ML20207D706
ML20207D706 | |
Person / Time | |
---|---|
Site: | Sequoyah ![]() |
Issue date: | 07/27/1988 |
From: | Jenison K, Mccoy F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II), NRC OFFICE OF SPECIAL PROJECTS |
To: | |
Shared Package | |
ML20207D541 | List: |
References | |
50-327-88-27, 50-328-88-27, GL-87-12, IEB-79-14, IEB-80-12, IEB-80-15, IEIN-86-081, IEIN-86-81, IEIN-87-062, IEIN-87-62, NUDOCS 8808150434 | |
Download: ML20207D706 (25) | |
See also: IR 05000327/1988027
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Report Nos.:
50-327/88-27, 50-328/88-27
Licensee: Tennessee thlley Authority
6N 38A Lookout Place
1101 Market Square
Chattanooga, TN 37402-2801
Docket Nos.:
50-327 and 59-328
License Nos.:
OPR-77 and DPR-79
Facility Name:
Sequoyah Units 1 and 2
Inspection Conducted: May 1, 1988 thru May 30 1988
Team Leaders:
k
Vf' 25,1106
K/M. /ep son, Senior Resident Inspector
Sete Signed
Team Members:
P. Harmon
G. Humphrey-
D. Loveless
W. Poert
Approved by:
W
~/!)788
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F. R. b)t:Co), Chief, Project Section 1
' 'Dat( Signed
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Division of TVA Projects
SUMMARY
Scope:
This routine, announced inspection was conducted on site in the areas
of:
operational safety verification; review of previous inspection
findings; followup of events; review of licensee identified items;
review of NRC Bulletins; and review of Inspector Fellowup Items.
Results: One violation was
identified - Violation 327,
328/88-27-01,
Inadequate Corrective Action for Improper Positive Reactivity Changes
(Paragraph 5).
4
No unresolved items or inspector follow-up items were identified.
BG00150434 800727
ADOCK 05000327
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REPORT DETAILS
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Persons' Contacted
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Licensee Employees
'H. Abercromb'ie, Site Director
J. Anthony, 0perations Group Supervisor
- R. Deecken, Maintenance Superintendent
~J'. Eynum, Assistant Manager of Nuclear Power
M. -Cooper,~ Compliance Licensing Supervisor
H. Elkins, Instrument Maintenance Group Manager
R. Fortenberry, Technical Support Supervisor
J. Hamilton, Quality Engineering Manager
.
M. Harding, licensing Group Manager
G. Kirk, Compliance Supervisor
'*J.
La Point, Deputy Site Director
- L. Martin, Site Quality Manager
R. Olson, Modifications Manager
R. Pierce, Mechanical Maintenance Supervisor
R. Prince, Radiological Control Superintendent
R. Rogers, Plant-Operations Review Staf f
S.. Smith, Plant Manager
,
J. Sullivan, Plant Operations Review Staff Supervisor
B. Willis, Operations and Engineering Superintendant
C. Whittemore., Licensing Engineer
'NRC Employee
F. McCoy
- Attended exit interview
2.
Exit Interview
'd
- The inspection scope and findings were summarized on June 8, 1988, with
those persons indicated in paragraph 1.
The Startup Manager 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.
NOTE:
A list of abbreviations used in this report is contained in
paragraph 9.
Inspection Findings:
One violation was identified, paragraph 5.
No unresolved items or inspector follow-up it'.ms were identified.
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3.
Operational Safety Verification (71707) Units 1 and 2
The inspectors reviewed applicable logs, including the shift logs, night
order book, clearance hold order- book, configuration log, and TACF log;
conducted discussions with control room operators; reviewed LER and PRO
documents; and confirmed the adequacy of corrective actions.
The
inspectors verified compliance with TS LCOs and that WRs had been
submitted as required including follow-up activities and prioritization of
work.
Tours of the auxiliary, control, and turbine buildings were conducted to
observe plant equipment conditions and to inspect corrective actions.
No violations or deviations were identified
4.
Licensee Action on Previous Enforcement Matters (92702)
(Closed) VIO 327, 328/87-61-01, Failure to Accomplish an Activity
Affecting Quality in Accordance with a Prescribed Procedure.
Eight of twenty-five licensed SR0s and R0s failed to complete required
procedure reviews within the prescribed time stipulated in OSLA-1, Nuclear
Generating Plant Operations Training Program Letter. OSLA-1 required
documentation of subsequent action changes to emergency (EI) and abnormal
operating procedures (AOI) within five shift days for snift operators,
however attachment XI-C which records the individual's review of procedure
changes, specified within thirty days. As part of the corrective action
for this violation, and to resolve the inconsistency, the licensee revised
OSLA-1 to agree with Attachment XI-C, requiring review within 30 days.
Changes to immediate action steps of the Els and A01s require review prior
to assuming the shift.
Additionally,- the
licensee has revised
Administrative Instruction (AI)-5, Routine Shift Work, to include a
requirement for all licensed operations personnel and shif t technical
advisors to review and sign for immediate action changes to EIs and A0Is
prior to assuming the shift.
The licensee's actions appear to be
adequate.
This item is closed.
(Closed) VIO 327, 328/87-66-01, Licensee Failed to Adequately Establish,
Implement, and Maintain Procedures for Configuration Control.
This event involved the generation and control of plant system configu-
ration control records. OSLA-58, Maintaining Cognizance of Operational
Status
Configuration
Control,
failed to adequately specify when
configuration control records should start (records only required af ter
check lists completed), and failed to specify an appropriate method for
deviating from SOI checklists (deviations not considered a procedural
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change). The licensee failed to adequately implement OSLA-58 requirements
for the use of working copies of SOI checklists (transfer of all pertinent
data), failed to implement requirements for recording deviations (date not
recorded next to initials), failed to maintain configuration control after
checklists were completed (documented configuration did not agree with
actual configuration) and failed to process revisions to SOI checklists
(log entries cleared without re performing the section of checklist that
had been revised).
The inspector has reviewed AI-58 Maintaining Cognizance of Operational
Status-Configuration Status Control (a conversion of OSLA-58); AI-2,
Authorities and Responsibilities for Safe Operation and Shutdown; AI-4,
Preparation, Revics, Approval and Use of Plant Instructions; AI-30,
,
Nuclear Plant Method of Operation, IR 327,328/87-66, the response to NRC
inspection reports 327,328/87-66 (K. P. Barr/S. A. White) letter dated
December 21, 1987; IR 327,328/88-06; Response to System Alignment
Verification for Unit 2 Heatup Violations, and OSLA-107, System Operating
Instruction Review and Verification Checklist.
The new AI-58 procedure,
as amended by the work ticket process, appears to provide adequate
correctin, of the examples of Violation 87-66-01.
The assignment of
managen.ent personnel to directly supervise the system alignment program in
conjunction with the training given to personnel involved in this program
also appear to be adequate. Finally, QA audit interaction with operations
,
on an extended basis has botn increased operator and management attention
in this area and identified a sharp reduction in the number of
configuration related errors. Violations 327,328/88-06-01, 88-06-02, and
88-06-03 also address the subject of configuration status in a more
generic sense. Therefore, because the specifics of this violation have
been corrected, the generic issues aisc appear to be corrected and the
ge ne r.i c issues will be reviewed again on the closure of the other
violations mentioned.
Violation 327,328/88-66-01 is closed.
(0 pen) VIO 327,328/87-71-01, Failure to Implement Adequate Design Control.
This item involved the licensee's failure, due to a misinterpretation of
GDC-56, to install caps on the ends of test connection lines for
instruments30-46B, 30-478, and 30-48B, thereby not providing the required
containment isolation capability.
The licensee's actions regarding this item were as follows:
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DCN X00028-B was issued to add the required caps.
The work was
completed on Unit 2 in accordance with WP 12635.
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The licensee has committed to install the required caps in Unit 1
prior to entry into mode 4.
This item remains open until corrective actions are complete.
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(0 pen) VIO 327,328/87-73-05, Failure to Perform Adequate .10 CFR 50.59
Safety Evaluations for Modifications to the Facility. Which Involved
Compensatory Measures.
As previously reported in Inspection Report 87-76, the inspectr.r has
reviewed the specific USQDs pertaining to this issue and determ',vJ them
to be adequate. TVA's response has been received, however a supplemental
response is due by June 6, 1988. Pending the receipt and review of this
supplemental response, this item remains open.
(Closed)' VIO
327,328/87-76-01,
Failure to Perform Adequate Post
Maintenance Testing
This issue involved an inadequate post maintenance test procedure that did
not test operation of a specific diesel generator voltage regulator under
rapid ioading/ rejection conditions.
The licensee's response to the violation (Gridley/NRC Document Control)-
dated May 4,
1988, (RIMS L44880504811) was reviewed and found to be
adequate.
The licensee has issued LER 327/87070, Revision 3,
that
addresses the generic concerns of this violation.
The licensee has
revised Standard Practice SQN 66, Post Maintenance Testing, to ensure that
testing requirements on a.iy voltage regulator maintenance work are
evaluated for each case until appropriate post maintenance tests are in
place. The licensee's corrective actions appear to be adequate and if
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implemented, will prevent recurrence.
Long term corrective action will be
tracked by LER 327/87070.
This Violation is closed.
(Closed) Unresolved. Item (URI) 327, 328/86-46-06, and IE Bulletin 80-12,
Decay Heat Removal (DHR) System Operability.
This issue involves the potential loss of RHR during Mode 5 operation.
Bulletin 80-12 directed the licensee to perform six reviews and report to
the commission regarding specific actions. The licensee responded on June
19, 1980 by letter from J. C. Ross to J. O'Reilly, indicating the results
of the licensee's reviews and analysis as required.
The licensee
indicated that procedures would be changed to include instructions on loss
of decay heat removal capability during the following modes of operation:
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The RCS partially drained, but before reactor vessel
head
de-tensioning.
The reactor vessel head de-tensioned, but still in place.
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The reactor vessel head removed, but before filling of the refueling
cavity.
A01-14, Loss of RHR S;.utdown Cooling, has been revised to include
alternate cooling for the modes identified above.
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NRC GL 87-12, Loss of RHR while the RCS is partially filled, is a
reiteration of the same basic material asBulletin 80-12, with more
emphasis on detail.
The: inspector reviewed A0I-14 and the conditions described above are
adequately addressed. GL-87-12 was reviewed as well as the licensee's
response (Rims L44 871002 805). GL 87-12 supersedesBulletin 80-12 and
URI 86-46-06, which was written to track some of the above menticned
conditions.
The licensee's actions appear to be adequate.
Therefore, URI 86-46-06, and Bulletin 80-12 are closed.
GL 87-12 will
remain open until the final corrective actions are complete.
(Closed) VIO 327/80-15-01, Piping Support Discrepancies.
NRC Inspection Report 327/80-15 reported drawing discrepancies and pipe
support discrepancies with several safety related supports which had been
previously signed off and inspected by TVA as satisfactory.
The TVA
response to the violation stated that specific discrepancies
had been
corrected and the licensee planned to conduct walkdown inspections and a
review of several hundred supports inside and outside containment.
This
inspection was to be done in conjunction with elements of NRC Bulletin 79-14, Seismic Analysis for As-Built Safety Related Piping Systems.
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TVA transmitted the results of the initial inspection by letter dated June
27, 1980.
The results indicated ". . . discrepancies found were randomly
distributed through several phases of the design, construction, and
operation process
TVA concluded that discrepancies would not
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prevent safe operation at low power; however, additional inspections would
be necessary to provide further assurance that the plant can be operated
at full power.
TVA later identified as part of the Design Baseline Verification Program
(DBVP) design problems with several thousand hangers which led to large
scale re-engineering and plant modification efforts.
The inspector reviewed the inspection report, the TVA acknowledgement and
the initial inspection summary.
The specific discrepancies associated
with Violation 80-15-01 were corrected.
Corrections of programmatic
deficiencies for Unit 2 were addressed under the NRC OBVP verification
inspections and found to be programmatically adequate.
Implementation
activities on Unit I are currently in progress for approximately 1300
hangers and Bulletin 79-14 remains open.
The generic issues associated
with the adequacy of Bulletin 79-14 will be addressed with the closure of
the bulletin.
This violation is closed.
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'(Closed)'URI 327, 328/87-30-06, EDG Circuit Cards.
On April 30,1987, .while trouble shooting erratic v'oltage swings on the-
1 A-A . EDG , the ' licensee determined that 'he voltage swing problem was
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associated with the voltage regular minimum-maximum cards.
The licensee,
with - concurrence and assistance of .the - vendor ~(Basler Electric),
disconnected the minimum-maximum circuit from the EDG voltage regulator
for all EDGs.
Subsequent calculations and operational testing conducted
by the licensee were submitted to the NRC's OSP for review.
OSP issued,
SER NUREG 1232. Volume 2, Part 1 on March 25, 1988, (Revised Preliminary
Report) which accepts the licensee's EDGs as adequate to support restart.
Portions of the SER are excerpted here to detail the boundaries of
acceptance and the licensee commitments:
In January 1988, TVA identified to the NRC data from surveillance
tests which raised significant questions about the operability of the
EDGs at Sequoyah.
These results were interpreted by TVA as
indicating both a possible aefect in the 2A-A generator's exciter
system and a more general problem in all generators in conforming
with voltage limits during loading as stated in RG'1.9.
A failed
component was replaced in the exciter system of the 1A generator
which corrected the first problem and lef t only' the more general
voltage problem.
A detailed review of the test data by the NRC
identified the following technical issues relevant to the second
problem:
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'The test results were worse than would be predicted by the
calculations methods used to model diesel generator performance.
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The test data, when extrapolated to post accident conditions,
showed that the diesel generators had less margin, in termt of
voltage behavior, than calculations had predicted.
Because of these issues, the NRC required TVA to undertake a major
analytic effort with the following objectives:
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To identify the reasons why calculations did not predict the
severity of the experimental results.
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To improve the calculational methods to provide greater
assurance that the calculational methods would conservatively
predict post accident behavior.
To quantify the margins available between diesel capability and
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specific electrical power system requirements in post accident
operation.
The specific margins in question were the following:
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minimum voltage during the loading sequence.
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diesel generator power rating (KVA and KW).
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motor performance (starting and stalling).
contractor performance _(pick up and drop out).
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motor operated valve performance (torque and timing).
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overcurrent protection misoperation (circuit breakers and
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control fuses).
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sequence timing error (overlap and maximum loading time).
These required analyses and calculations were submitted to the NRC on
March 1, 2, and 11, 1988.
Based on its review of the results of
these calculations the NRC reached the following conclusions:
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The most likely reason that the original calculations did not
bound the test data was that generator voltage did not stabilize
between successive steps in the loading sequence.
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The generator's inability to stabilize voltage was probably
caused by the use of a voltage regulator lacking the speed and
,
performance characteristics typical of those used in modern
nuclear applications.
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The most recent analysis methods used by -TVA bounded the
experimental data and predict the EDGs behavior in post accident
loading with adequate margin. However, this margin is less than
was expected when the plant was licensed.
The margin that remains is sufficient to assure safe operation
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of Sequoyah for restart and for the period of time until
corrective action is taken to improve the margin.
In the March 3,
1988, submittal, TVA committed to evaluate the
performance of the EDGs and implement corrective action . prior to
Unit I restart after the naxt Unit 1 refueling outage. This schedule
is acceptable to the NRC staff. The staff SER states that the TVA
calculations on which the staff's findings are based assume the
Sequoyah Unit 1 is in cold shutdown and must be revised to support
Unit i restart.
Further, the staff notes its reliance on TVA's
commitment to undertake,
after restart, a major review and
modification
effort
to
improve
performance
of
the
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regulator / exciter system."
Based on the above this URI is closed.
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5.
Licensee Event Report (LER) Follow-up (92700)
The following LERs were rev'.ewed and closed. The inspector verified that:
reporting requirements had been met; causes had been identified;
- corrective actions appeared appropriate; generic applicability had been
considered; the LER forms were-complete; the licensee had reviewed the
event; no unreviewed safety questions . were involved; and no further
violations of regulations or Technical Specification conditions had been
identified.
(Closed) LER 327/86038, Two EDG Starts Resulting from Personnel Error and
a Component Failure.
During performance of SI-60, 6.9 KV Unit Board Manual and Automatic
Transfer,
licensee personnel went to the wrong unit board to locate the
components specified in the SI. Operation of the identical components on
another unit board, resulted in inadvertent start of the EDGs.
Partial
cause of this event has been attributed to the close proximity of the Unit
Boards. The licensee's corrective action consisted of clearly identifying
the unit board separation with highly visible stripes and training
personnel to be aware of the unit board layout configuration.
A second
EDG start occurred while transferring Start Bus 28 from alternate to
normal supply.
Approximately two to three seconds after closure of
Breaker 1414, the breaker tripped causing a loss of voltage and starting
of the EDGs.
Investigation revealed that the contacts on the close-trip
switch were dirty.
The manual close-trip switch was replaced, and the
breaker was tested satisfactory and returned to service. While returning
the EDGs to normal following this inadvertent start, EDG 2A-A restarted.
Investigation revealed a loose fuse clip contact.
The fuse clip was
repaired, and the other EDG fuse clips were inspected and found to be
acceptable.
The licensee's correct;,6 ?ctions are adequate.
This LER is closed.
'(Closed) LER 327/87078, Revision 2,
Inadequate Procedure For Reactor
Coolant System Chemical Addition Resulted In Noncompliance With A TS
Action Statement.
This event report describes three occasions on Unit 1 and seven occasions
on Unit 2 when chemical additions to the RCS caused positive reactivity
changes in violation of TS requirements. The chemicals were flushed into
the system using up to 75 gallons of demineralized water which resulted in
a very small dilution of the RCS boron concentration.
The licensee has
calculated these decreases in boron concentration to be approximately four
parts per million (ppm) resulting in negligible positive reactivity
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changes. However, Action Statement (b) to TS 3.7.7 requires that in Modes
5 and 6, when the Control Room Emergency Ventilation System (CREVS) is
inoperable, all operations involving positive reactivity cheqges be
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suspended.
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The NRC inspector reviewed Revision I to the LER in late March .1988 and
identified the following three discrepancies:
The abstract and corrective action-sections of Revision 1 state that
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Technical Instruction (TI)-19, - Chemical Feed _ Controls, and- System
Operating Instruction- (S01)-62.5, Boric Acid Batching, Transferring -
and Storage System, had ~ been revised such that both procedures now
provide instructions for RCS chemical addition that will not cause an
RCS dilution.
In fact, both of these procedures had been revised
with temporary changes that expired on January 30 and February 15,-
1988, respectively-
Thus, when LER Revision 1 was issued on
.
February 25, 1988, neither of these . procedure changes remained in
effect.
In addition, two other station procedures that govern
chemical additions which could cause positive reactivity changes were
not addressed; SOI-62.2, Boron Concentration Control, and S0I-62.3,
Reactor Coolant Chemical Addition and Control.
The abstract and corrective action sections of LER Revision 1 state
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that "no planned activities, when either Unit 1 or Unit 2 is in
Mode 5 or 6, .will be performed which could potentially result in a
positive reactivity change if both trains of CREVS are inoperable."
However, there were no formal licensee controls or procedural
limitations in place to assure that such activities could not take
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picce.
The abstract and corrective action sections of the original LER
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contained a commitment to revise the TVA interpretation of the action
statement to TS 3.7.7 so that any planned evolution that could result
in a positive reactivity change is considered prohibited by IS. TVA
also committed to promulgate the details of this interpretation to
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licensed personnel by issuing a management directive by February 29,
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1988.
No management directive or other suitable notification had
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been issued to station personnel. Revisions 1 and 2 to the LER have
deleted references to these commitments without explanation and they
have been closed without action on the licensee's Corporate
Commitment Tracking System.
CAQR SQP 871660 which addressed these events was closed on
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January 23, 1988, specifying that a root cause analysis / recurrence
control was not' required.
In addition, the CAQR was erroneously
closed based on a temporary procedure change (expiring January 30,
1988).
The CAQR disposition stated that a procedure was developed
that would ensure no dilution would occur while under LC0 3.7.7
(TI-19 Revised) for addition to RCS.
The plant Operations Review
Staff was apparently not knowledgeable of the corrective action
status or actual actions accomplished at the time Revision 1 to this
LER was prepared and issued.
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The above concerns were discussed in detail with responsible licensee
personnel on April 1,-1988. No indication was made at this meeting that
there was any current intent by TVA to make additional reviews or take
further corrective action.
The licensee subsequently stated (late May
1988) that they had intended to implement additional long term corrective
actions in March, .However, the
licensee could 'not provide any
documentation (meeting notes, memos, log entries, commitments, telecon
notes, etc.) of this intent
that any additional reviews or actions were
assigned or in process. The failure to take corrective action to prevent
recurrence is a violation of 10 CFR 50, Appendix B, Criterion XVI and is
identified as Violation 327, 328/88-27-01.
As a result of the above identified concerns TVA has taken additional-
actions. Revision 2 to the LER was issued which provides more detail on
the positive reactivity change events and committed to permanent revisions
to TI-19, 501-62.2 and S01-62.3 as long-term corrective actions.
These
procedure changes have been reviewed by the NRC inspector and it appears
the procedural corrective action is adequate.
On April 12, 1988, TS
interpretation 111, was issued addressing plant evolutions that could
result in positive reactivity changes. The licensee's corrective actions
appear to be adequate.
This LER is closed.
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(Closed) LER 327/88008, Personnel not Properly Implementing Approved
Administrative Procedures Resulting in Inappropriately Exiting a TS Action
Statement on Radiation Monitoring.
This event involved operations personnel declaring radiation' monitors
(0-RM-90-134 and 141) operable prior to completion of the instrumentation
post modification test PMT-206, Radiation Monitoring Sample Flow Alarm
Function Test and Calibration.
The inspector reviewed the LER and the following corrective actions:
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The immediate corrective actions taken were to enter TS LC0 3.3.3.9
action statements and to resume taking grab samples for analysis.
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Operations personnel on shift were directed to review all PMT
documentation as verification before returning equipment to an
operable status.
This statement was also included as a change to
AI-30, Nuclear Plant Conduct of Operations.
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The personnel involved in the event have been counseled.
The licensee actions appear to be adequate.
This LER is closed.
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(Closed) LER 327/88011, An Inadequate Determination of the Effect of
Routine Diesel Testing Resulted in Both Trains of the Control Room
Emergency Ventilation System Being Inoperable.
This event -involved a routine maintenance' operation on EDG 1A-A that
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administratively made the CREVS train A inoperable while the train B CREVS
was declared inoperable to allow replacement of fire detectors in the
ductwork.
This resulted in both trains of CREVS being inoperable which
required entry into LC0 3.0.5.
This condition had existed from 11:40 p.m.
February 15, 1988 to 12:30 a.m.
February 16, 1988- (approximately 50
minutes) before the shif t supervisor determined that LC0 3.0.3 was in
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effect.
The inspector reviewed LER 327/88011, LC0 3.0.3, LCO 3.0.5 and S0182.1.
The licensee has established a WCG, to help the control room shif t crew
determine the impact of work activities on the plant and has revised the
50I-82 series of instructions to include warnings that certain steps will
cause the EDG to become inoperative.
These factors should reduce the
chances for similar events.
The licensee's actions appear to be
acceptable.
This LER is closed.
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(Closed) LER 327/88012, ESF Actuation.
CVI Caused by Electromagnetic
Interference (EMI).
This event involved an EMI actuated CVI. This problem had been identified
previously and a task force had recommended modifications to eliminate or
at leas' reduce spurious CVIs caused by radiation monitor (RM) circuit
induced EMI.
This event occurred before the modification ~ were installed.
s
The inspector has reviewed LER 327/88012, WP 7343-02 and WP 7344-02, ECN
L7343 and ECN L7344. The ECNs and WPs are the result of the task force
recommendations and they have been field completed. This issue was also
addressed as a CAQR which was closed in NRC Inspection Report
327,328/88-19. The licensee's actions appear to be adequate.
This LER is closed.
(Closed)
Radiation Monitor Technical
Specification
Surveillance Requirement Omitted From Surveillance Program Due to an
Oversight.
This issue involved four steam generator blowdown system RMs that had not
been adequately functionally tested.
These RMs are required to be
functionally tested to meet quarterly technical specification SR 4.3.3.9.
The inspector has reviewed LER 327/88013 and discussed the testing of
radiation monitors with the IM department.
Functional testing of
radiation monitors has been completed. SI-1 was reviewed and appeared to
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be adequate. This issue was-also addressed as a CAQR which was closed in
NRC Inspection Report 327,328/88-19. The licensee's. actions appear to be
adequate.
This LER is closed.
(Closed) LER 327/88015, TS SR to. Sample Containment not Properly
,
Implemented Before Containment Purge Operations Oue to Incomplete.
Procedure Review.
This event involved the failure ~ to
implement TS SR 4.11.2.1.2 which
requires upper and lower containment be sampled for radioactivity analysis
prior tMall containment purge operations.
The licensee determined that
the cause of this event was an inadequate procedure, SI 410.2, Containment
(Upper, Lower) Purge, that only required sampling of the portion of
containment to be purged.
The inspector has reviewed TS SR 4.11.2.1.2, SI-1, Surveillance Program,
Appendix F, ICF 88-0626 (issued March 18, 1988) and SI-410.2. ICF 88-0626
changed SI-410.2 and resolved the prob em. The licensee's actions appear
to be adequate.
This LER is closed.
(Closed) LER 328/88005, Loosening of Gland Seal Bolts on Speed Increaser
Lube Oil Pumps Causes a Potential Inoperability of Both Unit 2 CCPs.
This event involved the CCP speed increaser lube oil pump failure on CCP
2A-A due to the wrong lube oil pump (a 900 rpm versus the required
1800 rpm)
being
installed during
a
previous
pump
replacement.
Investigation of the Unit 1 and Unit 2 CCPs discovered that three of the
four CCPs had the wrong speed increaser lube oil pumps installed.
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The inspector has reviewed LER 328/88005, WR B257712, WR 8257714, WR
B247090, WR B247453, Maintenance Instruction (MI) 12.3.1 Centrifugal
Charging Pump Speed Increaser Inspection and Maintenance, ICF 88-0729 and
CAQRs SQP 880161 and SQP 880188. The three units with improper speed
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increaser lube oil pumps have had the pumps replaced or rebuilt to proper
requirements. CCP 1A-A had the proper speed increaser lube oil pump, CCP
IB-B had the pump replaced; CCP 2A-A had the pump rebuilt, and proper
gears, bearings and seals installed, CCP 28-B had the pump replaced.
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12.3.1 has been revised to require verification of proper components. The
licensee's actions appear to be adequate.
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This LER is closed.
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(Closed) LER 328/88007, Main Steam Isolation ESF Due to an Inadequate
Calibration Procedure.
This event involved . the lack of communications between operations and
maintenance department personnel. The root cause of this event was
inadequate maintenance department training on performance of procedure,
IMI-99 CC 10.313, Off-line Channel Calibration of Turbine-Impulse Pressure
Channel I during Mode 4 operation.
When operations personnel asked if
bi-stables would be tripped, maintenance technicians thought that tripping
of the bi-stables was not required.
When IMI-99 CC 10.3B Step 5.39.11
was performed the bi-stables were tripped and an ESF actuation resulted.
The inspector reviewed LER 328/88007, IMI-99 CC 10.313 and IMI-99 CC
10.48, and discussed the LER writeup and root cause determination with the
responsible P0RS member.
The licensee corrective action included establishing a work control group,
requiring a field verification of procedures prior to first time usage or
af ter revisions, and revising IMI-99 CC 10.38 and IMI-99 CC 10.4B to
require unit operator authorization prior to' tripping of bi-stables. The
licensee's corrective actions appear to be adequate to prevent recurrence.
This LER is closed.
(Closed) LER 328/88011 Revision 1, A Surveillance Requirement Used to
Verify Baron Concentration in.the Cold Leg Accumulators Was Not Performed
Within the Applicable Time Frames.
This event involved operations personnel not performing SR 4.5.1.1.1.b on
Accumulator 3 as required. The Reactor Coolant System (RCS) inventory had
been leaking into the accumulator, requiring the accumulator to be drained
periodically.
The inspector has reviewed LER 328/88011 Revision 1, 501-63.1, ECCS, SR 4.5.1.1.1.b, and a training letter issued April 27, 1988. The licensee's
actions appear to be adequate.
This LER is closed.
(Closed) LER 328/88012, Two Improper Operability Determinations Relating
to a Level Control Valve Resulted in not Entering the Applicable LCO in a
Timely Manner.
This event involved the failure to enter LC0 3.7.1.2 in a timely manner.
Two separate events were addressed in this LER.
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While performing SI-166.8, Increased Frequency Testing of Category A
and B Valves, level control valve 2-LCV-3-175 could not be controlled
from the control room.
This level control valve was not declared
inoperable at that time.
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Steam Generator Level Indicator 2-LT-3-107 was declared inoperable
due to excessive fluctuation. IM personnel determined that the sense
line for 2-LT-3-107 required backfilling.
Operations and IM
personnel determined that backfilling 2-LT-3-107 per MI
19.1.5,
Backfilling Sensing Lines for System 3 Transmitters Located on Panel
,
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L-184 ( Accumulator Room 3), would also make 2-LT-3-175 inoperable.
LCO 3.7.1.2 was not entered at the time that the-level transmitters
were made inoperable due to work performed by IM in accordance with
MI-19.1.5.
Contributing factors for both events were:
Failure to properly interpret Technical Specifications (TS).
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Faildre
to
understand TVA generated Technical. Specification
Interpretations (TSIs).
Failure to make log entries that contain adequate detail.
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Failure to identify the test director properly.
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The inspector reviewed, MI-19.1.5, which has been revised to address the
applicability of LCO 3.7.1.2 and to clarify the effect of backfilling the
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common sense line of LT-3-175 and LT-3-107.
The TSIs have been reviewed
for adequacy and clarity, and returned to the control room.
However,
those TSIs that require revision were not included and are not committed
to be returned to the control room prior to Unit I restart. Training has
been given on AI-6, Log Entries and Review.
The TSIs and MI-19.1.5
revisions are considered to be enhancements.
The licensee's actions
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appear to be adequate.
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This LER is closed.
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The following LERs remain open pending TVA completion of corrective
'
actions:
(0 pen) LER 327/88007, Opening of Blast Doors in Secondary Containment
Boundary Outside the Conditions Set for Surveillance Testing of the ABGTS
This report details a situation where the ABSCE configuration did not
coincide with that configuration used during SR testing used to verify TS
operability of the ABGTS.
Specifically, when the TS SR test was performed the blast doors to Units 1
and 2 reactor building were closed and the containment purge system (CPS)
was secured. However, the blast doors could be (and have been) opened and
the CPS operating on a unit in Modes 5 or 6 while the other unit is in
Mode 1-4. At the time of discovery both units were in Mode 5 and the ABGTS
system was not required to be operable. This problem has been identified
on CAQR SQP 880090.
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Short term corrective action was to close the blast doors and to tag the
Unit 1 CPS out of service prior to Unit 2 entry to Mode 4.
To allow
reopening of the Unit 1 blastdoors, an -evaluation of the data from the
previous performance of the ESTS annulus vacuum draw down test was
performed to ensure that the additional annulus inleakage would not'cause
the ABGTS to be declared inoperable.
Technical Instruction- (TI)-77,
Breaching the Shield Building, ABSCE or Control-.. Room Boundaries, was
revised to require an ' evaluation of additional inleakage from primary -
containment and the annulus prior to opening blastdoors (while the
opposite unit is in modes 1-4).
Long term corrective action included a programmatic review of the
administrative controls on the ABSCE boundary and to install logic to
isolate containment purge on an opposite unit auxiliary building isolation
signal.
The. licensee intends to submit a revision to the LER by
September 1, 1988 to provide details of the corrective actions taken.
The inspector reviewed Temporary Alteration Change Form number 1-88-0230
which placed the Unit 1 CPS out of service and the revision to TI-77 and
considers these actions adequate. The inspector verified that resolution
of CAQR SQP 880090 is classed as a Unit i restart item on TROI.
This item remains open pending resolution of CAQR SQP 88090.
(0 pen) LER 327/88010, Revision 1, An Inadequate Review of the Design Basis
of Two Engineered Safety Feature Actuated Valves Resulted in the Potential
for Plant Operation Outside of the Design Basis.
During a review of SI-166, Summary of Valve Tests for ASME Section XI, the
licensee determined the stroke time for valves LCV 62-135 and LCV 62-136
was in error.
Investigation disclosed the correct maximum allowable
stroke time (MAST) to be 15 seconds vice 28 seconds listed in SI-166. The
licensee reviewed previous stroke time records for the subject valves and
determined the stroke times to be within 15 seconds for all previous
tests.
The licensee confirmed the 15 second stroke time to be correct
per .TVA/ Westinghouse evaluations and the TS as approved in May 1987. TVA
retested the valves for both units in November and December 1987 and
verified the stroke times within the maximum allowable.
SI-166 was
revised to reflect the proper MAST.
The licensee determined that the cause of this event was an inadequate
determination of the design basis for these valves during the development
of SI-166. Recurrence control actions included a review of MAST for all
safety related motor operated valves against the requirements of the FSAR,
1
Design Criteria and the TS. This review is being conducted as part of the
corrective action for CAQR SQP 871446.
TVA has committed to make any
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necessary design document changes by July 29, 1988 and revise the FSAR if
necessary at the next scheduled FSAR update.
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The NRC inspector reviewed the change to SI-166 and other supporting
' documentation for this LER. The actions taken regarding LCVs62-135 and-
'62-136 appear to be adequate.
The review of design criteria against the
. MASTS identified 32' additional valves that had MASTS greater /less than
-specified in Design Criteria.
Pending completion of the. corrective actions for CAQR SQP 871446 this item
remains open.
(open) LER 327/87012, Loss of Shutdown Decay Heat Removal Resulting from
False Indications of RCS Level in Sightglass.
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This LER reported that shutdown decay heat removal was lost in Mode 5 for
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approximately 1 1/2 hours because of an inadvertent water level loss in
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The water level had been lowered to allow
for planned maintenance in one of the SGs and the event occurred when a
buildup of debris blocked the inlet to the sightglass and caused a faulty
level indication. The inspector reviewed the LER dated February 27, 1987,
revised LER dated April 30, 1987, and associated documentation which
stated the cause of the event and proposed corrective action.
The inspector made the following determinations:
SI-673, Reactor Coolant System Level Verification Using Sightglass or
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Tygon Hose, was revised to include the requirement for weekly
,
flushing of the sightglass and radiochemistry lab analysis of debris
collected from the flush water.
SI-673 was also revised to include the physical elevations of
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pertinent features in the open reactor coolant system such as RHR
nozzles, loop centerlines, SG manways, vessel flange, etc.
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The licensee evaluated the need to modify the design of the existing
1/2 inch sensing lines to the sightglass and determined that
increasing line size would have little or no effect on the problem.
The lines were designed with continuous upward slope, but were
observed to be discontinuous as the result of heavy objects being
placed on the tubing and loose hanger clips allowing tubing to slip.
WR B232170 was written to correct these deficiencies.
The task force which iavestigated the cause of the event recommended
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various actions which included the installation of a diverse level
indicating system for the RCS. CCTS item NC0-37-0077-002 was written
to trati this commitment which was scheduled for completion by the
,
next refueling outage.
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The inspector agreed that the corrective actions as stated in the revised
response should preclude recurrence of the event and that the proposed
schedule that the licensee has committed to is adequate.
However,
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installation of the diverse level indicating system remains to be
,
completed.
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This LER remains open.
(open) LER 328/88013, Rod Control System Deficiencies Caused Inaccuracies
in the Rod Group Demand Position Indication Resulting in Three Manual Reactor Trips.
These events involved three manual reactor trips initiated because of
problems associated with the rod group demand position indication.
Two
events were caused by faulty individual rod position indications and one
was caused by an open switch in the rod lift coil circuitry.
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The inspector has reviewed LER 328/88013, WR-B267610, replacement of group
1 demand step counter for control bank B, WR-7B290346, determine . problem
on open disconnect switch, and WR-8290792, troubleshooting to determine rod
control problem and to replace bad components if any.
The licensee's
actions appear to be adequate for Unit 2 restart.
However, long term
commitments associated with Unit 1 restart are being tracked by the CCTS
and are listed as follows:
NC0 88 0110 001, Continued Troubleshooting of Circuit Boards.
<
NCO 88 0110 002, Procedure to Cycle Rod Control Logic. Circuitry.
NCO 88 0110 003, Prepare a Supplement to this LER if Troubleshooting
.
Requires it.
NCO 88 0110 004, Clean and Test Demand Step Counters.
NCO 88 0110 005, Review of Technical Specification (TS) 3.1.3.3.
Necessary actions to close this LER are verification of completion of the
above corrective actions.
This LER will remain open.
(0 pen)
Nonconformance with
Configuration
Control
Requirements Following a Post Modification Test of a Radiation Monitor
Resulted in a Containment Ventilation Isolation.
This event involved an EMI induced signal in the radiation monitor of
sufficient strength to cause a CVI.
This EMI occurred when instrument
maintenance personnel returned the local sample pump switch to the normal
(run) position, without notifying operations. Thus the cause of this CVI,
was an actuation of a control without proper authorization. The licensee
has issued a memorandum to all IM personnel stating that operations
personnel shall be contacted, and the RM trip signal blocked before
performing any work on RMs capable of actuating ESFs.
CVIs have been
generated several times by EMIs resulting from contact chatter. A team
was established to determine causes of EMIs and to suggest a fix to
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eliminate or reduce the occurrences.
This team ' was established as a
result of a CAQR which was closed in NRC IR 327,328/88-19. Modifications
have-been completed to the involved RMs. The licensee's actions appear to.
be adequate. However, the licensee intends to issue a supplement to this
LER to detail specific actions taken to prevent recurrence.
This LER remains open pending issuance and review of the revised LER.
(0 pen) LER 327/88017 Inadequate Tagging of a Radiation Monitor Pump Switch
Results in a Containment Ventilation Isolation.
.This event involved an EMI induced signal of sufficient magnitude to cause
a RM actuated CVI. This CVI was generated in the same canner as the one
described in LER 327/88014.
The inspector has reviewed LER 327/88017 and concurred with the root cause
determination. The licensee has placed a hold order (H0) on the RM local
sample pump switch as an interim corrective action. Long term corrective
action will consist of a permanent sign on the RMs, that would indicate
that improper operation could cause ESF actuation and instruct that RMs be
blocked prior to any work performance.
A contributor to this event was
contact chatter (see LER 327/88014 in this report).
The licensee's actions
appear to be adequate.
,
This LER remains open pending review of completed long term corrective
actions.
(0 pen) LER 327/87030, Revision 1, Blown Fuses in Emergency Start Circuits
Results in Spurious Emergency Diesel Generator Starts on Two Occasions.
This event involves the spurious start of the standby diesel generators
due to mechanical failure of the FLAS-5 fuses supplied by Littlefuse, Inc.
The licensee contacted Littlefuse, Inc. and was informed that the
production process was changed after lots 2 & 3 were delivered to TVA.
This change should make the fuses more reliable. This LER also fulfills
the 10 CFR 21 report commitments.
The inspector has reviewed the LER, the licensee's response to the problem
and Information Notice (IN) 87-62, Mechanical Failure of Indicating-type
Fuses.
The problem appears to be in the manufacturing process of fuses
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prior to. lot number 4.
No fcilures have been attributed to fuses supplied
after lot 3.
The licensee has replaced all fuses from lots 2 and 3 and
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has completed testing of FLAS-5 fuses from lots 4, 6, 10, 11 and 12. The
test results indicate an average expected life of 80 months and a minimum
expected life of 23 months. The licensee has committed to evaluating the
application of FLAS-5 fuses by March 31, 1989. This commitment is tracked
by CCTS Number NCO 870 227 003, Evaluate Application of FLAS-5 Fuses.
This LER, remains open pending review of the TVA evaluation.
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(0 pen) LER 327/87043, Control Room Isolation Caused by Electromagnetic
Interference on a Radiation Monitor During Flow Switch Testing.
During surveillance testing of sample flow switches for the main control
room intake process radiation monitor 0-RM-90-206, electrical disturbances
from- the switch contacts -chattering caused an inadvertent control room
ventilation isolation due to a high radiation signal on radiation monitor
0-RM-90-126.
The plant staff verified that no high radiation levels
existed.
The licensee has revised Design Change Request (DCR)-2276 to
modify monitor 0-RM-90-206.
Modification of 0-RM-90-126 is complete.
Modification of 0-RM-90-206 is not complete.
This item remains open pending completion of corrective action.
(0 pen) LER 327/87070, Revision 1, Diesel Generator Voltage Low When Output
Breaker Closes Because of a Component Defect Found During Surveillance
Testing and Design Deficiency Outside of Plants Design Basis.
During the Conduct of Special Test Instruction (STI)-77, Loss of Offsite
Power with Safety Injection on October 21, 1987, EOG 2A-A failed to
achieve the required output voltage at the time of the output breaker
closure. The cause of the low voltage condition was determined to be due
to a defective voltage regulator component (installed November 8,1986)
and output breaker close permissive relays tolerance that allowed breaker
closure at less than minimum design voltage.
The defective voltage
regulator component was repaired and 10 CFR Part 21 notifications were
appropriately accomplished. The remaining EDGs and a spare component were
verified to be correct.
Following repairs to the voltage regulator, EDG
. voltage rise still exhibited slow development with respect to breaker
closure. Further investigation revealed that the EDG output breaker close
permissive relays CX and AX tolerance was excessive, allowing the breaker
to close prior to achieving the required voltage.
Adjustments were made to the EDG governor actuators to achieve the
required voltage output at time of breaker closure. STI-77 was completed.
The performance data was analyzed by NRC staff in March 1988. Results of
the analysis are presented in NUREG 1232, Volume 2, Part 1 (SER) dated
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March 25,
1988, Revised Preliminary Safety Evaluation Report which
accepted the DGs for Unit 2 restart. However, the SER noted that the
conclusions for Unit 2 restart assumed Unit 1 in a cold shutdown condition
and that additional TVA calculation is required to support Unit I restart.
This LER also identified the PMT following maintenance as inadequtte. The
guidelines issued May 28, 1988, were reviewed, discussed with the licensee
and appear adequate as interim guidelines. TVA has requested an extension
to the commitment in the LER (30 days after Unit 2 Mode 2) to provide PMT
procedures for EDG testing following maintenance.
The following items require resolution prior to this LER closure.
Acceptable TVA calculations supporting two unit operation.
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Acceptable PMT procedures.
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This LER remains open.
6.
Inspector Follow-up Items
Inspector Follow-up 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 and the inspector has either resolved the concern
identified, determined that the licensee has performed adequately in the
area, and/or determined that actions taken by the licensee have resolved
the concern.
(0 pen) IFI
27/85-17-06, 327/85-17-05, Review Licensee QA Exceptions for
Feed Water Isolation Valve Stem Replacement.
This IFI was initiated to evaluate TVAs process for controlling the
installation of safety related material that has been identified as
nonconforming or has inadequate documentation. This question arose when a
valve stem with outstanding licensee identified QA exceptions to vendor
documentatien was installed on valve 1-FCV-3-47.
Installation of items prior to final disposition of nonconformances is
addressed in Section 7.6 of AI-11, Receipt Inspection, Nonconforming
Items. QA Level / Description Changes and Substitutions.
This procedure-
requires
plant manager approval
for the installation,
and the
Nonconforming Item (NCI) tag to remain attached to the item and Site
Quality Managers approval. However, the responsibility for ensuring that
the equipment or associated system is not declared operable until the NCI
is cleared is given to the requesting organization. No formal controls or
documentation requirements are currently in place to assure that. equipment
or systems are not declared operable with outstanding non-conforming
conditions. CAQR SQN 880350 has been issued to address this concern.
It
is expected that AI-46, Supplemental Requirements for Release of Material
From Power Stores-Unit 0,
will be revised to provide the required
controls.
-The inspector reviewed Maintenance Request A-300025 which removed and
installed the replacement stem in 1-FCV-3-47 and supporting documentation.
The nonconformance was related to the fact that the vendor's Certificate
of Conformance (C0C) had not been provided.
Documentation shows the C0C was received and the nonconformance cleared in
May 1985.
The inspector reviewed the C0C. There is no technical concern
related to this individual material.
This item remains open pending completion of programmatic corrective
actions.
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(Closed) IFI 327,328/86-69-05, Possible Loss of ENS With Loss of Offsite
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Power.
This item involves the requirement, identified in IE Bulletin 80-15, that
the ENS have a reliable power source. Sequoyah has a station package that
requires on-site power. The ENS hcs an uninterruptible power supply that
is backed up by a battery operated power supply (thirty minute minimum).
ECN 7088 and WP 7088-1, installed an inverter powered from a reliable
backup source-that completes the necessary ENS modifications.
The inspector has reviewed Bulletin 80-15, NRC IR 327.328/86-69, the
licensee's response (Gridley/ Grace) dated November 5, 1986, ECN 7088, WP
7088-01 and the installation.
The licensee's actions appear to be
adequate.
This item is closed.
(Closed) IFI 327,328/87-19-01, Control of Shelf Life On In-Plant Stored
Material.
This item involved a question raised regarding the shelf life of material
requisitioned from power stores and subsequently stored for extended
periods- of time in the plant.
Specifically, the inspector raised
,
questions regarding the shelf life of the reactor cavity seal, which is
stored in the plant between uses.
In addition, the inspector requested
information regarding procedural controls for in plant storage of other
shelf life material.
Although the manufacturer specifies a warehouse shelf life of twelve years
for the reactor cavity seal, once the seal is used it may continue to be
reused as long as it is not damaged or worn or subjected to sufficient
radiation to cause material embattlement.
MI-1.2, Removal And Replacement
Of RPV Head And Attachments, Revision 24, Section 9.2.1, requires that a
visual inspection for damage or wear be performed and that a durometer
check be performed to detect material degradation prior to each use.
Between uses the seal is stored in an air-tight stainless steel box in the
Auxiliary Building, as it is not desirable to store it in the power stores
warehouse due to the presence of radioactive contamination.
On February 10, 1987, a Directive (R00-870210-910) was issued to all TVA
sites, requiring that no spara material be issued from power stores
without a specifically designated end use.
This action should prevent
storage of materials within the plant and preclude exceeding shelf life
intervals of material that is outside warehouse control.
Corrective
action for this item appeared to be adequate.
This item is closed.
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7.
10 CFR Part 21 Report Follow-up (90712)
(0 pen) Part 21 (P21)-86-01, Atwood and Morrill Main Steam Isolation Valves
Spring Failure Caused by Quench Cracks.
This issue was addressed in NRC Information Notice (IN) 86-81. .The
inspector reviewed the licensee's response to IN 86-81.
The inspector
verified that corrective actions had been completed on Unit 2 and that the
Unit 1 actions were being tracked on the licensee's TROI system under the
title, Isolation Valves and is scheduled for completion prior to entry
into Mode 3.
This item remains open.
(0 pen) P21-87-01, Failure in Silicone-Rubber Insulated Cables Manufactured
by Rockbestos, Anaconda-Erickson, and American, During High Potential
Test. (Possibly Damaged During Transport).
NRC Inspection Report 327,328/87-76, previous discussions between the NRC
and TVA management and TVA commitments specify additional testing of
selected cable samples during the next refueling outage.
The requirement
for performance of additional cable testing is committed to in the
licensee's CCTS (NC0 87 0322 002).
Failed silicone rubber cables have
been replaced.
This item remains open pending completion of additional testing.
(0 pen) P21-87-02, Defective MIS-5 Indicating Fuses.
This report was documented in LER 327/87030 (paragraph 6) and describes a
possible generic problem with Littlefuse, Inc., FLAS-5 fuses manufactured
prior to lot number 4.
Subsequent to lot 3,
Littlefuse changed the
production process and the solder material used in FLAS-5 fuses.
This
change appears to have corrected the observed problem. TVA has conducted
testing of FLAS-5 fuses from lots 4,
6,
10, 11 and 12.
The results
indicate an average expected life of 80 months with a minimum expected life
of 23 months.
TVA is cor.tinuing to evaluate the application of FLAS-5
fuses.
This evaluation is expected to be completed by March 31, 1989.
This issue remains open.
(Closed) P21-87-03, Miswired Diode Causes Slowed Voltage Response of EDG
Voltage Regulator.
Regulator Manufactured by Basler Electric.
On November 8, 1986, during troubleshooting of voltage fluctu6tions of the
2A-A EDG the licensee replaced the voltage regulator, supplied by Basler
Electric Company. Post maintenance testing failed to detect the defective
replacement part. However, during performance of STI-77, Loss of Offsite
Power With Safety Injection-2A-A Containment Isolation Test, the DG
(i
-
.
.
.
.
.
.
.
.
.
24
t'
voltage . failed to respond as specified.
Subsequent troubleshooting
revealed that a diode in the voltage regulator was installed incorrectly.
The licensee's corrective actions properly wired the diode in the voltage
,
regulator for 2A-A EDG, verified the diode was installed correctly in all
~
other EDG regulators, inspected a spare voltage regulator as properly
'
configured, and confirmed that this type voltage regulator is not used at
other TVA nuclear facilities.
A 10 CFR 21 notification was provided to
NRC Region II on November 18. 1987, followed by written LER 327/87070.
Failure of the post maintenance testing program to detect the defective
voltage regulator at the time of installation is being tracked by
Violation 327,328/87-76-01. Based on the review conducted, the licensee's
actions pertaining to this particular 10 CFR 21 item are adequate.
This item is closed.
8.
NRC Bulletins (92701)
(Closed)Bulletin 80-12, Oecay Heat Removal (DHR) System Operability
The closure of this oulletin is discussed in paragraph 4 under URI
327,328/86-46-06, a follow-up item to the bulletin.
This bulletin is closed.
.
9.
List of Abbreviations
-
Auxiliary Building Gas Treatment System
-
Auxiliary Building Secondary Containment Enclosure
AI
-
Administrative Instruction
Abnormal Operating Instruction
A0I
-
BIT
-
Boron Injection Tank
C&A
-
Control and Auxiliary Buildings
CAQR
-
Conditions Adverse to Quality Report
Centrifugal Charging Pump
-
Corporate Commitment Tracking System
CCTS
-
COPS
-
Cold Overpressure Protection System
-
Control Room Emergency Ventillation System
Containment Ventilation Isolation
-
-
Direct Current
DCN
-
Design Change Notice
-
-
Emergency Instructions
EI
-
-
Emergency Nctification System
Engineered Safety Feature
-
Flow Control Valve
-
GDC
General Design Criteria
-
GL
-
Generic Letter
,
.- - - .. - ..
-
.
.
.
.
.
.
.
.
-
25
HIC -
Hand-operated Indicating Controller
H0
-
Hold Order
-
Inspection and Enforcement
IEB -
Inspection and Enforcement Bulletin
IM
Instrument Maintenance
IMI
-
Instrument Maintenance Instruction
IR
-
Inspection Report
KVA -
Kilovolt-Amp
KW
-
Kilowatt
Kilovolt
KV
-
LER -
Licensee Event Report
LCO -
Limiting Condition for Operation
MI
-
Maintenance Instruction
NOV .-
Nuclear Regulatory Commission
NRC
-
OS LA -
Operations Section Letter - Administrative
OSLT -
Operations Section Letter - Training
OSP -
Office of Special Projects
PMT -
Post Modification Test
i
PORS -
Plant Operation Review Staff
PRO -
Potentially Reportable Occurrence
-
Quality Assurance
-
,
-
Regulatory Guide
Radiation Monitor
-
-
Reactor Operator
-
RWST -
Reactor Water Storage Tank
SER -
Safety Evaluation Report
-
-
Surveillance Instructien
System Operating Instructions
SOI
-
SR
-
Surveillance Requirements
Senior Reactor Operator
SR0
-
STI -
Special Test Instruction
TACF -
Temporary Alteration Control Room
TROI -
Tracking Open Items
TS
-
Technical Specifications
TVA -
Tennessee Valley Authority
Violation
-
Work Control Grouc
WCG
-
WP
-
Work Plan
-
Work Request