ML20246B651
| ML20246B651 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 06/26/1989 |
| From: | Brady J, Jenison K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20246B630 | List: |
| References | |
| 50-327-89-15, 50-328-89-15, NUDOCS 8907100046 | |
| Download: ML20246B651 (29) | |
See also: IR 05000327/1989015
Text
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Re' port Nos.:
50-327/89-151 50-328/89-15"
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Licensee: " Ten'nessee Valley Authority
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6N.38A Lookout Place
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1101 Market Square
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Chattanooga, TN. 37402-2801'
Docket'Nos.:
50-327'and 50-328-
License Nos.:
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Facility Name:
Sequoyah Units'I and 2
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LInspection_-Conducted:
May 6, 1989 thru June-5, 1989
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Inspector:
by
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K. Jepfrson, Senior Resid,ent Inspector
Date Signed
Accompanying, Personnel:
P. Harmon, Senior. Resident Inspector
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D. Loveless, Resident Inspector
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Approved by: [ [ [ M
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9. Brady, Actpfg Chief, Project Section 1
Date Signed
TVA Projects Division
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SUMMARY
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Scope:
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lhis routine monthly. inspection invo'lved inspection effort by the. Resident-
Inspectors in the area of operational safety verification including control
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room observations, operations performance, system lineups, radiation protec-
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tic.n, safeguards, and housekeeping inspections.
Other areas finspected included
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twaintenan'ce observations, surveillance testing observations, review of previous
inspection ~ findings,- follow-ep of events, review of licensee identified items,
and review of inspector follow-up items.
.Results:
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- The areas lof Operations, Maintenance, and Surveillance were adequate and fully:
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capable to support current plant operations. The observed _ activities'of theJ
control rsom operators were professional and well executed., <Significant
w'eaknesses were identified in the licensee's implementation of the 10 CFR 50.59
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safety" evaluation processes, the licensee's TS interpretation process, and the
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- independent qualified review process.
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The licensee had failed to follow procedures 'as .they relate to the configura-
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tion' of a Radiation . Monitor with respect to the detector cable.
This is
identified as VIO 327,328/89-15-02 (paragraph 4).
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The licensee had failed to establish, implement, and maintain written proce-
dures for controlling thermal power within license limits at steady state
operations.
This is identified as VIO 327,328/89-15-03 (paragraph 7).
The licensee had on three occasions failed to perform adequate reviews pursuant
to 10 CFR 50.59.
These examples, described below, were identified as VIO
50-327,328/89-15-04.
a.
Changing the surveillance instruction for the RHR system , without
performing an adequate review pursuant to requirements of 10 CFR 50.59
(paragraph 9.a).
b.
Changing the mode of operation of the BIT from continuous recirculation to
occasional recirculation without performing a review pursuant to the
requirements of 10 CFR 50.59 (paragraph 9.b).
c.
Changing the positions of the source range and intermediate range detector
locations which affected the intermediate range high neutron flux trip
without performing an adequate review pursuant to the requirements of
'10 CFR 50.59 (paragraph 9.c).
The licensee had on three occasions described below failed to establish,
implement and maintain adequate written procedures.
These examples were
identified as VIO 50-327.328/89-15-05.
a.
The licensee allowed a procedure that was known to be inadequate for RHR
pump and pipin0 venting to be performed (paragraph 9.a).
b.
The licensee failed to have an adequate procedure before changing the mode
of operation and the associated system lineup of the BIT (paragraph 9.b).
c.
The licensee failed to ensure that a safety evaluation accompanied the
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screening review form for tne IR and SR detector relocation as required by
SQA-119 (paragraph 9.c).
The licensee put two safety systems in an inoperable condition and failed to
comply with TS LCO action statements.
These examples, described below, were
identified as VIO 50-327,328/89-15-06.
a.
The BIT vas made inoperable during the period when recirculation flow was
stopped (paragraph 9.b).
b.
The IR high flux bistables were made inoperable when IR detectors were
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withdrawn to a new location (paragraph 9.c).
One non-cited violation was identified:
Violation of RWP procedure resulting in the contamination of two workers
(paragraph 2.e).
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One unresolved ' item was identified:
Apparent weakness in the safety evaluation program (paragraph 9).
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REPORT DETAILS
1.
Persons Contacted
Licensee Employees
J. Bynum, Vice President, riur. lear Power Production
- J. LaPoint, Site Directar
- S. Smith, Plant Manager
T. Arney, Quality Assurance Manager
- R. Beecken, Maintenance Superintendent
- M. Cooper, Compliance Licensing Manager
D. Craven, Plant Support Superintendent
- S. Crowe, Site Quality Manager
- R. Fortenberry, Technical Support Supervicor
J. Holland, Corrective Action Program Manager
J. Patrick, Operations Superintendent
R. Pierce, Mechanical Maintenance Supervisor
- M. Burzynski, Site Licensing Staff Manager
A. Ritter, Engineering Assurance Engineer
- R. Rogers, Plant Support Superintendent
- M. Sullivan, Radiological Controls Superintendent
- S. Spencer, Licensing Engineer
C. Whittemore, Licensing Engineer
- E. Whitaker, ISEG Manager
- G. Gault, Reactor Engineering Supervisor
- H. Hellums, Licensing Manager
- T. Flippo, QA Manager
- S. Crowe, Site Quality Manager
- C. Mason, 0&M Support Manager
- K. Yeller, PORC Oversight Supervisor
- F. kashburn, ISEG Manager
- V Blanco, Lead Nuclear Enginee*
- L. Cooco, Attorney
NRC Employees
- J. Brady, Acting Sectien Chie'
- Attended exit interview June 5, 1989
- Attended exit interview on June 9, 1989
Acronyms and initialisms used in this report are listed in the last
paragraph.
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2.
Operational Safety Verification (71707)
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a.
Control Room Observations
The inspectors conducted discussions with control room operators and
verified that proper control room staffing was maintained.
The
inspector also verified that access to the control room was properly
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controlled, and that operator behavior was commensurate with plant
configuration end plant activities in progress and with on going
control room operations.
The operators were observed adhering to
appropriate and approved procedures.
Additionally, the frequency of
visits to the control room by upper menagement was observed for
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adequacy and found to be acceptable.
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The inspector also verified that the licensee was operating the plant
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in a normal plant configuistion as required by TS.
When abnormal
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conditions existed, operatore were found to be . complying with the
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appropriate LC0 action statements.
The inspector verified that leak
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rate calculations were performd and that leakage rates were within
the TS limits,
The inspectors sampled instrumentation and recorder traces for
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indication of abnormalities and verified the status of selected
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control room annunciators to ensure that control room operators
understood the status of the plant.
Panel indications were reviewed
for the nuclear instruments, emergency power sources, the safety
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parameter display system, and the radiation monitors to ensure
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operability and operation within TS limits.
With the exception of
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VIO 327,323/89-15-03 and VIO 327, 328/89-15-04, example 3, operator
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actions with respect to control room indications were adequate,
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Control rod insertion limits were observed as specified in the TS,
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t!o viclations or deviations were observed.
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b.
Control Room Logs
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The inspectors observed control room operations and reviewed
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applicable logs including the shift logs, operating orders, night
order book, clearance hold order book, and configuration log to
obtain information concerning operating trends and activities.
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TACF log was reviewed to verify that the use of jumpers and lifted
leads causing equipment to be inoperable was clearly noted and
understood.
The licensee is actively pursuing corrections to
conditions requiring TACFs.
No issues were identified with these
specific logs.
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Plant chemistry reports were reviewed to confirm steam generator tube
integrity in the secondary side and to verify that primary plant
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chemistry was within TS limits.
A specific problem with dose
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equivalent Iodine occurred on Unit I during an approximately eight
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day period within this inspection period.
Dose equivalent Iodine
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increased by a factor of approximately five with no identifiable
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cause.
This condition was recorded, tracked, and trended by the
plant Chemistry section, but a concerted effort to identify the root
cause ' was not exerted until after the involvement of the Plant
Manager.
The resident staff will continue to monitor the licensee's
resolution of this problem.
In addition, the implementation of the licensee's sampling program
was observed.
Plant specific monitoring systems including seismic,
meteorological, and fire detection indications were reviewed for
operability.
A review of surveillance records and tagout logs was
performed to confirm the operability of the reactor protection
system.
No violations or deviations were observed.
c.
ECCS System Alignment
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The inspectors walked down accessible portions of the following
safety-related systems on Units 1 and 2 to verify operability, flow
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path, heat sink, water supply, power supply, and proper valve and
breaker alignment:
Component Cooling Water System (Unit 1)
Secondary Radiation Monitors (Units 1 and 2)
Feedwater System (Unit 2)
RHR System (Unit 1)
In I,ddition, the inspectors terified that a selected portion of the
containment isolation lineup was correct,
ho deviations or violations were identified.
d.
Plant Tours
Tours of the diesel , generator, auxiliary, conrol, turbine buildings,
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and exterior areas were conducted to observe plant equipmec
conditions, potential fire hazards, control of ignition sources,
fluid leaks, excessive vibrations, missile hazards and plant
houseke. iing and cleanliness conditions.
The plant was observed to
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be clean and in adequate condition.
The inspectors verified that
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maintenanc.e work orders had been submitted as required and that
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followup activities and prioritization of work was accomplished by
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the licensee.
The prioritization of work was discussed with the
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Maintenance Superintendent and Operations Management with respect to
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the increase of maintenance activities that are outstanding in the
control room.
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The inspector visually inspected major components for leakage, proper
lubrication, cooling water supply, and any general condition that
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could prevent fulfilling its functional requirements. The number of
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minor housekeeping and maintenance items such as small leaks,
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tagging, chipped paint, etc. is increasing.
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The inspector observed shift turnovers and determined that all
necessary information concerning the status of plant systems was
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addressed.
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No violations or deviations were observed.
e.
Radiation Protection
The inspectors observed HP practices and verified the implementation
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of radiation protection controls.
On a regular basis, RWPs were
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reviewed and specific work activities were monitored to ensure that
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activities were being conducted in accordance with the applicable
RWPs.
Workers were observed for proper frisking upon exiting
contaminated areas and the radiologically controlled area.
Selected
radiation protection instruments were verified operable and
calibration frequencies were reviewed. The following RWP was reviewed
in detail:
RWP 89-00012, CVCS Hold-up Tank Room B.
The activities that took place under this RWP involved the
contamination of two workers.
This particular RWP prohibited
entry into areas with contamination levels greater tha: 50,000
c'pm/100 square centimeters.
Contamination in the general area
where the two workers were performing activities was approxi-
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mately 9 mrad / hour /100 square centimeters.
When the two subject
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workers were discovered by a TVA Health Physics representative,
they were directed to leave the area.
Each worker was showered
several times and given a whole body count. One worker indicated
a 0.1% Maxiraum Permissible Organ Burden, Co-60.
Two Radiolo-
gical Incident Reports (RIR) were written by the licensee
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(89-67, 89-68).
No legal limits were exceeded and nb over-
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exposures took place.
Because the licensee identified this
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violation of RWP procedures, took prompt corrective actions, and
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is evaluating long term corrective actions under the referenced
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RIR, this issue will be categorized as Non-cited Violation
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327,328/89-15-01.
This violation is not being cited because the
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criteria specified in Section V.G. of the Enforcement Policy
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were satisfied.
NCV 327, 328/89-15-01 is closed.
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Safeguards Inspection
In the course of .the. monthly activities, the inspectors incl'uded a
review of the licensee's physical security program.
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of.various shifts of the 4 security force was observed in the. conduct
of daily . activities including: protected and vital area access
- controls; searching of personnel and packages; escorting of. visitors;
badge issuance and, retrieval;.and, patrols and compensatory posts.
In . addition, - the. inspectors observed protected area lighting, and
protected and vital area barrier integrity.
The inspectors verified
interfaces between the security organization and both operations and
' maintenance.
Specifically, the Resident Inspectors:
(1) witnessed firearms training and qualification
(2). interviewed individuals with security concerns
(3) ' visited central and secondary alarm stations
(4) verified protection of_ Safeguards Information
(5) . verified onsite/offsite communication capabilities
No violations or deviations were identified,
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The inspectors reviewed selected items to determine that the
licensee's problem identification system as defined in AI-12,
Corrective Action, was functioning.
CAQR's were routinely reviewed
for adequacy in addressing a problem. or event.
Additionally, a
sample of the following documents were reviewed for adequate handing:
.(1) Work Requests
(2) Potential Reportable Occurrences'
(3) ' Radiological Incident Reports
(4) Problem Reporting Documents
(5) Ccrrect-on-the-Spot Docuraeats
(6) Licensee Event Reports
Of the items reviewed, each was founc' to have been ident.ified by the
licensee with immediate corrective action in place.
For those issues
that required long term corrective action, the licensee was making
adequ6te progress.
No violations or deviations were observed.
Positive trends were identified in the operational safety verification
area.
Operations management changes were put in place to strengthen the
Operations Superintendent position. Personnel in the positions that
directly support the Operations Superintendent position appeared to be
taking a more active role in the operation of the units.
General
conditions in the plant were adequate.
The number of control room
maintenance and modification items increased during this inspection period
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as a result of an increased effort in the Operations Section to remove
unnecessary indications and difficulties for the operators.
Radiation
protection and security are adequate to continue two unit operations.
3.
Surveillance Observations and Review (61726)
Licensee activities were directly observed / reviewed to ascertain that
surveillance of safety-related systems and components was being conducted
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in accordance with TS requirements.
The inspectors verified that: testing was performed in accordance with
adequate , procedures; test instrumentation was calibrated; LCOs were met;
test results met acceptance criteria requirements and were reviewed by
personnel other than the individual directing the test; deficiencies were
identified, as appropriate, and any deficiencies identified during the
testing were properly reviewed and resolved by management personnel; and
system restoration was adequate.
For completed tests, the inspector
verified that testing frequencies were met and tests were performed by
qualified individuals.
The following activities were observed / reviewed with no deficiencies
identified except as noted:
SI-276, Auxiliary Feedwater Control Valves Operability.
SI-137.2, Reactor Coolant System Water Inventory.
SI-38, Shutdown Margin.
No trends were identified in the area of surveillance performance during
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this inspection period.
The area of surveillance scheduling and manage-
ment was observed to be adequate and improving.
The management of the TS
SI prcgram socears to have progressed from a reactive type process to a
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reutirely scheduled, adequately managed plant operation support activity.
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This is a significant improvement in the licensee's ability to control
plant activity schedules.
TS surveillance requirements were discussed at
the highest levels of the TVA onsite Nuclear Power organization.
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4.
Monthly Maintenance Observations and Review (62703)
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Station maintenance activities on safety-related systems and components
were observed / reviewed to ascertain that they were conducted in accordance
with approved procedures, regulatory guides, industry codes and standards,
and in conformance with TS.
The following items were considered during this review:
LCOs were met
while components or systems were removed from service; redundant
components were operable; approvals were obtained prior to initiating the
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work; activities were accomplished using approved procedures and were
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inspected as applicable; procedures used were adequate to control the
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activity; troubleshooting activities were controlled and the repair
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records accurately reflected the activities; functional testing and/or
calibrations were performed prior to returning components or systems to
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service; QC records :were maintained; activities ~ were accomplished by
qualified personnel; parts and materials used were properly certified;
. radiological controls were implemented; QC hold points were established -
where required and were observed; . fire prevention controls were
implemented; outside contractor force activities were controlled in
accordance with .the' approved QA program; and housekeeping was actively.
pursued.
The following work requests were reviewed:
This work activity involved the trouble shooting and repair of a post
accident monitor located in the turbine building.
The function of
the monitor is to measure the condenser exhaust' for potential .
activity which could result from a steam generator tube leak.
When inspected in the field, the RM detector cable was installed
(pressed in and - unsoldered) into the ' detector.
The inspector-
reviewed the configuration control sheet and determined that the
configuration control document did not agree with the arrangement =in
the plant.
IMI-134, Configuration Control of Instrument Maintenance
Activities,. states that the technician performing a work activity
shall list and initial any configuration changes in order on the
IMI-134 data sheet.
These changes include jumpers, wire lifts,
inhibits, temporary instrument settings,
unbolting flanges,
disconnecting tubing and pipe fittings, temporary connections, etc.
The technician 'is to list these items in sufficient detail to
uniquely identify each item.
Centrary to the above, the configura-
tion of the RM with respect to the detector. cable was not correctly
controlled.
This is VIO 327, 323/89-15-02 tor failure to follow
procedure.
In addition to the violatior3 cited, several other
weakness were identified.
Adequate corrective action for the
violation'will include correction of these weaknesses:
Work was completed over several days with no one person in
charge.
There was no indication of what trouble shooting activities ^iere
performed by the several shifts and crews that worked on the
components.
No post maintenance test was identified and an
adequate post maintenance test will be difficult in considera-
tion of the loss of control of trouble shooting activities.
Configuration control with respect to the processor portion of
the RM was not adequate in the cases of the CPU III 10889-01,
and I/O boards.
These boards were placed into and out of the
equipment without record or control of the work activities.
Several components were stored within the electrical cabinet
portion of the RM.
These included tools, spare connecting pins,
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and spare cables.
In addition, cigarettes were also stored
within the electrical cabinet portion of the RM.
One of the tools stored in the electrical cabinet portion of the
RM' was a Winchester crimper.
This tool belonged to the vendor
and was not controlled. or calibrated under the licensee's
crimper program.
In addition, several spare connecting pins
were identified in the cabinet .that did not have corresponding
-documentation in the work package.
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WR 8265544, Containment High Temperature Alarm.
During post maintenance.-testing of this repair the operator noted
that the annunciator test' circuit did not function like the other
panels.
This discrepancy, although partly by design, allowed for
some alarms to stay' in when they were not, in fact, in alarm.
The.
review. determined that the alarms would still perform their intended-
function.
The licensee is still trying to determine the source of
the problem.
WR B265525, RPI-13 Failure.
This work activity involved the extensive testing to the Rod Position
Indication System and the modification of a circuit connection.
The inspectors reviewed the outstanding maintenance activities which
directly affect the control room and control room activities.
The number
of outstanding maintenance activities is increasing and is very visible by
the tags and out-of-calibration stickers placed within the control room
horseshoa operating. area.
Both the Maintenance and Operations
Superintendents are aware that the number of outstanding maintenance items
is increasing. 16th' managers maintain that the large number of items is
an . indication of strong support by maintenance of the operators and an
attempt to maximize maintenance efforts.
The inspector determined that
the number and nature of outstanding maintenance items is presently not
affecting the safe operation of the plant.
One VID 327,328/89-15-02, and no deviations were identified in the area of
Maintenance.
5.
Management Activities in Support of Plant Operations
TVA management activities were reviewed on a daily basis by the NRC
inspectors.
Resident Inspectors observed that planning, scheduling, work
control and other management meetings were effective in controlling plant
activities.
First line supervisors appear to be knowledgeable and
involved in the day-to-day activities of the plant.
First line supervisor
involvement in the field has been observed and with the exception of the
RM activity described above appeared to be adequate.
Management response
to those plant activities and events that occurred during this inspection
period appeared timely and effective.
Examples of this management action
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were the responses - to the two personne1' contaminations, and a failed
RPI-E-13 on Unit 1.
One weakness appeared to.be the first and middle line
management being unable to explain the increase in Unit 1 dose equivalent
Iodine over a period of approximately eight days.
This issue is still
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under. consideration;with Plant Management involved in the resolution of.
the. problem.
The management activities which are discussed'in paragraph 9
and cited as VIO 327, 328/89-15-04 occurred during' previous inspection
periods.
6.
Site Quality Assurance Activities in Support of Operations-
During the7first part of this inspection period, the inspector discussed
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several recent issues involving the implementation of. SQEP-65, External
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Interface Control, with - the Site QA Manager.
The inspector expressed
concern that the SQEP-65 process could be used to bypass established
review programs.
.This process is used by DNE to transmit technical
information to the plant as a response to plant initiated questions.
The
site QA Manager established three audits of the SQEP-65, P-QIR' process.
The QA monitoring, documented in the reports listed below, determined that
. the P-QIRs were being - used as design' output. documents.
This allowed
calculations to be performed and utilized by the plant without a USQD as
required by 10 CFR 50.59 and without a 'PORC review.
A 100% review was
conducted and the problems that were identified were all associated with
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.non CSSC equipment.
The resolution of the problem was an agreement
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between site QA and the site Manager of DNE to discontinue'the implementa-
tion of.the SQEP-65 process and affects corrective actions through one PRD
and two CAQRs.
Monitoring Performed
QSQ-M-89-558, SQEP 65 Nuclear Sample
Q5Q-M-89-556, SOEP-55 Mechanical Sample
Q5Q-M-89-557, SQEP-65 Electrical Sample
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Corrective Actions
PRD SQQ 890299P
CAQR SQQ 890297
CAQR SQQ 890303
The licensee's corrective actions appear to be adequate..
The inspector discussed QA involvement in plant activities with the site
QA Manager.
The site QA Manager stated that he is continuing to operate
under the previously established written agreement between the previous
site QA Manager and the Plant Manager.
This agreement established QA
being notified of plant events and participation in the line investigation
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of plant issues.
7.
NRC Inspector Follow-up Items, Unresolved Items, Violations (92701, 92702)
(Closed) URI 327,328/89-12-02, BIT Recirculation.
This item is resolved
and is1 now being tracked under VI0s 327,328/89-15-04,05,06 which is
discussed in paragraph 9.b. of this report.
This item is closed.
(0 pen) URI 50-327,328/89-12-01, Licensed Power Indication.
Inspection Report 327,328/89-12 identified several questions concerning
the control of licensed thermal power.
Sequoyah operating licenses for
both units limit reactor thermal power to 3411 MWth.
The following four
evaluations were suggested:
a.
An evaluation to determine which of the two power indications should
be used to comply with the license condition which limits thermal
power to 3411 MWth.
b.
An evaluation of a specific eight hour period where thermal power, as
indicated by the U0908 computer point, indicated an eight hour
average in excess of 3411 MWth.
c.
An evaluation of the same eight hour period for compliance with axial
flux difference TS requirements stated in TS 3.2.1.
d.
An evaluation of a specific period of 52 minutes where thermal power
exceeded 100.75%.
1
This discussion addressed items a, b, and d above.
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On April 19, 1989, at approximately 9:50 a.m. , the inspector noted that
{
the Unit 1 power range nuclear instruments were reading approximately 101%
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reactor thermal power.
The Unit Operator informed the inspector that they
had an increase in the grid power factor that had caused an increase in
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turbine power.
The governor valve limiters installed on the Sequoyan
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turbine generators are not used by the utility to mitigate these power
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swings.
)
At this time the reactor power as indicated en the plant computer (P-250)
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data point U1118 (instantaneous reactor thermal power) was 3415.6 MWth.
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The operator explained that he had been reducing power over several
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minutes to reduce the average power over his 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shift (i.e. , by
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approximately 3:00 p.m.) to less than licensed power of 3411 MWth.
The
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inspector discussed with the operator that both his instantaneous and his
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8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> average power were above the licensed limit.
The inspector
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questioned whether correcting the condition over a 5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> period was
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adequate. At 10:00 a.m. the P-250 point U0908, showing the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> average
of U1118, was reading 341t 0 MWth.
Tte operator took action to reduce
power, and by 10:00 a.m. the one hour aserage reactor power shown on P-250
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point U0018 was indicating less than licensed power, and by 12:01 p.m. the
8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> average power was down to 3408.6 MWth.
The operators explained that they were operating under TS interpretation
.119 which referenced guidance NRC has issued in this area.
The licensee
inferred from the ' guidance that the- average ;,ower level over any
eight hour shift- should not exceed the full steady-state licensed power
level and that it was acceptable to briefly exceed the licensed power
level by as much as 2% for as long as 15 minutes', but in no case should
that 102% power level be exceeded.
The 11cea:d s TS interpretation
referenced an August 22, 1980 internal NRC memorandum titled Discussion of
" Licensed Power Level", which was written to give inspectors. guidance and
address a uniform basis for enforcement.
The menorandum clearly states
that this guidance is not an NRC-wide agreement. Oiscussion with the NRR
technical staff indicates that the memo assumed trat the licensee was
controlling their power excursion and taking an active approach to
reducing the power back below 100% should they exceed that valce.
The inspector noted a period, beginning at 6:33 a.m. on April 19, 1989,
during which reactor power went above 3411 MWth, proceeded to 3436.1 MWth .
(approximately -100.75%), and returned to 3411 MWth at 7:37 a.m.
During
this time there was no apparent operator action to reduce reactor power.
Op..rators apparently waited for the grid power factors to drive thermal
power back down.
The inspector discussed this issue with the plant manager.
He committed
to maintain the plant below 100% reactor power as indicated on the nuclear
instruments until these issues can be reviewed and an acceptable
!
alternative operating method can be determined, reviewed and implemented.
The NRC recognizes that the licensee desires to operate at rated thermal
!
power on a continuous basis.
The NRC also recognizes that power tran-
sients will occur which cause thermal power to change without operator
actions.
The NRC expects that the licensee will take prompt corrective
action to reduce thermal power to the license limit when it is discovered
)
to be above the license limit.
Issuance of a TS interpretation to
l
operators which infers the acceptability of operating above the licensed
!
power limit in other than a transient condition is of concern to the NRC.
]
In addition, once thermal power is discovered above the license limit, the
failure by the TS interpretation to require immediate operator action to
reduce power to the license limit is also of concern.
The inspectors were
unable to find any procedures or other guidance provided to operators
which address the proper actions to take for power transients during
steady state operation at Rated Thermal Power.
This is considered a
violation of TS 6.81 for failure to establish, implement, and maintain
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written procedures and is identified as VIO 327, 328/89-15-03.
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During the review of this issue the inspector also reviewed the following
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documents and found them to be acceptable:
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RTI-8
Statepoint- (sic)- Data Collection and Thermal Power
Measurement
51-78-
. Power. ' Range Neutron Flux Channel Calibration by Heat
Balance Comparison.
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TI-2
Calorimetric Calculation,
The review of these procedures also showed that the _ calibration of the
nuclear instruments and the P-250 data point U1118 was acceptable and
shows adequate. reliability of these indications.
Items a, b, and 'd are administratively closed and will be -addressed
further under the review of the violation.
Item c above will. remain open
and continue to be tracked as unresolved item URI 327,328/89-12-01.
8.
Licensee Event Report Followup (92700)
UNIT 2
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(Closed) LER.327/88-043, Revision 1, Inadequate Fire Watch Patrol Resulted
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In A Noncompliance With Technical Specification 3.7.12.
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On September 30, 1988, the inspector brought to the attention of TVA
management a concern involving the adequacy of the SQN fire watch patrols
which was-later cited as VIO 327,328/88-46-01, Failure to Perform Adequate
Fire Watches.
The fire watch patrols are required by action statement (a) to Limiting
Condition for Operation (LCO) 3.7.12 as a compensatory measure for several
non-functional fire barriers in the Auxiliary and Control buildings. This
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action statement requires verification of the operability of area fire
detectors on at least one side of the breached penetration and the
establishment of an hourly fire watch patrol whenever a fire barrier
penetration has been determined to be non-functional.
TVA management
i
initiated an investigation which concluded that the individual who was
assigned to the fire watch route "A" had not made one of the required
hourly patrols because he was not in the protected area for the 10:00 a.n.
EDT to 11:00 a.m.
EDT hourly fire watch on September 28.
The other
individual who was assigned to fire watch route "B" had made all required
,
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hourly patrols, but his route durations appeared inadequate to survey his
assigned fire watch route properly.
No malicious intent was found and
this event is considered to be an isolated occurrence.
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TVA admitted that the reason for these occurrences was that management had
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failed to provide clear and concise guidance on how fire watch individuals
must perform their duties.
Further, management did not provide for
periodic checks of fire watch patrols to ensure adequate implementation of
'
the fire watch program.
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To correct the situation TVA developed and implemented the following
steps:
a.
QA personnel will continue periodic monitoring activities.
b.
Fire Protection Section supervisor has designated specific
individuals to perform periodic monitoring activities.
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c.
Red hardhats were provided to fire watch personnel for unique
identification.
d.
Fire Operations Section Instruction Letter (FDSIL)-10 has been
revised to enhance the fire watch program.
The new revision includes
the following:
Added door numbers to the fire watch checkoff sheets.
Fire watch personnel can not swap routes without prior approval
from the Fire Operations shift supervisor.
Fire watch personnel are required to wear the uniquely
identified hardhats.
Fire watch personnel are required to carry the section
instruction letter and route sheets with them while walking
their routes.
Fire watch personnel are required to follow the route identified
on the route sheet unless not allowed by plant conditions.
Fire watch ruutes were revised to ensure more effective use of
fire watch personnel and reduce confusion of the routes,
e.
Formal training to fire watch personnel has been provided on the
revised section instruction letter which emphasized the importance of
accurate logs and adherence to hourly fire watch routes to fulfill TS
requirements.
!
f.
A veroal warning has been issued to all fire watch personnel stating
that if fire watch personnel fail to perform their duties,
appropriate disciplinary actions will be taken up tu and including
termination.
g.
Follow up training for fire personnel througn observation of
performance will be provided by fire protection engineers to ensure
continued compliance with fire waten requirements as outlined in the
FDSIL-10.
This action will be completed by December 31, 1988.
h.
TVA will evaluate the feasibility of watchman stations for monitoring
fire watch activities by December 31, 1988.
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The inspector verified that above corrective actions had been implemented
and performed a review of additional fire watch activities.
This review
included monitoring of fire watch routes, reviewing and observing watchman
station logs and reviewing VAACS printouts of the fire watches' progress
through the plant.
Fire watch coverage now appears to be adequate.
LER 327/8P-043 is closed.
9.
Event Follow-up (93702)
a.
Sequoyah Unit I and Unit 2 were placed in an unanalyzed condition
during several performances of RHR systelr surveillance.
This
condition was caused when improperly prepared revisions to 51-128.1,
RHR Pump and Piping Venting,51-128.2, Residual Heat Removal Pump
1A-A,51-128.3, Residual Heat Removal Pump 1B-B, SI-128.4, Residual
Heat Removal Pump 2A-A, and SI-128.5, Residual Heat Removal Pump
2B-B, were used to perform system surveillance
on twenty-six
separate occasions from March 22, 1988 through April 20, 1989. The
revised procedures caused RHR system lineups to be changed to
configurations which effectively rendered both trains of RHR
inoperable for short periods of time.
This condition is described in
LER 89-011; Root Cause Investigation Report, SQA 186, Root Cause
Assessment Adverse Action / Condition, dated May 2,1989; and CAQR
SQP89039. Placing both trains outside the design basis and in an
unanalyzed condition should have been identified in the licensee's
procedure change reviews as an Unreviewed Safety Question.
None of
the licensee's investigation reports concluded that an Unreviewed
Safety Question existed as defined by 10 CFR 50.59.
Thr: changes to the surveillance instructions were made in May 1987
frr SI 128.1 and December 1988 for 51-128.2, 128.3, 128.4, and 128.5.
s
Une review process for 51-128.1 included complction of an Unreviewed
Safety Question Determination worksheet, SQA 119, Evaluation of
Changes, Tests, or Experiments, and a qualified informal PORC review
documentation.
These independent reviews concluded that no unre-
viewed safety question, reduction of safety margin, or possibility of
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increasing the chances for an accident or malfunction of a different
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type than those previously analyzed would result from the issuance or
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use of the instruction.
During accident conditions, the RHR pumps provide automatic low-head
safety injection flow to the RCS following a safety injection signal
(SIS) and also provide suction flow to the centrifugal charging pumps
(CCPs), and the safety injection pumps after swapover to the con-
tainment sump from the refueling water storage tank (RWST).
Specif-
ically, at least one train of low-head safety injection flow is
assumed to be delivered to all four RCS cold legs.
During a LOCA,
three of these flow paths deliver water to the RCS while the fourth
is assumed to be spilling from the faulted leg.
Any configuration
which would result in less than three cold leg injection paths is
currently outside the design basis of Sequoyah Nuclear Plant.
Also,
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the' design basis of the plant does not include having the hot leg
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injection path open via FCV-63-172 while the cold leg paths are also
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open. . This condition would reduce the estimated flow to the cold
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legs to less than that 7.isumed in the accident analysis.51-128.1, originally issued in May 1967, has a step which momentarily
opens FCV-63-172 to facilitate RHR system venting. Opening the 63-172
valve places the RHR system in a configuration that is outside the
design basis if perforried during Modes 1, 2 or 3, when the RHR system
is required to be operaDie.
The licensee's investigations revealed
l
24 separate instances when the 63-172 v.31ve had been opered while the
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cold legs ' were also aligned during modes when the RHR system was
required to be operable.
SI-128.2,128.3,128.4, and 128.5 were revised in December 1988 to
include steps to momentarily shut FCV-63-93, FCV-63-94, (cold leg
isolation valves), and to shut FCV-74-33 and FCV-74-35, (P.MR
cross-connect isolation valves) to facilitate RHR pump ve: sting.
Shutting either the 63-93 or the 63-94 valve will isolkte two
separate cold legs, putting the plant into an unanalyzed condition.
Closing either 74-33 or 73-35 isolates the cross-connect, which,
considering the single failure of either pump, would invalidate the
assumption that either pump could supply all four cold legs during an
accident.
TVA investigation revealed. that the instruction had been
used on two separate occasions, since the incorporation of the change
in December 1988, while the plant was in Modes 1, 2, or 3.
The
instances occurred on Unit 1 on February 24, 1989, and March 1, 1989.
Of the 26 performances with these surveillance since their modifi-
cation, none placed the RHR systems in an inoperable condition for
periods of time in excess of the time limitations contained in the
action statements of TS 3.b.2 or of TS 3.0.3.
However, operators
were unaware that thest: action statements were applicable because of
the inadequate surveillance.
The SQA-119 screening review form completed to document the
December 1988 revisions to the SI-128.2 through 128.5 instructions
determined that the change consisted only of a format change and the
testing methods remained the same.
As such, the licensee determined
that no further evaluation was necessary and a safety evaluation of
the change was not performed.
This conclusion was concurred in by
the independent qualified reviewers as documented on the change
control document attached to the instruction change package.
The
processes used to evaluate the safety, adequacy, deportability, and
applicability of the 10 CFR 50.59 rule in general failed to
determine, for SI-128.1 through 128.5, that an unreviewed safety
question existed.
This is considered to be a violation 10 CFR 50.59
for processing a change that involved an unreviewed safety question
and is identified as VIO 327,328/89-15-04, example a.
The licensee discovered the problems with the procedures when the
responsible supervisor was informed by the SOS on duty that he
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believed the surveillance instructions were unacceptable. based on his
determination that the hot leg isolation valve could not be open at
?
the same time as the cold leg' isolation valves, and that shutting the
cross-connect isolation valves and/or the pump discharge isolations
could cause both trains to be inoperable.
A scheduled performance of
SI-128.1 was stopped by the Operations staff on April 10.
The
responsible section, Mechanical Test, reviewed the procedure and the
file for that procedure and discovered that NRC IE Information Notice 87-01, described a condition that could render both trains of RHR
The described situation involved the shutting of the RHR
~
cross-connect valves and/or the pump discharge isolation valves.
The
notice warned recipients that this configuration would cause both
trains to be inoperable.
When the responsible mechanical test
engineer realized that the procedures involving pump venting,
SI-128.2 through 51-128.5, isolated two of the four cold legs,
Condition Adverse to Quality (CAQR) report SQP890239 was initiated.
The C.AQR concluded that TS LCO 3.0.3 should have been entered during
the performances of RHR venting procedures.
The NRC was notified on
April 13 that the past performances of the procedures had placed the
plant in TS 3.0.3, and in a configuration outside the design' basis.
The responsible Mechanical Test engineer believed that he was able to
control all performance
of SI-128.1 through SI-128.5 because he was
responsible for issuing the performance packages.
He did not,
therefore, make immediate corrections to the procedures.
On April
17, a performance package of SI-128.1 war prepared by the Planning
and Scheduling Section and sent to the ec' trol room.
SI-128.1 was
performed on April 20, with the attendant errors still in the
procedure.
This instance was discovered when the Mechanical Test
engineer was attempting to locate the last controlled copy of the
performance package and called the control room.
He was then told
that SI-128.1 had been performed that morning.
The performance of
SI-128.1 after it was known to be inadequate constitutes a failure to
maintain adequate written procedures as required by TS 6.8.1 and is
identified as VIO 327,328/89-15-05, example a.
b.
Unit 2 entered Mode 3 at 4:30 p.m. on April 6,1989.
At 4:55 p.m.,
the BIT to BAT recirculation was stapped and the BIT recirculation
valves .ere shut to stop back leakage from the RCS to the BIT.
This
backleakage was causing dilution of the BIT and the BAT.
The inspector questioned this procedue b9cause the recirculation
path provides the only method of ensuring tcat the proper BIT volume
is maintained.
Assurance that the EIT is full, (900 gallons), is
provided by the absence of a low flow alarm ra the recirculation path
out of the BIT.
TS 3.5.4.1 require!, the BIl to be nperable in Mode 3
with a minimum volume of 900 gallons.
TN action statement for this
requirement is to restore the BIT to ac sperable condition within one
hour or be in Hot Standby and borated to the appropriate shutdown
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margin within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
,
The licensee's position was that even though no on-line verification
..
of BIT volume exists when the BIT is off recirculation, there is no
reason to suspect that the volume is being reduced.
Further, the
surveillance requirement = is to verify that proper volume is present
every 7 ' days.
The licensee therefore concluded that taking the BIT
off recirculation is allowed ,as long as the surveillance to verify
volume is - performed within 7 days.
This position was based.on an
existing licensee TS interpretation.
After the Resident Inspector
,
notified the plant staf f of their concern regarding this issue, the
BIT was placed back on recirculation at 10:02 a.m. on' April'10,1989.
Plant Management agreed that recirculation will be maintained until
the issue would be resolved.
The question of whether the BIT was inoperable was reviewed by the
NRC technical staff at the request of the Resident . Inspector and the
NRC Project Manager.
The FSAR states in part that to prevent cold
spots and stratification within the tank during normal operation,_the
contents of the boron injection tank are continuously recirculated.
A sparger is provided on the . inlet to the tank which, coupled with'
recirculation, will prevent channeling and ensures homogeneity of the
. boric acid solution.
Any large scale leakage is detected by a flow
indicator and alarm located on the discharge line.
The staff
concluded that the BIT must remain on recirculation .to ensure that
proper volume is maintained and to prevent possible boron crystal 11-
zation inside the BIT.
The staff concluded that the BIT was inoper-
able during the time the BIT. was off recirculation and that this
change in mode of operation of the BIT involved an unreviewed safety
question.
Processing a change to the facility or procedures which
are described in the FSAR without performing a review pursuant to
10 CFR 50.59 is considered to be a violation and is identified as
VIO 327, 328/89-15-04, example b.
The inspector questioned the licensee's determination that taking the
BIT off recirculation did not require entry into the LCO.
In
addition, the inspector questioned whether a proper review pursuant
to 10 CFR 50.59 had been completed.
The normal recirculation lineup
'
is specified in 501 62.1, Chemical and Volume Control System.
AI-4,
Preparation, Review, Approval and use of Site Procedures /Instruc-
tions, establishes the method and controls required by TS for
revision and use of site procedures.
AI-58, Maintaining Cognizance
I
of Operation Status Control, implements the review and approval
I
requirements of TS when handwritten instructions are used where an
approved procedure does not exist.
The change in the BIT system
lineup was required to be controlled by approved procedures.
Neither
of these methods were used.
Changing the BIT lineup without an
approved procedure is considered to be a violation of TS 6.8.1 for
failure to follow procedures and is identified as VIO 327, 328/
89-15-05, example b.
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AI-58, Maintaining Cognizance of Operation Status Control, Appendix
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governs the production of handwritten instructions when an
approved S0I procedure does not exist. It requires only that the SOS
review and approve the procedure, and that an Independent Qualified
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Reviewer, whict may include the STA or other shift personnel who are
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certified as Qualified Reviewers, also review and sign the change
l
form.
This process was not performed until April 8,1989, two days
I
after the configuration change had been made and challenged by the
'
inspector.
The AI-58, Appendix E worksheet determined that there was
)
no unreviewed safety question, and that no safety issues were
'
involved.
l
During the period.that the BIT was not on recirculation with the BAT,
the licensee attempted to meet Surveillance Requirement 4.5.4.1 by
l
performing Surveillance Instruction 3, Daily, Weekly and Monthly
Logs.
51-3 verifies that the BIT has an adequate borated water
volume while the BIT is in recirculation with the BAT.
SI-3
]
accomplishes this in step 3.1.4, by verifying that greater than 900
,
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gallons volume exists in the BIT by the absence of the low recir-
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culation flow alarm FIS-63-43.
The validity of using the absence of
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the alarm to- determine BIT volume as required by surveillance
requirement 4.5.4.1 is based on the BIT being in recirculation.
The
FSAR states that the purpose of this alarm is to detect any large
scale leakage within the BIT.
The alarm will only perform this
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function if the BIT is in recirculation.
During the period that the
BIT was not in recirculation, approximately 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, the low
!
recirculation flow alarm was actuated.
Also, because of the
previously identified problems with valve leakage, performing SI-3
after the reestablishment of recirculation only ensures that the
volume was present during the period of recirculation and does not
ensure that the volume was or will be correct during the period in
i
which recirculation was or will be stopped.
Therefore SI-3 did not
adequately implement surveillance requirement 4.5.4.1 to determine
the operability of the BIT during periods when recirculation was
stopped, and as such did not prove operability of the BIT during the
absence of recirculation.
This is a violation of TS 3.5.4.1, BIT
Operability, and is identified as VIO 327,328/89-15-06, example a.
c.
At 6:00 a.m. May 5,1989, the SOS on duty noticed that the Unit 2
Intermediate Range (IR) High Flux trip bistables were not indicating
!
tripped while the reactor plant power was at 73%.
The setpoint of
the bistables is approximately 25% reactor power as indicated on the
power range monitors.
Further observation indicated that the IR
channels were low compared to the same instrument readings on Unit 1.
The IR channels, N-35 and N-36 were declared inoperable and the LC0
for TS 3.2.1 was entered.
LCO 3.2.1 allows continued operation with
both IR channels inoperable if power is above 10%.
Licensee review
of this event revealed that during the recent outage, the storage
baskets for the Source and Intermediate Range detectors had been
pulled back from their normal position against the reactor vessel
wall to a maintenance position approximately 21 inches from the
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normal position.
This configuration changed the amount.of leakage
neutron flux at the detectors and caused them to indicate lower than
actual power levels.
This event is described in LER 50-328/89-006.
The detector arrangement has Source Range (SR) channel N-31 and IR
channel N-35 in one basket and SR N-32 and IR N-36 in the other-
basket.
The storage baskets are located 180 degrees apart outside
the vessel wall.
The baskets and detectors had been retracted on
March 19 and March 20 for N-31/N-35 and N-32/N-36 respectively, while
Unit 2 was shutdown for refueling.
This change was performed ir. ar.
attempt to reduce noise spiking that had occurred intermittently on
,
the SR channels.
Moving the detectors back allowed the detector
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cables to straigSten out slightly and reduce crimping and chafing
that was postulated to be causing the spiking.
In addition to the
position change, plant personnel removed ground straps which
connected the SR and IR detectors in each basket.
This work was done
under WP 1192-01 and controlled by IMI-92-SRPC.
IMI-92-SRPC, which
controls detector positioning, was revised by ICF 98-0314 after the
detectors had been withdrawn to the maintenance position.
The major
change incorporated by the ICF was to delete the steps returning the
baskets to the normal position.
This position change was recommended
by Westinghouse in a letter dated February 28, 1989 to optimize
detector output (reduce noise on the channel).
The ICF and the accompanying screening review did not identify that
the detector movement would require readjustment or recalibration of
the source range channels.
Further, the ICF and the screening review
did not mention that the IR channels would also be affected.
A
screening review is required by SQA 119, Evaluation of Changes,
Tests, or Experiments,. to evaluate changes and insure that safety
evaluations are performed when the changes could have an affect on
,
nuclear safety or create an unreviewed safety question.
The
!
instrument technician who prepared the ICF and the accompanying
screening review concluded that changing the position of the
l
detectors did not involve a change in the facility as described in
the FSAR.
The screening review form question for this item, Question
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- 1, is marked "NA".
Question #2 is stated as: "Does the proposed
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change involve new procedures or instructions or revisions to
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instructions or procedures that would make descriptions in the SAR or
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plant TS invalid and which could impact nuclear safety ?".
The
i
preparer marked this answer box "Yes".
The screening review form
states that marking "Yes" requires a safety evaluation to be
performed.
Despite this requirement to initiate and perform a safety
evaluation, neither the preparer nor reviewers of the ICF and its
accompanying screening review recognized the requirement when
reviewing the screening review form and therefore did not perform a
safety evaluation.
SQA 119, Revision 13, controls the screening
review and independent review process.
AI-43 controls the Indepen-
dent Qualified Reviewer process that requires independent review of
the ICF and the accompanying screening review.
Failure to follow the
requirements of SQA-119 and AI-43 by not ensuring that a safety
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20
s
evaluation accompanied the screening review form as required is
considered a violation of the requirements of TS 6.8.1 for failure to-
follow written procedures and is identified as VIO 327,328/89-15-05,
example c.
In the space for " Justification" on the screening review form, the
preparer concluded that "... Rep;sitioning the detectors will not -
adversely affect operation of the detectors in any way...".
The
change to the position of the intermediate range detectors was
sufficient to cause the IR High Flux trip bistables to be inoperdie
which was discovered after the unit was restarted.
The screening
review form was reviewed and approved by an Instrument Engineer.
This review was inadequate because three separate errors made by the
preparer were not identified and corrected.
The' question marked
"Yes" which required a cafety evaluation was not complied with, the
effect of moving the baskets did not address the impact on the IR
channels, and the ' conclusion that the detectors would not be
adversely affected was not correct.
In addition, the completed ICF
and the accompanying screening review form were reviewed and
concurred with by three separate independent qualified reviewers
including personnel from Systems Engineering and QA.
None of these
independent reviews recognized any of the three errors associated
with the ICF and the screening review.
None of these separate
reviews questioned the operability of the source range detectors and
associated reactor trips, the involvement of the IR detectors and
their associated reactor trips, the missing safety evaluation, or the
impact of moving detectors away from their source.
The change in the
position of the detectors resulted in both IR high' flux bistables
being inoperable.
The review performed pursuant to 10 CFR 50.59 was
not adequate in that it did not identify the effect the procedure
change would have on the nuclear instrumentation or reactor trip
system and is identified as VIO 327,328/89-15-04, example c.
The maintenance and modifications performed on the instruments were
performed while the plant was in Mode 5, Cold Shddown.
Operations
personnel (the duty Shift Operating Supervisor, 503). are notified of
work in progress under the requirements of the Vo.k Package (WP)
program.
Operations was therefore aware that the detectors were to
be withdrawn, but assumed that following the ground strap removal the
detectors would be returned to their original position.
The decision
to leave the detector baskets withdrawn was made as a modification to
the original WP.
The change to the original scope of the WP was not
processed as a major change, and the ICF was not marked as requiring
notification of the SOS for a configuration change even though the SR
I
detectors, which the operators were using to monitor the core, were
)
directly affected.
The fact that this change could be processed
without the notification or the knowledge of the licensed control
room operating staff is of concern.
In addition, NRC inspectors had
suggested during the closecut of Unit 2 containment during the early
'
1988 restart that the required position of the detector baskets be
added to the containment closecut checklist.
Had that been done, an
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21
additional check on proper detector location may have helped keep the
operators informed of what was happening with the equipment they were
using to monitor the core.
Subsequent to the maintenance and modifications performed on the
detectors, Unit 2 was taken from Cold Shutdown (Mode 5) to Power
Operation (Mode 1).
Surveillance
performed on the SR channels
effectively adjusted them which eliminated any effects of the
location change.
The SR channels were, therefore, operable during
the UnN 2 Mode changes associated with the startup.
Comparable
surveillance do not adjust the IR channels until after the plant
reaches 100% power, at which time the IR High Flux Trip setpoint is
checked and adjusted to coincide with a setpoint equivalent to 25% as
determined by the Power Range excore detectors.
The trip setpoint
for the bistable was set prior to plant startup at a current value
calculated to correspond with 25% power.
As a result, mode changes
were made with both IR high flux bistables inoperable. This situ-
ation wa; the direct result of a failure to control configuration
of systems and equipment during the maintenance and modification
process.
Due to three trips which occurred while low in power, a
total of 4 separate startups and the attendant mode changes were made
with the IR channels inoperable.
TS 3.3.1.1 requires the IR channels
to be OPFDABLE in Modes 1, 2, and when the reactor trip breakers are
closed with tuc1 in the reactor vessel.
TS 3.0.4 prohibits entry
into an operational mode unless the conditions for the LC0 are met
without reliance on provisions contained in the ACTION requirements.
Contrary to this requirement, mode changes were made on four separate
occasions during the time period from April 13, 1989 to April 25,
1989 with both IR high flux bistables inoperable.
This is a viola-
tion of TS 3.3.1.1 and TS 3.0.4 and is designated VIO 327,328/
89-15-06, example b.
After the licensee staff realized that the IR channels were
inoperable, the appropriate LCOs were entered, the IR High Flux
reactor tt ip was adjusted to a setpoint of 25% power range equiva-
lent, and safety evaluations were performed for three areas.
The
first SE was completed on May 6, 1989, and provided justification for
continued operation with the SR and IR detectors in their withdrawn
position.
The second safety evaluation was completed on May 7, and
provided assurances that the IR channels could be adjusted to reflect
the change in relative position.
The third safety evaluation was
complete on May 8, and determined the impact or contribution of the
!
condition on the previously discussed startups.
The safety evalua-
tion determined that the plant was safe in its present configuration,
that the intended adjustment could be made to restore the IR channels
to Operable status, and that having the IR channels inoperable did
not cause unsafe conditions to exist during the plant startups.
{
1
Instances of improper changes to plant systems and components and/or
control processes have occurred recently at Sequoyah without accompanying
,
safety evaluations.
These include the BIT / BAT recirculation issue
described in IR 327,328/89-12-02 and this report, the RHR system lineup
change that resulted in the plant being in an unanalyzed condition as
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22
described in LER 50-327/89-011 and this report, the IR and the SR
relocation issue which affected the high neutron flux trip as described
in
this
report,
isolation
of
a
pressurizer
PORV with reliance on a TS interpretation rather than a safety evaluation
as described in URI 327,328/89-12-02, and operation with continued high
voltage on the 6.9 kv shutdown boards described in IR 327,328/89-14.
g
Apparent weaknesses in the safety evaluation program, will be addressed as
l
i
URI 327, 328/89-15-07, and will include:
(1) the licensee method of implementing 10 CFR 50.59
(2) independent qualified reviewer program
(3) the experience review program feedback to the safety evaluation
process
10.
Exit Interview (30703)
The inspection scope and findings were summarized on June 5 and on June 9,
1989, with those persons indicated in paragraph 1.
The Senior Resident
Inspector described the 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 inspectors during the inspection.
Inspection Findings:
This routine monthly inspection involved inspection effort by the Resident
Inspectors in the area of operational safety verification including
control room observations, operations performance, system lineups,
radiation protection, safeguards, and housekeeping inspections.
Other
areas inspected included maintenance observations, surveillance testing
observations, review of previous inspection findings, follow-up of events,
review of licensee identified items, and review of inspector follow-up
items.
Results:
The areas of Operations, Maintenance, and Surveillance were adequate and
fully capable to support current plant operations.
The
observed
activities of the control room operators were professional and well
i
executed.
Significant weaknesses were identified in the licensee's
implementation of the 10 CFR 50.59 safety evaluation processes, the
licensee's TS interpretation process and the independent qualified review
process.
(Closed) NCV 327, 328/89-15-01, " Licensee Identified Minor Contamination
of Two Individuals under RWP 89-00012"
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(0 pen)
VIO 327, 328/89-15-02, " Configuration of RM 90-404 Detector
Cable was not controlled"
(Closed) URI 327, 328/89-12-02, " Recirculation of BIT"
(0 pen)
VIO 327, 328/89-15-03, " Reactor Operation above 3411 MWth"
(Closed) LER 327/88-043, Revision 1, " Inadequate Fire Watch Patrol
Resulted in a Non-compliance with TS 3.7 12"
(0 pen)
VIO 327, 328/89-15-04, "Three Examples of Inadequate Unreviewed
Safety Question Determinations per 10 CFR 50.59"
l
(0 pen)
VIO 327, 328/89-15-05, "Three examples of Failure to Establish
and Implement Procedures per TS 6.8.1"
(0 pen)
VIO 327, 328/89-15-06, "Two Examples of Failure to Comply with
TS LC0 Requirements"
(0 pen)
URI 327, 328/89-15-07, " Apparent Weakness in the Safety
Evaluation Program"
11.
List of Acronyms and Initialisms
ABGTS-
Auxiliary Building Gas Treatment System
ABI -
Auxiliary Building Isolation
ABSCE-
Auxiliary Building Secondary Containment Enclosure
AFW -
Aardnistrative Instruction
AI
-
A01 -
Abnormal Operating Instruction
AVO
Auxiliary Unit Operator
-
AS05 -
Assistant Shif t Operating Supervisor
ASTM -
American Society of Testing and Materials
,
'
BIT -
Boron Injection Tank
BFN -
Browns Ferry Nuclear Plant
C&A
Control and Auxiliary Buildings
-
CAQR -
Conditions Adverse to Quality Report
Component Cooling Water System
-
Centrifugal Charging Pump
-
CCTS -
Corporate Commitment Tracking System
CFR -
Code of ' Federal Regulations
COPS -
Cold Overpressure Protection System
CPU -
Central Processing Unit
-
CSSC -
Critical Structures, Systems and Components
CVCS -
Chemical and Volume Control System
Containment Ventilation Isolation
-
Direct Current
-
DCN
Design Change Notice
-
1
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Diesel Generator
j
.DG
-
DNE -
Division of Nuclear Engineering
i
Engineering Change Notice
f
-
ECCS -
'
1
-
Emergency Instructions
)
EI
-
Emergency Notification System
-
<
E0P
Emergency Operating Procedure
-
Emergency Operating Instruction
-
ERCW -
Essential Raw Cooling Water
l
Engineered Safety Feature
-
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FCV -
Flow Control Valve
l
FSAR -
Final Safety Analysis Report
GDC
General Design Criteria
-
GOI -
General Operating Instruction
GL
Generic Letter
-
HVAC -
Heating Ventilation and Air Conditioning
l-
Hand-operated Indicating Controller
-
' H0
Hold Order
-
-
Health Physics
ICF
Instruction Change Form
-
IDI
Independent Design Inspection
IFI
Inspector Followup Item
-
IM
Instrument Maintenance
-
IMI
Instrument Maintenance Instruction
-
IN
NRC Information Notice
-
I/O -
Input Output
IR
Inspection Report
-
KVA
Kilovolt-Amp
-
KW
Kilowatt
-
KV
-
Kilovolt
LER
Licensee Event Report
-
LCO
Limiting Condition for Operation
-
Licensee Identified Violation
LIV
-
LOCA -
Loss of Coolant Accident
Main Contrni Room
-
MI
Maintenance Instruction
-
-
Maintenance Report
MSIV -
NB
NRC Bulletin
-
NOV -
NQAM -
Nuclear Quality Assurance Manual
NRC
Nuclear Regulatory Commission
-
OSLA -
Operations Section Letter - Administrative
OSLT -
Operations Section Letter - Training
Office of Special Projects
-
Precautions, Limitations, and Setpoints
-
Preventive Maintenance
NCV -
Non-cited Violation
Parts Per Million
-
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[
Post Modification Test
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PORC -
Plant Operations Review Committee
l
P0RS -
Plant Operation Review Staff
P-QIR
Plant Quality Information Requested / Released
-
PRO
Potentially Reportable Occurrence
l
,
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Quality Assurance
t
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l
Quality Control
-
i
RCDT -
Reactor Coolant Drain Tank
Reactor Coolant Pump
-
RCS -
,
l
-
Regulatory Guide
RHR -
-
Radiation Monitor
R0
-
Reactor Operator
Rod Position Indication
-
RPM -
Revolutions Per Minute
RTD -
Resistivity Temperature Device Detector
Radiation Work Permit
-
RWST -
Refueling Water Storage Tank
Safety Evaluation Report
-
-
Surveillance Instruction
-
SMI
Special Maintenance Instruction
-
50I -
System Operating Instructions
SOS
Shift Operating Supervisor
-
SQM -
Sequoyah Standard Practice Maintenance
SQEP -
Sequoyah Engineering Projects
SQRT -
Seismic Qualification Review Team
SR
-
Surveillance Requirements
SR0
Senior Reactor Operator
-
550MI-
Safety Systems Outage Modification Inspection
SSQE -
Safety System Quality Evaluation
i
SSPS -
Solid State Protection System
'
-
STI -
Special Test Instruction
TACF -
Temporary Alteration Control Form
TAVE -
Average Reactor Coolant Temperature
TDAFW-
Turbine Driven Auxiliary Feedwater
TI
-
Technical Instruction
TREF -
Reference Temperature
TROI -
Tracking Open Items
TS
-
Technical Specifications
TVA -
Tennessee Valley Authority
UHI -
Upper Head Injection
U0
Unit Operator
-
URI -
Unresolved Item
USQD -
Unreviewed Safety Question Determination
VDC -
Volts Direct Current
VAC -
Volts Alternating Current
VAACS-
Vital Area Acess Control System
1
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26
WCG
Work Control Group
-
,
WP
Work Plan
-
- .
Work Request
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