ML20209G230
| ML20209G230 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 04/08/1987 |
| From: | Harmon P, Jenison K, David Loveless, Mccoy F, Poertner W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20209G174 | List: |
| References | |
| 50-327-87-02, 50-327-87-2, 50-328-87-02, 50-328-87-2, NUDOCS 8704300504 | |
| Download: ML20209G230 (21) | |
See also: IR 05000327/1987002
Text
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UNITED ShATES
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NUCLEAR REGULATORV COMMISSION
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ATLANTA, GEORGn A 30323
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Report Nos.:
50-327/87-02, 50-328/87-02
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Licensee: Tennessee Valley' Authority
6N38 A Lookout Place
1101 Market Street
Chattanooga, TN 37402-2801
Docket Nos.:
50-327 and 50-328
License Nos.: OPR-77 and DPR-79
Facility Name:
Sequoyah Units 1 and 2
Inspection Conducted:
January 6, 1987 thru February 5, 1987
Inspectors
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D. P. Loveless, Resident)Mipec,t r
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Approved by:
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F. R."NcCoy,' ChieW5est40n 1
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Division of TVA Projects
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SUMMARY'
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, Scope:
This routine announced inspection involved inspection onsite by the
= resident inspectors in - the areas of:
operational safety verification
(including operations performance, system lineups, radiation protection,
safeguards and . housekeeping inspections); maintenance observations; review of
previous inspection findings; followup of events; review of licensee identified
items; and review of inspector followup items.
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Results: Two violations were identified.
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327,328/87-02-01,
failure to adequately control
field changes,
paragraph 7
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327,328/87-02-02,
failure to establish, maintain and implement safety-
related procedures, paragraphs 9 and 13
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Four unresolved items were identified.
327,328/87-02-03,
modification to control room ceiling, paragraph 5
327,328/87-02-04,
access control, paragraph 5-
327,328/87-02-05,
adherence to Health Physics requirements, paragraph 5
327,328/87-02-07,
two spills of reactor coolant, paragraph 9
No deviations were identified.
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REPORT DETAILS
1.
Licensee Employees Contacted
- H. L. Abercrombie, Site Director
- L. M. Nobles, Acting Plant Manager
- B. W. Willis, Acting Power Plant Superintendent
- B. M. Patterson, Maintenance Superintendent
- R. J. Prince, Radiological Control Superintendent
- M. R. Harding, Licensing Group Manager
- L. E. Martin,' Site Quality _ Manager
- D. W. Wilson, Project Engineer
R. W. Olson, Modifications Branch Manager
- J. M. Anthony, Operations Group Supervisor
- R.'V. Pierce, Mechanical Maintenance Supervisor
M. A. Scarzinski, Electrical Maintenance Supervisor
- H. D. Elkins, Instrument Maintenance Group Manager
J. T. Crittenden, Public Safety Service Chief
R. W. Fortenberry, Technical Support Supervisor
- G. B. Kirk, Compliance Supervisor
-D. C. Craven, Quality Assurance Staff Supervisor
- J. H. Sullivan, Regulatory Engineering Supervisor
J. L, Hamilton, Quality Engineering Manager
D. L. Cowart, Quality Engineering Supervisor
H. R. Rogers, Plant Operations Review Staff
R. C. Burchell, Compliance Engineer
- R. H. Buchholz, Sequoyah Site Representative
Other licensee employees contacted included technicians, operators, shift
engineers, security force members, engineers and maintenance personnel.
- Attended exit interview
2.
Exit Interview
The inspection scope and findings were summarized with the Plant Manager
and members of his staff on February 4, 1987.
The two violations
described in this report's summary paragraph were discussed.
No devia-
tions were discussed.
The licensee acknowledged the inspection findings.
The licensee did not identify as proprietary any of the material reviewed
by the inspectors during this inspection.
During the reporting period,
frequent discussions were held with the Site Director, Plant Manager and
other managers concerning inspection findings.
3.
Licensee Action on Previous Inspection Findings (92702)
(Closed) Violation 327,328/84-38-02, Failure to have Adequate Nonconfor-
mance Report Procedure.
The licensee stated in their response to the
violation dated March 15, 1985, that the violation was not caused by the
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lack of procedure.
They suggested instead that the violation resulted
from a lack of management control which did not ensure the timely
reporting of the potential nonconformance.
The licensee committed to instructing all office of engineering employees
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at Sequoyah and in Knoxville on the timeliness requirements of documenting
and reporting potential nonconforming conditions.
This information was
presented by memorandum to all Sequoyah engineering project employees on
April 9, 1985.
Knoxville employees were informed during informal meetings
with their supervisors as designated in an April 9,1985 memorandum from
John A. Raulston.
Furthermore, the licensee has committed to training all
office of engineering and nuclear power employees on the use of their new
procedures on conditions adverse to quality in response to NRC Order
85-49. This training was completed in October 1985.
This item is closed.
4.
Unresolved Items
Unresolved items (URIs) are matters about which more information is
required to determine whether they are acceptable or may involve viola-
tions or deviations.
Four unresolved items were identified during this
inspection, and are identified in paragraphs 5 and 9 of this report.
(Closed) URI 327,328/86-60-11, Control of Safety-Related Pumps.
This item
concerned failure of the emergency core cooling system (ECCS) pum
meet the flow balance requirements of Technical Specification (TS) ps to
4.5.2.h
during the performance of the surveillance instructions, failure of the
turbine driven auxiliary feedwater (AFW) pump to reach rated speed and
flow, acceptability of AFW pump dischar
errors in the inservice inspection (ISI)ge piping vibratiori, and technical
procedure for RHR testing.
TS 4.5.2.h requires the performance of a flow balance test during shutdown
following completion of modifications to the ECCS subsystem that alter the
subsystem flow characteristics.
As a result of concerns raised in
Westinghouse bulletin NSD-TB-80-11 concerning pump degradation,llanceSequoyah
began performing the ECCS flow balance on n 18 month survei
interval.
The performance of the flow balance test
indicated that at the
end of the 18 month period the flow balance for the centrifugal charging
pumps would be just below the minimum acceptable value.
The licensee has
received a change to the TS to reduce the required value for the sum of
the injection line flow rates (excluding the highest flow rate) from 346
gpm to 316 gpm.
The flow balance is now conducted with the pump mini-flow
valves open where previously they were closed.
However, discussions with
the mechanical test engineers responsible for conducting the surveillance
determined that the TS change to a lower flow requirement should imarove,
the pass rate of this surveillance activity.
The licensee has a'so
initiated a design change request to replace the positive displacement
charging pump with a centrifugal pump.
This will reduce the run times on
the centrifugal charging pumps therefore the pump degradation between flow
balances would be reduced.
Further review of the licensees performances
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of the ECCS flow balances and the status of the design change request to
replace the positive displacement pump will be inspected during routinely
scheduled NRC ASME Section XI testing inspections. This URI is closed.
Concerning the failures of the turbine driven auxiliary feedwater pump to
reach rated speed and flow, the inspector reviewed the Licensee Event
Reports associated with these events for adequacy. The inspector reviewed
the licensee's potential reportable occurrence (PRO) files to determine if
this constituted a recurring problem when running the turbine driven
auxiliary feedwater pumps. These events appeared to be isolated incidents
and not recurring failures with similar root causes. PR0 1-87-56 has been
issued against the AFW System.
The other items addressed in this URI concerning pump performance were
addressed in NRC Inspection Report 86-71 and are considered to be
resolved. URI 327,328/86-60-11 is closed.
(Closed) URI 327,328/86-71-05, Containment Spray System Field Change
,
Request.
This item is addressed as violation (VIO) 327,328/87-02-01 in
paragraph 7d of this report. This URI is closed.
(Closed) URI 327,328/86-71-06 Control Auxiliary Building and Containment
Isolations. This item is addressed as violation (VIO) 327,328/87-02-02.in
paragraph 9a. This URI is closed.
(Closed) URI 327,328/86-71-04 Adequacy of Surveillance Instruction (SI)
400.1, Liquid Effluent Batch Release
A discrepancy was identified between two different methods used to calcu-
late background radiation on effluent releases.
For the particular
release that the inspector observed, the prccedural discrepancy did not
have any safety significance.
SI 400.1 and Technical Instruction 18,
radiation monitoring, are both under review in the licensee's surveillance
instruction review program.
This URI will be evaluated under the overall
appraisal of the licensee's surveillance instruction review program URI
327,328/86-60-10 described in paragraph 7 of this report.
URI 327,328/
86-71-04 is closed.
(Closed) URI 327,328/86-42-02, Vendor Supplied Requirements
This issue was resolved in subsequent environmental qualification inspec-
tions conducted at Sequoyah nuclear plant and is therefore closed.
5.
Operational Safety Verification (71707)
a.
Plant Tours
The inspectors observed control room operations, reviewed applicable
logs, conducted discussions with control room operators, observed
shift turnovers, and confirmed operability of instrumentation.
The
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inspectors verified the operability of selected emergency systems,
and verified compliance with TS limiting conditions for operation
(LCO). The inspectors verified that maintenance work orders had been
submitted as required and that followup activities and prioritization
-of work was accomplished by the licensee.
Tours of the diesel generator, auxiliary, control, and turbine
buildings, and containment were conducted to observe plant equipment
conditions, including potential fire hazards, fluid leaks, and
excessive vibrations and plant housekeeping / cleanliness conditions.
The inspectors walked down accessible portions of the following
safety-related systems on Unit 1 and Unit 2 to verify _ operability and
proper valve alignment:
_
residual heat removal system
component cooling water system
safety related heat tracing
During a tour of the control room the inspectors observed water
dripping into pans hung in the false ceilings and onto the ceiling
itself.
The licensee stated that they were in the process of
installing drip pans and gutters in the false ceilings to catch the
water from the leaking control building roof. This modification was
being performed under work request (WR) 8214608 with a plant opera-
tion review committee (PORC) approved unresolved safety question
determination (USQD) in place. The inspector questioned the use of a
WR to perform a safety related modification and the corrective action
time frame. These issues will be reviewed during the next inspection
period. This item is considered unresolved and will be identified as
URI 327,328/87-02-03.
No violations or deviations were identified.
b.
Safeguards Inspection
In the course of the monthly activities, the inspectors included a
review of the licensee's physical security program. The performance
of various shifts of the security force was observed in the conduct
of daily activities including protected and vital area access
controls; searching of personnel and packages; badge issuance and
retrieval; patrols and compensatory posts; and escorting of visitors.
In addition, the inspectors observed protected area lighting, and
protected and vital areas barrier integrity. The inspectors verified
various interfaces between the security organization and the opera-
tions and maintenance organizations.
Specifically, the inspectors
inspected security during outages and verified protection of safe-
guards information.
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During a tour of the Unit 2 west valve room the inspectors-identified
that a security guard did not appear to have direct control over
access to the open valve room which is a vital area.
This issue was
discussed with Regional management and will be identified as URI
327,328/87-02-04 until resolution by . a security specialist is
effected.
No violations or deviations were identified.
c.
Radiation Protection
The inspectors observed Health Physics-(HP) practices and verified
implementation of radiation protection control.
On a regular basis,
radiation work permits (RWPs) were reviewed and specific work activi-
ties were monitored to ensure the activities were being conducted in
accordance with applicable RWPs.
Selected radiation protection
instruments were verified operable and calibration frequencies were
reviewed.
During a routine tour of the auxiliary building, the inspectors
observed a group of workers gathered around one train of the emer-
gency gas treatment system (EGTS) charcoal tray beds.
The system had
been breached, and one worker had his head and part of his shoulders
within the internal boundary of the system.
The worker inside the
EGTS system appeared to be performing a task with a pair of channel-
lock type pliers.
When questioned, another worker stated that an HP
technician had taken a smear of the area in addition to frisking the
area.
The worker further stated that the HP technician had directed
the workers to wait until he had counted the smear before entering
the internals of the EGTS charcoal tray bed area to commence work.
The inspector questioned the worker who was inside the EGTS system
boundary and he stated that he was aware of the instruction given by
the HP technician.
The outside of the EGTS was not marked poten-
tially contaminated, however, it was marked " notify HP prior to
entry."
Survey number D-87-0162 found no detectable transferable
surface contamination.
Because of the low safety significance of
this particular issue, no violation will be issued at this time.
However, adherence, to directions from HP technicians and compliance
with HP practices will be reviewed as URI 327,328/87-02-05.
No violations or deviations were identified.
6.
Monthly Surveillance Observations (61726)
The inspectors observed / reviewed TS required surveillance testing and
verified that testing was performed in accordance with adequate
procedures; that test instrumentation was calibrated; that LCOs were met;
that test results met acceptance criteria requirements and were reviewed
by personnel other than the individual directing the test; that deficien-
cies were identified, as appropriate, and that any deficiencies identified
during the testing were properly reviewed and resolved by management
personnel; and that system restoration was adequate.
For completed tests,
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the inspector verified that testing frequencies were met and tests were
performed by qualified individuals.
The inspector reviewed the performance of SI-244, Periodic Functional
Tests of Radiation Effluent Monitoring Instruments, which was intended
to satisfy TS surveillance requirements for the third quarter of 1986.
The licensee documented that this surveillance had been missed in its
September 1986 monthly report.
The inspector reviewed the recalibration
of flow indicator FI-77-230 and level transmitter LT-27-225.
This issue
was reported by the licensee in LER 327,328/86-37.
The licensee is currently in a large scale surveillance instruction (SI)
review.
The inspector reviewed that process and determined that the
licensee is now in the third iteration of this review effort.
Significant
technical issues have been identified by the site quality assurance (QA)
organization, which may have contributed to this iteration process.
The
following issues were evaluated by the inspector during a review of the
licensee s SI review process:
LER 327,328/86-30 Inadequate TS Procedure
LER 327,328/86-34 Failure to Perform Surveillance on Time
LER 327,328/86-37 Failure to Perform Surveillance on Time
LER 327,328/86-40 Inadequate Isolation Valve Leak Test
LER 327,328/86-43 Inadequate Containment Leak Rate Test
LER 327,328/86-44 Inadequate Emergency Core Cooling System Procedure
LER 327,328/86-48 Inadequate Emergency Core Cooling System Procedure
LER 327,328/86-50 Inadequate Surveillance Instruction
URI 327,328/86-71-04 Procedural Discrepancies
The above issues will be included with the issues identified in inspector
followup item IFI 327,328/86-60-10, and are administratively closed.
7.
Monthly Maintenance Observations (62703)
a.
Station maintenance activities of safety-related systems and compo-
nents 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 work; activities were accomplished using approved procedures
and were inspected as applicable; procedures used were adequate to
control the activity; troubleshooting activities were controlled and
the repair record accurately reflected what actually took place; func-
tional testing and/or calibrations were performed prior to return-
ing components or systems to service; quality control records were
maintained; activities were accomplished by qualified personnel;
parts and materials used were properly certified; radiological
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controls were implemented; QC hold points were established where
required and were observed; fire prevention controls were imple-
mented; outside contractor force activities were controlled in
accordance with the approved quality' assurance (QA) program; and
housekeeping was actively pursued.
b.
On January 7, 1986, the inspectors observed workers replacing Raychem
heat shrinkable tubing following the Raychem inspections.
Workers
discussed with the inspector the color coding of the tags that were
affixed during Raychem signoffs.
The work was covered by special
maintenance instruction (SMI)-2-302-1, Walkdown Procedure for 10 CFR 50.49 Splices Located in Unit 2 Containment Penetrations, R1.
The
procedure was PORC approved and appeared to be adequate.
The speci-
fic work observed involved the preparations and determination of
proper Raychem packages on cables 2-PM-1626-IV and 2-PM-1741-IV.
This specific work was authorized under WR B217558 and B217559
respectively.
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The inspector witnessed a portion of work plan 12195, AFW system
c.
Cavitating Venturi Post Maintenance Test on MDAFW pump 2A-A.
This
post maintenance test (PMT) plan was being performed because the
cavitating venturi associated with MDAFW pump 2A-A had been replaced
by MR A-522793.
The PMT consists of three parts; performed with the
steam generator (SG) pressure less than 100 psig, with SG pressure at
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approximately 385 psig and finally with SG pressure at 1005 psig.
The inspector witnessed the performance of the first part of the work
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plan.
No discrepancies were identified.
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d.
Portions of design change request (DCR) 2259 were reviewed in order
to resolve URI 327,328/86-71-05.
DCR 2259 was written to change the
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sequencing time of the containment spray (CS) pumps and other loads
when a loss of offsite power is followed by a need to commence con-
tainment spray.
The closure of this particular issue is consid-
ered by the NRC and TVA to be necessary prior to the startup of
either unit.
A discussion was held with department of nuclear
engineering (DNE) and Sequoyah nuclear plant compliance personnel to
resolve whether certain field generated changes to engineering change
notice (ECN) documents were adequately accomplished.
The following
documents were reviewed:
ECN L6715
DCR 2259
Field Change Request (FCR) 4873
Nuclear Engineering Procedure (NEP) 6.1, Variances / Expansions
,
Sequoyah Engineering Procedure (SQEP) AI-11A, FCR Handling
Sequoyah Administrative Instruction (AI)-19, Part IV, Plant
Modifications After Licensing
Work Plan 12227
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After a discussion with the licensee, it was determined that:
(1) The original DCR was not reviewed during the review and approval
process for FCR 4873.
Attachment 2, of NEP-6.1, which is used
during the initial ECN review and approval process, was not used
by either the design engineer or the review engineer to consider
whether FCR 4873 affected certain aspects of ECN L6715.
(2) It is not a normal practice for DNE engineers to review the
affected work plan or other issued FCRs prior to approving a
specific FCR.
(3) The reviews of WP 12227 and FCR 4873 conducted under AI-19 and
SQEP-AI-11A incorrectly determined that FCR 4873 was within the
scope of ECN L6715.
FCR 4873 addressed the removal of admin-
istrative control markings (hold clouds) and electrical circuits
within the hold clouds on certain as-constructed electrical
drawings.
These circuits and hold . clouds were addressed by
another ECN.
ECN L6715 did not address the addition or deletion
of the hold clouds on the affected drawings.
Therefore,
FCR 4873 was not within the original scope of ECN L6715, yet it
was reviewed with respect to ECN L6715.
The FCR was not
reviewed with respect to the ECN to which it really applied, and
consequently, an inadequate review was done prior to implement-
ing the FCR.
10 CFR 50 Appendix B, Criterion III, states that design changes,
including field changes, shall be subject to design control measures
commensurate with those applied to the original design.
NEP-6.1,
states that changes to any design documents which deviate from the
approved scope of work will be evaluated against the original work
scope.
As described above, FCR 4873 was not reviewed and approved
subject to design control measures commensurate with those applied to
the original design.
This is a violation VIO 327,328/87-02-01.
Three additional issues were identified during the review.
(1) SQEP-AI-11A Attachment A sections 12 through 14 were not com-
pleted when the inspector reviewed this completed FCR.
(2)
It was not clear what type of technical review the assigned
approving technical supervisor completed prior to approving this
FCR.
It appeared that the supervisory (DNE management) review
was administrative in nature and did not consider the technical
aspects of the FCR.
(3) The drawings reviewed and approved by the plant oaerations
review committee (PORC) were marked up by modificat ons person-
nel and were not the drawings approved by the DNE engineers.
The drawing revision numbers on the SQEP-AI-11A document did not
match the drawing revision numbers on the AI-19 document.
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this case the drawing changes were found to be the same.
However, the FCR process did not ensure the documents were the
same.
These issues will be resolved during a review of the licensee's
corrective action for VIO 327,328/87-02-01.
8.
Licensee Event Report (LER) Followup (92700)
The following LERs were reviewed and closed.
The inspector verified that:
reporting requirements had been met; causes had been identified; correc-
tive actions appeared appropriate; generic applicability had been consid-
ered; the LER forms were complete; the licensee had reviewed the event; no
unreviewed safety questions were involved; and no violations of regula-
tions or TS conditions had been identified.
LERs Unit 1
327/84-070 - Failure of Containment Isolation Valves, due to Sediment
Buildup on the Valve Stems.
During the performance of the weekly portion of SI-3, " Daily, Weekly,
and Monthly Logs," the ice condenser glycol system containment
isolation valves failed to stroke to the fully closed position on
demand.
The obstruction was a sediment buildup on the stem of the
valves.
The stems were cleaned and polished and the 0-rings and
diaphragms replaced.
The licensee was unable to determine the source
of the buildup.
There have been no indications of any similar
buildup on these or other valves on the glycol system prior to the
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event or to date.
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The inspector determined through
a records review and interview
process, that the event had been discussed with maintenance person-
nel.
The LER states that the results of a further investigation will
be supplied in a supplemental report.
To date this report has not
been issued.
The inspector discussed this discrepancy with licensee
personnel and the report will be issued in the near future.
The
publishing of this supplemental report will be tracked as IFI
327,328/87-02-06.
327/84-42 -
Incomplete Radiation Monitor Modification
327/86-52 -
Administrative Control of High Radiation Areas
LERs Unit 2
328/83-129 -
Loop 3 & 4 Main Steam Line Isolation Valves Would Not Close
Due to Excessive Heating During Operation.
The licensee determined that the valve packing had dried out during
operation.
Subsequent cooling during the refueling outage had caused
the packing to harden around the stem.
This prevented the valve from
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closing properly.
The valve stems were lubricated and the valves
tested satisfactorily.
Licensee review of past events showed that
the valves had been tested during the previous outage.
On the basis
of these tests the licensee determined that no further work was
required on the valves.
The valves are now tested and lubricated
every refueling outage regardless of the valve stroke-time.
This
recommendation was implemented in the Sequoyah preventative
maintenance program.
The inspector reviewed documentation available
to ensure that the problem has not occurred since that time.
This
item is closed.
328/86-08 -
Inadvertent Phase A Isolation As a Result of Personnel
Error In Failing to Follow Procedures.
Events described in
this LER contributed to the issuance of Violation 327,328/
87-02-02, described in Paragraph 9.
This item is closed.
328/86-09 -
Inadvertent Control Room Isolation Resulting from Deficient
Workplan.
Events described in this LER contributed to the
issuance of Violation 327,328/87-02-02, described in
paragraph 9.
This item is closed.
328/86-10 -
Inadvertent Phase B Isolation Occurred During Testing.
Events described in this LER contributed to the issuance of
Violation 327,328/87-02-02, described in Paragraph 9.
This
item is closed.
328/86-11 -
Inadvertent Containment Ventilation Isolation From Electro-
magnetic Interference On A Radiation Monitor
9.
Event Followup (93702, 62703)
During the inspection period covered by Inspection Report 327,328/
a.
86-71, three engineered safety features (ESF) initiations occurred
that were identified as URI 327,328/86-71-06.
The three events have
been determined to constitute a violation of TS 6.8.1.
(1) On November 30, 1986, a containment ventilation isolation (CVI)
was initiated when an instrument mechanic (IM) failed to follow
IMI-99, RT-106A.2, Response Time Test for Containment Lower Rad
Monitor 2-RM-90-106A.
The procedure stipulates that the IM is
required to have the operator place the appropriate channel
block switch in the " block" position prior to inserting a test
lead.
The IM inserted the test leads without having the block
switch in the correct position and a CVI resulted.
(2) On December 1,1986, a control room isolation (CRI) occurred
when a deficient workplan (WP) was used to perform a test on CRI
handswitch 2-HS-31A-7A.
The workplan was in error in that the
test procedure in WP 12268 required that the test personnel lift
one wire at the handswitch.
Subsequent review of wiring diagram
45N631-2 showed that additional parallel wire should have been
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specified in the procedure.
When the handswitch was placed in
the " actuate" position, a CRI occurred.
Inclusion of the second
wire in the procedure would have prevented this event.
(3) On December 3,1986, a containment phase "B" isolation occurred
when an inadequate procedure was used to perform instrument
maintenance instruction IMI-99, " Reactor Protection System
Response Time Test for Containment Pressure Channel III,
RT-16.7."
The procedure requires that only the train being
tested be placed in " test" prior to tripping the high-high
containment pressure bistables.
The test did not indicate that
the bistables feed both trains of the reactor protection system.
A phase "B" isolation was initiated by the train not in " test"
when the test was initiated.
TS 6.8.1 requires that written procedures be established, implemented
and maintained for surveillance and test activities of safety-related
equipment.
Contrary to this requirement, the licensee failed to
properly implement the procedure in instance 9.a.(1) above, and
failed to adequately establish and maintain procedures in instances
9.a.(2) and 9.a.(3).
These three examples constitute violation VIO
327/328-87-02-02.
b.
On January 6, 1987, the plant was notified by the division of nuclear
engineering (DNE) tnat a problem existed with the trip settings for
certain safety related breakers.
The significant condition report
(SCR) SQNEEB86124 R0, "SQN Auxiliary Power System class 1E Equipment
Ampacity Study," stated that the calculated maximum loading that
could occur during a loss of coolant accident (LOCA) exceeds the
feeder breaker continuous current setpoints for the control and
auxiliary building ventilation boards 1Al-A, 2Al-A and 2B1-B.
These ventilation boards supply safety related loads such as Emer-
gency Gas Treatment System (EGTS), Auxiliary Building Gas Treatment
System (ABGTS) and control room emergency ventilation fans as well as
various room coolers.
At approximately 4:00 p.m. (EST), the plant
determined that there was not an immediate operability problem
because certain (ESF) signals were blocked in the present mode (cold
shutdown
mode 5).
If the plant were at power, the potential for
maximum LOCA loading would cause the boards to be inoperable.
For
example, board 1Al-A could have a continuous loading of up to 444
amps following a LOCA.
The feeder breaker from the 480V shutdown
board to the vent board 1Al-A is set at 392 amps per the setting
sheets and other plant drawings.
Because the maximum expected
loading on these boards is within the capacities of these breakers,
the plant resolution will be to reset the breakers
to a higher
amperage capacity and change the drawings.
The inspector questioned why the licensee had not discovered this
when the vent boards are tested on a 18 month interval in SI-26.1A,
-26.1B, -26.2A and -26.28, " Loss of Offsite Power with Safety
e
12
Injection - D/G 1A-A,1B-B, 2A-A and 2B-B," respectively.
These
tests, however, do not ensure that cyclic loads (e.g. , room coolers
that cycle on temperature) are running.
Therefore, the maximum LOCA
loading was never experienced.
Resolution of this issue is expected
to be part of the design basis verification program.
This item will be reviewed under URI 327,328/86-60-10.
c.
During the inspection period the licensee began a program to inspect
the feedwater lines in response to the feedwater break event at Surry
Power Station.
d.
During the inspection period two separate spills of reactor coolant
system (RCS) fluid from open steam generator (SG) manways occurred.
Unit 1 SGs are presently undergoing heat treatment of the first two
rows of tubes to relieve stresses induced in manufacturing.
This
necessitates draining down the RCS to a level Lelow the SG primary
nozzles to allow robot aided access to the tube sheets.
A non-
watertight cover is placed over each RCS nozzle to prevent tools and
debris from entering the RCS loops.
recirculation and the primary access manways are opened.
Proper
water level in the RCS is maintained by use of an installed sight-
glass monitored by a television camera and a TV monitor in the
control room.
The first event occurred on January 28, 1987, and was caused by a
partially plugged sightglass.
The operators in the control room
noticed water level above the maximum allowed and began lowering
level by reducing charging pump flow. Water level in the sightglass
did not respond, and the operators continued to decrease charging
flow rate.
The residual heat removal (RHR) pump amperage began to
fluctuate, indicating pump cavitation.
The operators stop
pump and entered abnormal operating instruction (A01)-14, ped the
Loss of
RHR Cooling."
The operators began to raise water level to restore
adequate pump suction.
This caused actual water level in the RCS to
be increased until the SG channel heads began to fill and water
spilled out the manway.
The operators were alerted and stopped the
evolution.
After blowing down the sightglass, level indication was
restored and proper level attained.
The RHR pumps were then started
and RHR cooling restored.
Approximately 500 gallons of RCS water was
spilled.
The second event occurred on February 1,1987.
system were in the same configuration as in the first event.
The
operators were attempting to perform SI-166.3, " Stroke Time Testing
of FCV-63-1," which is the isolation valve for the normal RHR suction
from the refueling water storage tank (RWST).
When the valve was
opened, water from the RWST began to fill the RCS through the RHR
suction line connected to RCS Loop D, and then spilled from the open
SG manways.
The two series valves in the RCS suction line had not
been shut by the operator prior to opening the RWST suction. With
.
13
the RCS open to atmosphere, the elevation head in the RWST was enough
to cause flow back to the RCS.
The operator could not reclose
FCV-63-1 until the open limit was reached (approximately 25 seconds
After the valve was reclosed, water continued to spill
from the manway.
The operators suspected that valve FCV-63-1 had not
been fully shut by the motor operator, and sent an auxiliary operator
to the valve to manually shut it.
The auxiliary operator reported
that the local handwheel was turned approximately 10 turns in the
shut direction before valve travel stopped.
Total handwheel travel
is approximately 450 turns.
The rapid filling of the RCS displaced 6
of the 8 SG nozzle covers.
Total water spilled was estimated at 3000
to 4000 gallons.
Neither event resulted in personnel contamination.
These issues will be tracked as URI 327,328/87-02-07.
10.
Inspector Followup Items (93701)
Inspector followup Items (IFIs) are matters of concern to the inspector
which are documented and tracked in inspection reports to allow further
review and evaluation by the inspector.
The following IFIs have been
reviewed and evaluated by the inspector.
The inspector has either
resolved the concern identified, determined that the licensee has per-
formed adequately in the area, and/or determined that actions taken by
the licensee have resolved the concern.
(Closed) IFI 327,328/86-31-06, Moisture Entrained in Sense Lines Could
Cause Stress Corrosion Cracking in Bellows of Containment Sump Level
System.
(Closed) IFI 328/84-38-05, Procedure to Address Early Implementation of
Emergency Plan on Inadvertent Safety Injection.
(Closed) IFI 327,328/86-28-03, Component Cooling System Inleakage.
Increasing activity levels in the component cooling system (CCS) have been
observed since early 1985.
Efforts to locate the source of the leak were
not successful until the most recent investigation.
The location of the
leak has been identified as the 2B containment spray (CS) pump seal water
heat exchanger, which was found to have a sheared 0 ring at the inboard
seal.
This 0-ring provides the barrier between the CCS in the seal water
heat exchanger and the fluid in the pump casing.
The leak was found by
comparing cobalt 60/58 ratios from each component served by the CCS
system.
The unit 2 RCS was determined as the source of the radioactive
fluid.
The CS seal water heat exchangers were initially discounted as
possible leak points since the CS pumps are lined up to take suction from
the refueling water storage tank (RWST), and not from the RCS.
Local
samples at the Unit 2 Train B CS pump were conclusive.
The CS pump suction has the capability of switching to the ECCS recircu-
lation sump in the post-LOCA scenario.
This sump suction is by way of the
line common to the RHR Pumps.
Isolation from the sump and RHR system is
through a single valve, 2-FCV-72-20.
When this valve is opened (or leaks
by) the CS pump suction is exposed to RHR suction pressure. With the RHR
system on RCS recirculation, the pressure at this point is approximately
_
-
.
14
100 psig.
The leak was intermittent, which complicated the search.
It
was later determined that leakage only occurred when CCS pressure dropped
below 100 psig, such as when major components were placed in service. The
licensee is presently disassembling the other three CS pumps to inspect
'.
their seals.
IFI 86-28-03 is closed.
(Closed) IFI 327,328/85-27-03, Modifications to Che.nical Volume Control
System (CVCS)
(Closed) IFI 327,328/85-27-05, Flow Induced Vibrations on CVCS
(Closed) IFI 327,328/86-15-04, CVCS line Tee
.(Closed) IFI 327,328/86-20-05, Word Processing Errors
(Closed)IFI 327,328/86-42-04, Vendor Manuals
11. Review of Employee Concern Element Reports (TI251574)
Based on a review the following element reports the following general
areas of interest were identified:
a.
Many of the reports did not completely address the concern as
expressed by the individual,
b.
In some reports the scope and depth of the investigation appeared not
to be acceptable,
c.
There appeared to be a disconnection between some of the element
reports and the document files.
d.
Corrective action in some cases appeared not to resolve the concern
and in others the indicated corrective action did not include ancil-
lary issues.
The following employee concern element reports were reviewed during this
inspection period:
OP308.06 SON
OP307.02 SQN
C015101 SQN
OP308.05 SQN
OP301.05 SQN
OP306.01 SQN
MAS-86-001
OP310.02 SQN
C015109 SQN
OP301.11 SON
'
OP301.07 SQN
OP313.07 SQN
OP301.12 SQN
OP307.08 SQN
OP308.01 SQN
OP309.05 SQN
C015102 SQN
OP313.02 SQN
C01509 SQN
OP307.06 SQN
OP305.01 SQN
OP313.09 SQN
C015105 SQN
OP301.01 SQN
OP301.11 SQN
OP309.01 SQN
-
. _ . . _ . - . . - . -
-
. - - , - - - . _ . .
_ _ ~ - . . -
-_
.
-.
--
-
"
.
15
OP301.08 SQN
OP307.11 SQN
OP313.07 SQN
OP310.03 SQN
OP310.01.SQN
12. Experience Review (93702)
A potentially reportable occurrence report (PRO) 1-87-013, was issued
January 10, 1987.
The report described a condition where water could be
lost from the ECCS recirculation sump area in a post-LOCA event.
The
postulated loss was described as through leak paths at the air-return fans
on the upper containment deck level, and/or through a failed divider deck
seal.
The licensee has made a preliminary evaluation to determine the
amount of water that could be lost from the ECCS path (ref. SCR
SQNNEB8623R0).
The preliminary result of this analysis indicates that
water in the sump could fall below the sump swapover elevation (+13.7
ft.), but would be above the minimum level to prevent pump cavitation
(+8.0).
A second issue involved the possible failure of one of the two air-return
fans during spraydown of the area above the upper deck.
Fan A-A, located
above accumulator room 3, has the potential for having water from the
post-LOCA spraydown pooling above the open fan inlet.
The fan has been
estimated to have a water flow rate of approximately 650 gpm. This value
is in excess of the 50 gpm value guaranteed by the manufacturer.
personnel have been in informal contact with Duke Power personnel, alert-
ing them to the possibility of an arrangement that could allow water to
pool above their fans.
Resolution of SCR SQNNEB8623R0 is IFI 327,328/
87-02-08.
13. ColdWeatherPreparations(71714)
On January 16, 1987, the inspectors observed work in progress per general
operating instruction (G0I) - 6H, Apparatus Operations - Freeze Protec-
tion.
During this performance the inspector noted several discrepancies.
Step 17.A. and 18. A. both list a number of thermostats to be checked.
These steps require that the following be checked:
"All Circuits greater
than or equal to 75 degrees F - Check thermostats Set at 75 degrees F."
,
The auxiliary unit operator (AV0) conducting the procedure told the
l.
inspector that temperature indications did not exist for these circuits.
Additionally, the thermostats were set ranging from 40-150 degrees F.
They were left at 75-150 degrees F even though the procedure stated that
they were to be set at 75 degrees.
The response received was that the
'
procedure simply meant to make sure the thermostats were set greater than
75 degrees.
The inspector discussed the events with other AU0s and the Unit Operator
on shift.
The responses ranged from the fact that some thermostats were
not good enough to hold the circuit at or above 75 degrees without setting
!
them above that point, to the statement that it really did not matter as
long as the circuit remained above freezing.
i
,
_ . . . . _ __. _ - _ _,_ ,_ ._ _ ~._,_._ _ .., _. _ _ _ . . - - _ _ . _ _ _ . _ . _ _ . _ _ _ . _ - _ _ _ _ _
_
.
16
The inspector found that circuit thermostats for circuits 369P, 367P, 369S
and 367S were not identified in the panels designated in the procedure.
The AU0 accompanied by the inspector could not locate these thermostats.
The inspector was told by other AU0s that they did not exist.
One AVO
informed the inspector that he knew the thermostats did not exist, but
believed that the problem was identified on a WR.
No WR was identified.
Another AVO stated that he had in the past circled the errors and stated
that they were wrong. The inspector reviewed past performances of the G01
performed on December 19, 1987, January 2 and 9, 1987, and found that all
steps were signed off except on January 2, 1987. During this performance
circuits 365S, 363S, 369P and 367P were scribed out with no initials or
justifications.
The inspector went on a second tour with a different AVO on January 27,
1987.
During this tour the AVO showed the inspector the appropriate way
to determine the circuit temperatures.
Also using hand written cabinet
labeling he determined that the circuits were supplied by thermostats that
controlled more than one circuit.
The inspector verified this to be the
case by reviewing the circuit drawings with instrument maintenance person-
nel.
Neither of these points had been addressed by the original perfor-
mance or by the procedure.
The inspector reviewed the January 23, 1987
performance and found that circuits 365S, 3635, 369P, 367P and 367S were
all circled and annotated, "not found or non-existant."
TS 6.8.1 requires that written procedures be established, implemented and
maintained covering activities affecting quality.
G01-6H, Freeze Protec-
tion Checklist, is one such procedure.
Contrary to the above, on January 16, 1987, G01-6H was performed without
verifying the temperatures of circuits as required.
In addition, the
performance did not check certain thermostats as required because they
were not found.
Finally, circuit thermostats were not set to 75 degrees
per procedure specifications.
This is a violation of TS and shall be
identified as another example of Violation 327,328/87-02-02.
In addition, the inspector determined several labeling discrepancies which
included incorrect panel numbers, hand written labeling, and incorrect
switch labeling.
These cases were identified to the licensee personnel,
and will continue to be followed as housekeeping violation VIO 327,328/
85-32-02.
14.
Sustained Control Room and Plant Obcervation (71715)
During the inspection period, meetings were held with the Sequoyah opera-
tional readiness restart group for the purpose of coo-dinating the NRC
restart inspection effort.
The licensee presented, for information, two
operational readiness procedures; standard practice SQA-190, "Sequoyah
Activities List Restart Item Disposition," and standard practice SQA-191,
" Evaluation of Operational Readiness Prior To Plant Restart."
These
procedures establish TVA's restart criteria and describe the closecut
process including independent reviews.
The Sequoyah restart criteria,
. - _ _
- _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _
I
.
(
17
!
Appendix A to SQA-190, is provided below for information:
a.
The item identifies a specific deficiency which has significant
probability of leading to the inoperability of a system required for
startup or operation by the appropriate TS.
b.
The item identifies a programmatic deficiency which has a high
probability of causing or has caused a specific deficiency which
meets No. 1 above.
NOTE:
To assist in the determination of required for restart
relative to TS as in criteria No. I and 2 above, an affir-
mative answer to any or the following questions requires
consideration of the item for restart based on TS
requirements.
(1) Does the item directly and adversely affect safety-
related equipment function, performance, reliability,
or response time?
(2) Does the item indirectly and adversely affect safety-
related equipment power supply, air supply, cooling,
lubrication, or ventilation?
(3) Does the item adversely affect primary containment
integrity?
(4) Does the item adversely affect secondary containment
integrity?
(5) Does the item adversely affect control room
habitability?
(6) Does the item adversely affect systems used to process
radioactive waste?
(7) Does the item adversely affect fire protection or fire
'
loads?
(8) Does the item adversely affect the ability of a system
or component to meet its safety function during a
l
design basis event by impacting the seismic analysis,
i
single failure criteria, separation criteria, high
i
energy line break assumptions, or equipment qualifica-
tion?
(9) Are the programs such as radiological health, secur-
ity, radiological emergency preparedness, or quality
assurance which are necessary for safe conduct of
operations of the plant adversely affected?
_ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
~
,
"
.
18
(10) If not corrected prior to restart, could it lead to an
uncontrolled release or spread of radioactive contarri-
nation beyond the regulated area?
c.
The item identifies a specific deficiency that results in a failure
to comply with NRC regulations and no variance has been approved by
NRC.
d.
TVA has committed to NRC to complete the item prior to restart.
e.
The item identifies a specific deficiency which has a significant
probability of leading to a personal injury during plant operation.
f.
The item identifies a specific condition which has a forced outage
risk (probability X outage length) during the next cycle in excess of
the critical path time to correct the condition prior to restart.
The above restart criteria was evaluated against the criteria established
in volume 2 of the revised Sequoyah nuclear performance plan (SNPP). The
criteria listed in SQA-190 was the same as that in the SNPP with the
exception of items e and f above.
Item e addresses the personnel safety
aspect, where item f addresses schedule considerations.
Item f appears to
place a high degree of importance on schedule; however, when considered
along with the other established criteria it appears to be acceptable.
Additionally, SQA-190 has attempted to define the degree of significar.ce
of an item, where the SNPP implies the item must lead to inoperability of
equipment or systems before it becomes a restart item.
Implementation of
the above criteria will be evaluated during subsequent inspections.
15. Liquid and Gaseous Effluents (90713)
The report listed below was reviewed by Regional inspectors to verify
reporting requirements of technical content, data collection, acceptance
criteria, and handling of deficiencies noted.
Report reviewed was as follows:
Sequoyah Effluent and Waste Disposal Semi-annual Report,1st Half 86,
dated August 29, 1986
No violations or deviations were identified.
16.
IE Bulletins (92703)
(Closed) 327,328/86-BU-03, Potential Failure of Multiple ECCS Pumps Due to
Single Failure of Air Operated Valve in Minimum Flow Recirculation Line.
A review of the licensee response to IEB 86-03 (dated November 14,1986)
was conducted in the region.
In their response, the licensee stated that
.
.
19
the single failure vulnerability discussed in the bulletin did not exist
at Sequoyah due to the following:
The active ECCS consists of the safety injection system (SIS),
residual heat removal system (RHRS), and the centrifugal charging
portion of the chemical and volume control system.
The SIS has two safety injection pumps; each pump has a minimum flow
recirculation line connected to a common return line to the refueling
water storage tank (RWST).
A motor-operated flow control valve is
located in the recirculation line for each pump and in the common
return line to the RWST. The valve on the common return line to the
RWST is normally open and fails "as-is" and is remote-manual con-
trolled. Because it is the only isolation valve on the comon return
line from the safety injection pump discharge to the RWST (minimum
flow recirculation line), the design of the control circuit is such
that no spurious actuation will be able to energize the opening and
closing coils for the valve operator.
Emergency instructions call
for the valve to be closed before transferring SI pump suction to the
containment sump during recirculation mode of accident mitigation to
prevent transfer of radioactively contaminated water to the RWST. As
such, this system does not represent a concern in this area.
The RHRS has two trains and each train has its own separate and
redundant minimum flow recirculation line.
Each recirculation line
has its own normally-closed, fail "as-is," motor-operated globe
valve.
The control logic for each valve is identical.
With the
respective RHRS pump running and switches in the normal / auto posi-
tion, the valve will automatically open for flow below a prescribed
setpoint. The valve can also be opened remote manual.
Two centrifugal charging (CC) pumps share a comon minimum flow
recirculation line which has two motor-operated globe valves in
series.
These valves are normally open with the power removed to
satisfy 10 CFR 50 Appendix R requirements.
This item is closed.