Auxiliary Building in elevation 653.
-
Test Deficiencies - AI47, Conduct of Testing.
-
' SI-3, Oly, DN-1 - RM-90-119 failed to pass SI-302, WR
B769333 written.
-
SI-167, DN-4 - Valve 0-26-1755 Tagged Closed, H0
1-88-1221 lifted.
-
SI-531, DN-4 - 0-VLV-26-895, replace stem per WR
B296912
-
SI-540, DN-1 - Extinguisher removed when building taken
,
down.
-
SI-2, DN-2 - XA-55-60-2 will not clear, WR B283717
written
-
- Problem Reporting Document - A-12 Part III, Corrective Action,
Appendix I, Processing of Problem Reporting Document.
-..
-_n--,--------n-x---- - , - - - - - . ------_--------_---.-_--_-_----------.--___--------------_------.ww-----_.---------------------------,---------------------_-.---___--------_-----_---,-----,------__------a
- _ - . _ . _ - _ _ _
- 3.;.
E d, L 58
-
SQP 880007P - 56 of.72 WR's already closed in last 6
months did not include a failure cause.
I -
SQP 880027P - 6 uses of AI-47 deficiency log in
WP5591-01-1 outside interpretation of AI-47.
-
SQP 880034P - Worst. case vertical drop conduit walkdown
sketch was found in error.
-
SQP 880041P - The 1987 full scale-general fire drill has
not been entered into plant QA record
system.
-
SQP 880077P - The Discriminator voltage for channel N31
was charged on a WR and the WR closed prior
to resetting the channel.
-
. NQA Audit Report (correct-on-the-spot) - QMI-328,
-
SQA 88816, COTS 1 - Employee's confidentiality was not
tagged. (on the employee concern
report file)
-
SQA 88816, COTS 2 - ECP only closed 1 of 2 concerns -
expressed in exit interview.
-
SQA 88804, COTS 1 - SDR 88-036-02 was not responded to.
-
SQA 88804, COTS 2 - No EA review of SQP 871677.
- -
SQA 88812, COTS 1 - An M&TE investigation did not include
all activities for which the equipment
was used.
-
SQA 88812, COTS 3 - Procedure revision request QSB-384 was i
issued to correct incorrect reference )
in NQAM Part II section 2.4. j
i
-
QA Surveillance Reports (Correct-on-the-Spot) -AI-32, Quality J
Assurance Surveillance, QMI-702.6,
-
QSQ-M-89-015, COTS-1 - A copy of the workplan was not at
the site.
]
1
-
QSQ-M-87-048, COTS-1 - A copy of the implementing
procedure was not at the worksite.
1
-
QSQ-M-88-986, COTS-1 - The M&TE utilized in SI-93 was l
'
logged, however, an improper
SI-90.3 was listed.
1
l
______m__ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ ___________._____.______________________o
.. .,- -
,
g.(;; y
p n'.. j
,
c 59
,
'
-
QSQ-M-87-174, COTS-1 - The MI-6.20 utilized in-WR'B275782
was not verified to be the latesti
revision.
-
QSQ-M-88-128; COTS-1 - Penetration 774 ABA 12018000 was'
missing.a metal tag.
-
.QC-Inspection Rejections-'(Inspection Reports'and
. Correct-on-the-Sp>t) - AI-20,-QC InspectionLProgram, AI-11,
, Receipt Inspection and Non conforming Items.
--~ IR WBB-2489 . magnetic particle inspection - COTS on
cleanliness.
.
* -
-IRW88-2505 - Fixup' inspection COTS on' documentation.
- '
IRM88-0326 - Quality control inspection COTS on
documentation.
-
IRE 88-0977 - Quality control of gasket COTS to replace.
-
Licensee Event Reports - AI-18, Plant. Reporting Requirements.
-
327/88-039 - Unauthorized personnel.using a portable:
radio inside Unit 1 containment generated a
reactor trip signal.
-- "327/88-022 - Reactor trip signals generated from
electromagnetic interference caused by
welding machine operated near source range
nuclear instrument' cabling.
-
327/88-043 - Inadequate firewatch patrol resulted in a
noncompliance with TS 3.7.12.
-
328/88-030 - The failure to identify that the effects of
excessive post-trip reactor coolant system
cooldowns could have caused noncompliance
with shutdown margin requirements.
-
328/88-006 - Engineered safety feature main steam line
isolation and reactor trip due to maintenance
activities and inherent instrument response
during steam plant heatup.
-
Security Degradation / Incident Report - PHYSI-29, Security
Degradation / Incident Reporting.
-
87-088-10 - Deficient barrier from the protected area into
a vital area as a result of a HVAC duct
penetration.
,
,
L_-___-__-_-______-______ b
_ _ _ . - _ _ _ _ _ _ _ - _ _ _ _ _ _ _
-
4o .
.o
60
:
-
87-087-10 - Deficient Vital Area Barrier.
-
87-063-04 - Individual Piggybacked Through Vital Area
Door.
-
87-087-05 - Failed to Return Badge Within Allowable Time
After Leaving ERCW.
-
88-094-05 - Individual Accepted Wrong Badge At Access
- Portal.
-
88-118-06 - Failed to Return Badge Within Allowable Time
After. Leaving ERCW.
2. The June 30, 1988, Semiannual Component- Failure Trending Report was
reviewed. It identified the following components as having trends
and documented the corrective actions taken:
,
1-FCV-087-0021 UHI ISOLATION VALVE
1-VLV-002-0521 AUTOMATIC MAKEUP LCV ISOLATION VALVE
! 2-FSV-043-0287-A POST ACCIDENT SAMPLING VALVE
2-FSV-082-0231 DIESEL GENERATOR AIR START SOLEN 0ID
0-FSV-082-0160 DIESEL GENERATOR AIR START SOLEN 0ID l
1-VLV-062-0901 MIXED BED DEMIN INLET VALVE
1-PMP-002-0020 HOTWELL PUMP
1-FCV-043-0022 REACTOR COOLANT HOT LEG SAMPLE HEADER
CONTAINMENT ISOLATION VALVE
0-VLV-082-0501-1A1A DIESEL AIR START COMPRESSOR RELIEF
VALVE
0-VLV-082-0534-1A2A DIESEL AIR START AIR TANK RELIEF VALVE
2-FT-068-0006A RCS LOOP COOLANT FLOW TRANSMITTER
1-TS-030-0194 714 PENETRATION ROOM COOLING FAN
1-PDIS-067-0491E/F SWITCH
1-FT-001-0010B-E STEAM GENERATOR MAIN STEAM HEADER
FLOW TRANSMITTER
1-FS-090-0106A LOWER COMPT. AIR MON. PART. LO FLOW
SWITCH
2-FM-003-35 i to p CONVERTERS
2-PX-003-170 ISOLATED DC POWER SUPPLY
2-FR-002-0035 CONDENSER HOTWELL PUMP FLOW RECORDER
0-TR-082-5036/1 DIESEL GENERATOR WINDING RTD.
]-H2AN-043-0200 CONTAINMENT HYDROGEN ANALYZERS
1-RM-080-0106A RADIATION MONITORS
1-VLV-067-0585C-A ERCW RETURN UCV COOLER CHECK VALVE
1-PMP-002-0033 HOTWELL PUMP
0-PMP-067-452 ERCW PUMP
f 2-FCV-063-0064 SIS ACCUMULATOR TANK N2 HEADER INLET
VALVE
1-FCV-001-0182 STEAM GENERATOR BLOWDOWN CONTAINMENT
ISOLATION VALVES
i
-__-___________A
. _ _
'
Ng y .:
e
g:.,
:# L
'
61 .j
l
.;e '
I
'-
-1-LCV-003-0174- STEAM GENERATORzLEVEL CONTROL VALVE
2-FSV-043-0287-A POST ACCIDENT SAMPLING CONTAINMENT
.
AIR PNL ISOLATION VALVE
"
< - 1-FCV-003-0191 FEEDWATER LONG CYCLE VALVES AND HEATER
, . . ISOLATION VALVES-
2-ENG-082-0002B1 DIESEL GENERATOR ENGINES
0-VLV-082'0501-1A1A ' COMPRESSOR-RELIEF VALVE
- .. 1-FCV-062-0086 CHARGING FLOW REACTOR COOLANT SYSTEM-
. COLD LEG LOOP FLOW CONTROL VALVE
- 0-VLV-082-05161A1A' - TANK CONNECTION SHUT 0FF VALVE
-
1-PMP-082-0001B/2-B' ~ DIESEL GENERATOR LUBE-0IL CIRCULATING
PUMP
1-PCV-001-0012-B STEAM GENERATOR MAIN STEAM HEADERL
PRESSURE RELIEF CONTROL VALVE
,
0-VLV-082-0528-1828L COMPRESSOR RELIEF VALVE ,
s !- .
__ ______-_.__m____. _____.___m__ _ _ _ . _ . _ _ . _ _ _. _ _ _ -
}}