ML20036C355
| ML20036C355 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 05/25/1993 |
| From: | Balmain P, Brian Bonser, Skinner P, Starkey R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20036C352 | List: |
| References | |
| 50-424-93-07, 50-424-93-7, 50-425-93-07, 50-425-93-7, NUDOCS 9306160171 | |
| Download: ML20036C355 (19) | |
See also: IR 05000424/1993007
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UNITED STATES
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NUCLE AR REGULATORY COMMISslON
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REGION il
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101 MARIETTA STREET, N.W.
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ATLANTA, GEORGI A 30323
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Report Nos.:
50-424/93-07 and 50-425/93-07
Licensee: Georgia Power Company
P.'O. Box 1295
Birmingham, AL 35201
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Docket Nos.:
50-424 and 50-425
License Nos.: HPF-68 and NPF-81
Facility Name: Vogtle 1 and 2
Inspection Conducted: March 28 - May 1, 1993
Inspector: h8
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Bon
nior Resident Inspector
Date Signed
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R. D. Stark y'
sident Inspector
Date Signed
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9 Zo 9'3
P. A. BO ma
dent Inspector
"Date Signed
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Accompanied by: J . l,4 Starefos
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Approved by:
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P/ Skinnet,' Chief
Date Signed
Reactor Projects Section 3B
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Division of Reactor Projects
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SUMMARY
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Scope:
This routine, inspection entailed inspection in the following
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areas: plant operations, surveillance, maintenance, plant
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modifications, refueling activities, review of overtime records,
-and follow-up of open items.
Results:
One violation, three non-cited violations (NCV), and one
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unresolved item (URI) were identified.
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The violation occurred from a failure to take adequate corrective
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action when troubleshooting an illuminated bistable light on the
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main control board.
Inadequate preparation of the troubleshooting
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activities subsequently resulted in a Unit I loss of decay heat
removal (paragraph 8b). .
One-NCV involved the performance of a procedural step out of
sequence during performance of a Unit 1 Engineered Safety Features
g6260171 930525
ADDCK 05000424
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Actuation System Test resulted in the loss of Unit 1 Train B IE
switchgear (paragraph 2d). The second NCV involved a failure by
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reactor engineering to properly implement a procedure during Unit
I low power physics testing and resulted in a manually initiated
reactor trip. This event is also considered a weakness in
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implementing reactivity controls during low power physics testing
(paragraph 2f). A third NCV concerned the absence of written
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documentation for_ authorization to exceed Technical Specification
and procedural guidelines for excess overtime (paragraph 7).
One URI was identified concerning modifications made to the Unit 1
and Unit 2 sequencer automatic test insertion (ATI) features. The
modifications did not receive a 10 CFR 50.59 review because the
ATI feature is not explicitly discussed in the Final Safety
Analysis Report, although it.is; discussed in the licensing basis
Safety Evaluation Report (paragraph 5b).
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REPORT DETAILS
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Persons Contacted
Licensee Employees
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- J. Beasley, Assistant General Manager Plant Operations
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S. Bradley, Reactor Engineering Supervisor
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- W. Burmeister, Manager Engineering Support
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S. Chesnut, Manager Engineering Technical Support
- C. Christiansen, SAER Supervisor
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- C. Coursey, Maintenance Superintendent
R. Dorman, Manager Training and Emergency Preparedness
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- W. Dunn, Unit Superintendent
G. Frederick, Manager Maintenance
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M. Griffis, Manager Plant Modifications
M. Hobbs, I&C Superintendent
K. Holmes, Manager Operations
- D. Huyck, Nuclear Security Manager
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- W. Kitchens, Assistant General Manager Plant Support
- R. LeGrand, Manager Health Physics and Chemistry
G. McCarley, ISEG Supervisor
R. Moye, Plant Engineering Supervisor
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- M. Sheibani, Nuclear Safety and Compliance Supervisor
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- M. Slivka, ISEG Senior Tech Specialist
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W. Shipman, General Manager Nuclear Plant
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- C. Stinespring, Manager Administration
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- J. Swartzwelder, Manager Outage and Planning
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C. Tynan, Nuclear Procedures Supervisor
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J. Williams, Supervisor Work Planning and Controls
Other licensee employees contacted included technicians, supervisors,
engineers, operators, maintenance personnel, quality control inspectors,
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and office personnel.
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Oglethorpe Power Company Representative
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T. Mozingo
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NRC Resident Inspectors
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- B. Bonser
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D. Starkey
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- P. Balmain
J. Starefos
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- Attended Exit Interview
An alphabetical list of abbreviations is located in the last paragraph-
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2.
Plant Operations - (71707)
a.
General
The inspection staff reviewed plant operations throughout the
reporting period to verify conformance with regulatory
requirements, Technical Specifications, and administrative
controls.
Control logs, shift supervisors' logs, shift relief
records, LC0 status logs, night orders, standing. orders, and
clearance logs were routinely reviewed. Discussions were
conducted with plant operations, maintenance, chemistry health
physics, engineering support and technical support personnel.
Daily plant status meetings were routinely attended.
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Activities within the control room were monitored during shifts
and shift changes. Actions observed were conducted as required by
the licensee's procedures. The complement of licensed personnel
on each shift met or exceeded the minimum required by TS.
Direct
observations were conducted of control room panels,
instrumentation, and recorder traces important to safety.
Operating parameters were verified to be within TS limits.
The
inspectors also reviewed DCs to determine whether the licensee was
appropriately documenting problems and implementing corrective
actions.
Plant tours were taken during the reporting period on a routine
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basis. They included, but were not limited to the turbine
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building, the auxiliary building, electrical equipment rooms,
cable spreading rooms, NSCW towers, DG buildings, AFW buildings,
and the low voltage switchyard.
During plant tours, housekeeping, security, equipment status and
radiation control practices were observed.
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The inspectors verified that the licensee's health physics
policies / procedures were followed. This included observation of
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HP practices and review of area surveys, radiation work per=its,
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postings, and instrument calibration.
The inspectors verified that the security organization was
properly manned and security personnel were capable of performing
their assigned functions.
Inspectors observed that persons and
packages were checked prior to entry into the PA; vehicles were
properly authorized, searched, and escorted within the PA; persons
within the PA displayed photo identification badges; and personnel
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in vital areas were authorized.
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b.
Unit 1 Summary
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The unit began the period with the reactor defueled. The unit
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entered Mode 6 and began core reload on April 5.
The core reload
was completed on April 11, Reactor vessel head tensioning was
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completed and the unit entered Mode 5 on April 15. The unit
entered Mode 4 on April 21, Mode 3 on April 22, and Mode 2 on
April 24. The reactor was taken critical during physics testing
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on April 24 but reentered Mode 3 following a manual reactor trip
on April 25 due to an unexpected large negative reactivity
insertion. The unit was taken critical and returned to mode 2 on
April 26.
Physics testing was completed and the unit entered Mode
1 on April 26. At the close of the inspection period the unit had
reached 94% power.
c.
Unit 2 Summary
The unit began the period operating at 100% power and operated at
full power throughout the inspection period.
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d.
Loss of Unit 1 Train B IE Switchgear During ESFAS Test
On April 15, the licensee was performing procedure 14667-1, Train
B DG and ESFAS Test, section 5.2, Loss-of-offsite Power / Safety
Injection. The unit was in Mode 6, Refueling, with the reactor
head set, and RHR A was providing shutdown cooling.
Prior to the
manual initiation of the UV/SI signal as directed in the
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procedure, the IB DG had been paralleled to the safety-related
4160 V switchgear, 1BA03, and "B" Train RHR, CS, AFW, NSCW, and
CCW pumps had been started. When the UV/SI signal was manually
initiated the normal incoming supply breaker from the "B" RAT to
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1BA03 tripped as expected, the loads on 1BA03 were shed and then
sequenced back on to IBA03 which was still being powered by the IB
DG. All loads sequenced automatically on to IBA03 as expected.
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At 10:28 a.m., approximately five minutes after all loads had been
automatically sequenced onto IBA03, the IB DG output breaker
tripped open resulting in all power being lost to IBA03.
Consequently, all the "B" train safety related pumps tripped. Of
particular concern to the operators was the loss of NSCW which
supplies cooling water to the DG and they manually initiated an
emergency trip of the IB DG. During this entire evolution A Train
safety-related equipment was not affected and RHR A continued to
provide shutdown cooling. Operators also entered TS Action
Statement 3.9.8.2, since less than two trains of RHR were
operable. TS 3.9.8.2a requires that the inoperable train be
restored as soon as possible. At 10:40 a.m., power was restored
to IBA03 from' the normal supply via the "B" RAT. At 10:43 a.m.,
RHR "B" was available for operation and TS 3.9.8.2a was exited.
The cause of this event was the failure of personnel directing
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procedure 14667-1 to' perform procedure steps in the proper
sequence.
Specifically, step 5.2.20, which requires that the SI
signal be reset, was completed prior to step 5.2.19, which
requires that the 151V phase A time-overcurrent relay at the IB DG
be manually tripped. The time-overcurrent relay trips the 186B
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lockout relay. During a non-emergency start of the DG a trip of
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the 186B lockout relay will cause the DG to trip.
However, this
trip is bypassed during an emergency start of the DG.
In this
event, the SI signal had been prematurely reset and therefore the
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UV/SI signal, which initiated the DG emergency start and would
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have prevented a DG trip on a trip of the 186 B lockout relay, had
been removed.
Subsequently, when the Phase A time-overcurrent
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relay was tripped in step 5.2.19, which caused the 186B relay to
trip, resulting in the DG tripping.
The inspector was present in the control room during this ESFAS
test evaluation. The evolution was thoroughly briefed to all
participants by the test director and the Manager Operations.
During the test the test director became distracted when
questioned by other personnel in the control room as to when the
SI could be reset. He mistakenly directed that the SI be reset
and by doing so performed a procedural step out of sequence. The
test director was counseled. The inspector witnessed the
procedure again when it was successfully reperformed and was
satisfied that it was adequate and that the mistake which was made
previously was due to personnel error in not performing the
procedural steps in sequential order.
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Performing a procedural step out of sequence was a failure to
implement procedure 14667-1, Train B DG and ESFAS Test, section
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5.2, Loss-of-Offsite Power / Safety Injection; and represents a
violation of TS 6.7.la.
This violation will not be subject to
enforcement action because the licensees' efforts in identifying
and correcting the violation meet the criteria specified in
Section VII.B of the NRC Enforcement policy. This violation is
identified as NCV 50-424/93-07-01, Failure To Follow Procedure.
e.
Unit 1 Containment Ventilation Isolation
On April 11, with Unit 1 in Mode 6, a CVI occurred due to high
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radiation on containment low range area radiation monitors 1 RE-
002 and 1 RE-003, during placement of the reactor upper internals
in the reactor vessel. The highest reading on 1 RE-002/003 was
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approximately 50 mR/Hr which was expected for the evolution in
progress.
The CVI actuation setpoint was 15 mR/hr.
Procedure 12000-C, Post Refueling Operations, requires that 1RE-
002/003 be reset to a higher alarm setpoint during this evolution.
Several factors contributed to the setpoints not being increased
as required. The USS was involved with several activities and his
attention was diverted away from the task at hand which he felt
was to complete the setting of the upper reactor vessel internals
on his shift. The internals were set late in the shift while
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control room turnover activities were being conducted. The USS
was using procedure 12007-C, Refueling Operations, and had delayed
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transition to procedure 12000-C because step 4.6.1- of 12007-C
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could not be signed off until completion of " core alterations"
which includes setting of the upper internals. The USS felt that
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it was inappropriate to transition to procedure 12000-C until all
steps in procedure 12007-C had been signed off. These factors
contributed to the USS not reviewing ahead in procedure 12000-C to
the step which requires the adjustment of setpoints for 1RE-002
and 003 to a higher value to prevent a CVI from occurring.
A pre-job briefing had been conducted by the Manager Operations
but no caution was stated during that briefing about a possible
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CVI with regard to 1 RE-002/003.
Procedure 93240-C, Reactor
Vessel Assembly-Disassembly, which was used as a guide for that
briefing, did not have either a caution or a specific step
requiring notification to adjust the setpoints of 1 RE-002/003
prior to moving the upper internals.
The following corrective actions have or will be taken regarding
this event.
The USS was disciplined regarding the importance of
procedural compliance. Appropriate operations personnel will be
reminded of the importance of taking time to review upcoming
procedural steps. Steps in procedure 12007-C referring to Post
Refueling Operations will be deleted and will be added to
procedure 12000-C.
Procedure 93240-C will be revised to include a
step to verify with the USS that RE-002 and RE-003 setpoints have
been raised prior to movement of the reactor internals or head.
This event will be reviewed during licensed operator
requalification training and will be included in the operations
required reading book. The inspector will review the corrective
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actions during the review of the associated LER.
f.
Unit 1 Manual Reactor Trip Due to Large Negative Reactivity
Insertion During Physics Testing
On April 25, while performing Unit I reload low power physics
testing at less than one percent power, a large negative
reactivity insertion occurred during rod worth measurements. The
licensee determined that the reactivity insertion exceeded -40 pcm
and then initiated a manual reactor trip in accordance with
Abnormal Operating Procedure 18003-C, Rod Control' System
Malfunction, due to symptoms of a potentially dropped or
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misaligned rod.
The inspector reviewed the sequence of events with licensee
personnel.
Immed ately prior to the negative reactivity
insertion, rod worth measurements (rod swaps) were being taken in
accordance with procedure 88002-C, Reload Low Power Physics
Testing. A rod bank exchange of SDB and CBD was in progress.
SDB, the reference bank, which was nearly fully inserted was being
withdrawn and CBD was alternately being inserted. When both banks
were in the lower portion of the core and as CBD was being further
inserted the reactor operator observed a higher than normal
negative start up rate indication.
Reactor engineering test
equipment indications showed that reactivity went off-scale in the
negative direction and neutron flux had decreased rapidly.
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on these indications, reactor engineering personnel initially
concluded that this abnormality was potentially caused by a
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dropped RCCA. Control Room personnel then initiated a manual
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reactor trip in accordance with the abnormal operating procedure,
which is the appropriate action for the conditions observed.
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Following the reactor trip, the licensee took extensive actions to
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determine if a RCCA had dropped, or had become detached from its
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drive rod and dropped. The inspector noted that these were
conservative actions performed with the reactor shutdown and
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included performance of rod control system current measurements
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during the withdrawal of CBD and verifications of rod alignment on
DRPI while exercising all rod banks to determine that all RCCAs
were latched.
In addition, the NSSS vendor performed a computer
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analysis of the core response which modeled SDB and CBD in the
lower portion of the core.
This analysis determined that CBD had
a very large differential rod worth when inserted to approximately
80-100 steps.
The testing and analysis indicated that the
negative reactivity insertion was due to the insertion of CBD and
not due to a dropped RCCA.
This was also confirmed by flux
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mapping after the Unit I startup at about 30% power.
The inspector reviewed procedure, 88002-C, and noted that during
rod worth measurements the engineer responsible for performing the
test is required by steps 8.8.23.2 and 8.8.23.6 to control
reactivity insertions from -20 pcm to +20pcm. These steps note
that the insertions may be increased at the discretion of the
responsible engineer, however, the reactivity indication only
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measures a band from about -50 pcm to +50 pcm.
The inspector
reviewed with the reactor engineering supervisor the strip chart
data collected and noted that the magnitude of reactivity swings
during the withdrawal was allowed to approach the top of scale
prior to going off-scale when CBD was inserted from step 87.
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Based on the review of this' data, the inspector concluded that
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reactivity was allowed to move off-scale and was not controlled as
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procedure 88002-C intended.
In addition it was determined that
the responsible engineer was distracted for several seconds during
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the CBD insertion in this area of high worth.
Discussions with personnel conducting the test revealed that the
abnormality that resulted in the high worth of CBD and the
magnitude of the reactivity insertion was unexpected. The
inspector observed the reperformance of rod worth measurements for
CBD following the restart of Unit 1 on April 26 and noted that the
control of the test, communications, and attention in the control
room was very good.
Allowing reactivity to go off-scale while performing rod worth
measurements is a failure to implement procedure 88002-C and
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represents a violation of TS 6.7.la.
This violation will not be
subject to enforcement action because the . licensees' efforts in
identifying and correcting the violation meet the criteria
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specified in Section VII.B of the NRC Enforcement policy. This
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violation is identified as NCV 50-424/93-07-02, Failure To
Implement low Power Physics Test Procedures for Controlling
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Reactivity. This violation also represents a weakness in
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implementing reactivity controls during low power physics testing.
The inspector concluded that the corrective actions and successful
reperformance of the CBD rod worth measurement discussed above
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were adequate.
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g.
Unit 1 Unplanned ESFAS Actuation When Train A Undervoltage
Occurred During Sequencer Testing
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On May 10, during functional testing at the Train A sequencer
manual test panel following a design change, a signal was
initiated that caused the normal incoming feeder breaker to the
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Train A 4160V bus to open. The bus was energized through this
breaker and upon opening, an actual undervoltage condition
occurred on the bus which resulted in the actuation of a load shed
and UV Loading Sequence. The licensee determined that the cause
of the breaker opening was due to a test pulse which was not
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completely filtered by the sequencer's logic circuitry.
A
sequencer manual test (System-UV) was being performed from the
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manual test panel at the sequencer to verify operation of the
sequencer logic for an UV signal. The licensee determined that
electrical interaction of circuit cards caused a synchronization
offset between two circuit cards and a 200 ms test pulse was not
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completely filtered allowing part of the signal to be processed as
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an actual signal.
The licensee corrected the deficiency by
decreasing the test signal to ensure that it is filtered. The
inspector will review the corrective actions during review of the
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LER.
Two non-cited violation were identified.
3.
Surveillance Observation (61726)
a.
General
Surveillance tests were reviewed by the inspectors to verify
procedural and performance adequacy. The completed tests reviewed
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were examined for necessary test prerequisites, instructions,
acceptance criteria, technical content, data collection,
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independent verification where required, handling of deficiencies
noted, and review of completed work. The tests witnessed, in
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whole or in part, were inspected to determine that approved _
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procedures were available, equipment was calibrated, prerequisites
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were met, tests were conducted according to procedure, test
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results were acceptable and systems restoration was completed.
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SURVEILLANCE NO.
TITLE
14460-2
ECCS Flow Path Verification
88002-C
Reload Low Power Physics Testing
T-ENG-93-10
DP Stroke Test for Valves 1HV-8116 and 1HV
8111A
28810-1
Battery Service Check and 18 Month
Inspection
54804-1
ATWS Mitigation System Actuation Circuitry
Quarterly Surveillance
14667-1
Train B D/G and ESFAS Test, Section 5.2
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34324-C
Channel Calibration of 1RE-0003
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Air Vented from ECCS Return Line
On April 15, during the performance of surveillance procedure
14460-2, ECCS Flow Path Verification, the licensee observed that a
small volume of air vented from valve 2-1204-X4-451, RWST Return
Line Vent, before a clear stream of water appeared. The licensee
then initiated a DC to evaluate corrective actions.
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The inspector was initially concerned about the impact of this
condition on the ECCS and whether the surveillance performed on
April 15, met the acceptance criteria referenced in TS 4.5.2b.1.
The inspector was also concerned that this deficiency did not
receive a timely engineering evaluation.
The inspector reviewed these concerns with licensee management and
determined that the surveillance requirements of TS 4.5.2b.1,
Emergency Core Cooling Systems, were met. Confusion arose
initially regarding the acceptance criteria because the criteria
in procedure 14460-2 was not worded the same as the TS. The
licensee made a procedure revision to change the wording to
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reflect the TSs. The inspector concluded, based on these
discussions, that the volume of air was minimal and since it was
located in the return line to the RWST, there would be no impact
on the operability of the ECCS. The inspector reviewed the
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proposed troubleshooting plan with system engineering and found
the actions acceptable, and also determined that the engineering
evaluation was performed in a timely manner.
The inspector had no
concerns following this review.
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c.
Unit 1 D Train Battery Cell Failure
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On April 22, during the normal weekly battery. surveillance on the
Unit 1 D Train battery, cell number 16 was found to have a low
voltage of 2.086 volts. The cell was jumpered out and the battery
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was returned to service. On April 29, cell 16 was replaced.
The inspectors were concerned that battery ID was experiencing
cell failures again immediately following replacement of nine
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weaker cells during the ongoing refueling outage. After replacing
weaker cells with new cells it is not uncommon for the new cells
to force the voltage down on the weaker cells in the battery. A
review of this issue with an NRC battery expert identified that
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this phenomenon is not unexpected. The inspectors concluded from
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this event that battery cell failures during weekly surveillances
would probably continue on the Unit 1 A, B, and D batteries. The
C Train battery was completely replaced during the Unit I
refueling outage.
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No violations or deviations were identified.
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4.
Maintenance Observation (62703)
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The inspectors observed maintenance activities, interviewed personnel,
and reviewed records to verify that work was conducted in accordance
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with approved procedures, TSs, and applicable industry codes and
standards. The inspectors also verified that redundant components were
operable, administrative controls were followed, clearances were
adequate, personnel were qualified, correct replacement parts were used,
radiological controls were proper, fire protection was adequate,
adequate post-maintenance testing was performed, and independent
verification requirements were implemented. The inspectors
independently verified that selected equipment was properly returned to
service.
Outstanding work requests were reviewed to ensure that the licensee gave
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priority to safety-related maintenance activities.
The inspectors witnessed or reviewed the following maintenance
activities:
MWO NOS.
WORK DESCRIPTION
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19202661
Inspect Batteries in Unit IB Sequencer - Battery
Backup Module
19301557
Implement DCP 92-VIN 0171-0-2 Which Changes Out
Sequencer Controller A Circuit Board
19300565
Recalibrate Loop 2 Delta T Setpoints to
Implement Power Uprate DCPs
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MWO NOS.fcont.)
WORK DESCRIPTION (cont.)
19203444
Reland Wires on Junction Box on MSIV IHV- 3026B
19301796
Test and Rework IPSV-8510A
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19203466
AFW Trains A and B Hydro
No violations or deviations were identified.
5.
Plant Modifications (37828)
a.
Relocation of Containment Atmosphere Monitor DPM, IRX-2562
During the current Unit I refueling outage, IR4, the inspector
reviewed DCP 91-VIN 0194-0-1, DPM 1-RX-2562 Relocation. The
inspection included a review of the 10 CFR 50.59 Safety Evaluation
of the DCP, a review of the DCP, and a walkdown of work in
progress and final installation.
Radiation monitor 1RE-2562 has two sections which monitor
containment atmosphere. Section IRE-2562A monitors containment
atmosphere particulate and section IRE-2562C monitors radioactive
gas.
Prior to this design change, the DPM 1RX-2562, for radiation
monitor 1RE-2562 was located in the Auxiliary Building room B08.
This room is designed to maintain a temperature of 100 degrees F
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or less and the actual temperature is normally near 100 degrees F.
This hot environment is believed to have been the cause of various
DPM component failures. As an interim measure prior to this DCP,
the licensee had installed a portable air conditioning unit to
cool the DPH cabinet internals.
This DCP relocated the DPM to an
adjacent room with a cooler environment. The relocation involved
installation of a seismically mounted junction box, running new
conduit and cable to the new location, core drilling at the new
location, and mounting the DPM at the new location. Subsequently,
I&C and Chemistry performed testing to ensure mor.itor operability.
Radiation Monitor 1RE-2562 was returned to service prior to the
end of 1R4. The inspectors will monitor the performance of IRE-
2562 to confirm that the relocation design change achieved its
desired results. The inspector determined that the DCP, Safety
Evaluation, and installation were adequately performed. A similar
modification is planned for 2RX-2562 during a future refueling
outage.
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b.
Review of Sequencer Design Changes
During this inspection period the inspector reviewed the
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implementation of DCP 92-VIN 0171, Modify Sequencers IA and IB
Circuit Boards, and DCP 93-VIN 0003, Sequencer Automatic Reset.
The inspector also reviewed the completed documentation for
procedures T-ENG 93-04 and T-ENG 93-05, A Train and B Train
Sequencer Functional Tests, which were performed to demonstrate
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the operability of the sequencers.following the completion of the
DCPs.
Several problems occurred during functional testing which required
additional modifications to the original design change for the
controller A circuit cards or repairs to the cards.
Part of
procedure T-ENG 93-04 is performed in the manual test mode; on
several initial test attempts the sequencer would exit from this
mode. This condition was not desired and troubleshooting
activities found and repaired a broken capacitor which eliminated
this fault. An unplanned ESFAS actuation, which is described in
paragraph 29, also occurred during functional testing. The design
error that caused the actuation was determined when the condition
was recreated under procedure T-ENG 93-18, Sequencer Load Shed
Anomaly Test - A Train. The error was corrected by revising the
DCP to shorten a test pulse duration so that it will not process
as a valid signal in the sequencer logic. Another potentially
significant problem identified during testing was a 36 second
delay in the manual block function, which would have blocked any
operator initiated actions for equipment actuated through the
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sequencers for 36 seconds following the completion of a loading
sequence. The inspector discussed the test results with system
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engineering personnel and determined that the operability of the
Unit 2 sequencers was not affected.
Following the ESFAS actuation, the inspector reviewed the
licensing basis documented in the NRC SER for Vogtle and noted
that the ATI function is explicitly described in section 8.4.8,
Load Sequencing Design. During the previous Unit I refueling
outage, ATI was found to interfere with the sequencer logic and
would delay a valid stepping sequence; this problem applied to to
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both units. DCP 92-VIN 0171 was initiated to correct this design
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fl aw. The licensee prepared Temporary Modifications 1-92-008 and
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2-92-016 to disable ATI on the sequencers for both units. When
the temporary modifications were prepared the licensee did not
prepare a 10 CFR 50.59 evaluation since the ATI feature is not
explicitly described in the FSAR. Then the inspector discussed
the 50.59 evaluation process with engineering and
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engineering / technical support personnel, he identified that the
licensing basis in the SER is not generally referred to for 50.59
evaluation applicability determinations when plant modifications
are made.
The inspector was concerned that the sequencer modification
represented a change to the facility and was made without a 10 CFR 50.59 evaluation. The inspector. is also concerned that the
licensing basis documented in a NRC SER is not reviewed when plant
modifications are made. Additional information is needed to
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assess the significance of these concerns and this item is
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identified as Unresolved Item 50-424,425/93-07-03:
Licensing
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Basis Reviewed to Determine 10 CFR 50.59 Applicability.
One unresolved item was identified.
6.
Refueling Activities - Unit 1 Fuel f ailure Analysis (60710)
The inspectors reviewed results of the licensee's fuel inspection
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activities that were conducted following the Unit I refueling outage.
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The licensee's inspections identified two leaking fuel assemblies. One
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assembly, 2F22, was a first cycle Vantage 5 assembly. The other
assembly, SE34, was a second cycle LOPAR assembly. The Vantage 5
assembly was reconstituted and reloaded into the core, while the LOPAR
assembly was stored in the spent fuel pool.
During UT inspections of
the Vantage 5 assembly, one rod was found to be leaking. The damaged
rod was removed and examined with high resolution video. The inspector
reviewed the video and noted that it showed a small fretting hole
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located on the lower portion of the rod. The upper portion of the rod
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exhibited obvious hydriding of the cladding. One area in this hydrided
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region had weakened and resulted in a rupture of the cladding.
Based on
review of the video, the licensee determined the fuel rod failure was
caused by debris impinging on the bottom of this rod, which fretted
through the cladding. Water that subsequently entered the rod caused
hydriding of the upper portion of the rod. The licensee also performed
a reactor vessel F0SAR inspection to remove debris from the RCS. No
substantial debris which would have caused the fuel rod failures was
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identified. The cause of the failure of the second assembly was not
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determined since it was permanently removed.
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The inspectors concluded that the potential for future fuel failures was
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minimal, based on the review of the licensee's determination that the
fuel rod failures were confined to two fuel assemblies, and on the
results of the FOSAR inspections.
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No violations or deviations were identified.
7.
Review of Overtime Records
During this report period the inspector reviewed a sample of overtime
records for members of the plant staff who perform safety-related
functions. The review was to verify compliance with TS 6.1.le, Plant
Staff, which provides guidelines to limit the use of overtime, and
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procedure 00005-C, Overtime Authorization. A sampling of time sheets
!
was reviewed for personnel in the operations, maintenance and health
physics / chemistry departments for the period of March 20-April 2, 1993,
during refueling outage IR4.
No concerns were identified in the
Operations, I&C Maintenance, or the Health Physics / Chemistry Department.
In the Electrical / Mechanical maintenance area the inspector identified
thirteen examples of personnel who worked more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in a seven
day period without written authorization as required by procedure
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00005-C. The hours ranged from 76 hours8.796296e-4 days <br />0.0211 hours <br />1.256614e-4 weeks <br />2.8918e-5 months <br /> to 99 hours0.00115 days <br />0.0275 hours <br />1.636905e-4 weeks <br />3.76695e-5 months <br /> worked during a
seven day period. The Manager Maintenance stated that verbal approval
for the excessive overtime had been given, but neither the maintenance
department nor the inspector could locate the required written
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documentation.
The licensee subsequently completed documentation forms
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when the original forms could not be found.
T.S. 6.2.2.e requires that
any deviation from the overtime guidelines shall be authorized by the
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applicable department superintendent, or higher levels of management, in
accordance with established procedures and with documentation of the
basis for granting the deviation.
Procedure 00005-C further requires
that such overtime authorization and its basis shall be recorded on a
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form similar to Figure 1 of Procedure 00005-C. This is a violation of
This NRC identified violation is not being cited because
criteria specified in Section VII.B of the NRC Enforcement Policy were
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satisfied.
This non-cited violation is identified as NCV 50-424/93-07-
04, Failure To Document Approval of Excess Overtime.
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One non-cited violation was identified.
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8.
Follow-up (90712) (92700) (92701) (92702)
The Licensee Event Report and follow-up items listed below were reviewed
to determine if the information provided met NRC requirements. The
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determination included:
adequacy of description, verification of TS
compliance and regulatory requirements, corrective action taken,
existence of potential generic problems, reporting requirements
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satisfied, and relative safety significance of each event.
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a.
(Closed) Part 21, Potential Problem with DG Jacket Water Pump Gear
Cooper Energy Services notified the NRC by a letter dated March 9,
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1993, of a potential defect with the jacket water pump gear in
DSRV Enterprise Standby Diesel Generators.
Vogtle was identified
as having these gears installed on its emergency diesel
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generators.
Cooper stated that.the potential defect was the
result of incorrect machining by the gear supplier and that the
gear should be replaced in a scheduled manner based on the.
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availability of acceptable replacement parts.
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Cooper Energy Services issued Amendment #1 to its Part 21
notification on April 1.
That notification stated that in cases
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where the initial inspection of the pump idler gear shows no
indication of interference with the water pump gear teeth, no
further inspection or replacement would be required.
Cooper found
that interference in the idler / water pump gear mesh manifests
itself in only a few hours and that gears which have operated in
excess of twenty hours with no visual wear pattern should not be
affected.
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During Unit I refueling outage IR4 the licensee inspected the 1A
and IB DG jacket water pump gears and found no indication of the
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potential defect mentioned in the Part 21 notification. However,
the licensee, as a precautionary measure, replaced the gear with
an improved design gear. During the next Unit 2 refueling outage
the Unit 2 DG jacket water pumps will be inspected and the gears
replaced if necessary. Unit 2 jacket water pumps have operated in
excess of twenty hours each with no indication of pump gear
problems.
The inspector was satisfied that the licensee had
appropriately addressed the concern identified in the Part 21
notification.
b.
(Closed) URI 50-424/93-04-02, Review Causes of Loss of
Decay Heat Removal Event
The URI addressed inadequacies in design change package reviews
and the performance of a reactor protection system modification
with only one train of decay heat removal in service. This event,
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which resulted in a momentary loss of decay heat removal, was
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caused by the RHR inlet isolation valve,1HV-8701B, closing. As
discussed in NRC IR 50-424,425/93-04, the event occurred when an
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I&C technician, troubleshooting the cause of an illuminated
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bistable light prior to performing a functional test on the design
change to the reactor protection circuitry, removed a card in the
RHR valve autoclosure circuitry. The URI was opened pending
completion of the review of this event.
The inspectors concluded, after further review, that the cause of
this loss of decay heat removal was an inadequate review by I&C
personnel of the modification and other available material which
described the circuitry being modified. Prior to performing the
troubleshooting to clear the bistable light, I&C personnel did not
identify from their review, that the RCS wide range pressure loop
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P-408, shared the circuit card that was removed. A more thorough
review by the I&C foreman before directing the technician to
continue troubleshooting should have identified the potential
problem. As a result of the inadequate review, the technician was
poorly prepared to troubleshoot, creating additional problems
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which challenged core safety.
The inspectors also concluded that there were other contributing
causes to this event. The initial review of the DCP by I&C during
the implementation planning stages did not identify all the
circuit cards to required to be removed to deenergize each process
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loop prior to performing the DCP.
In this case, all power was not
removed from the loop before work was performed.
The RCS wide
range pressure loop was not identified as sharing part of the
process loop on which the design change was being performed and as
a result work was performed on a partially energized loop. The
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cause of the bistable light which initiated the troubleshooting
process and resulted in this event, was suspected to be a contact
between two pins on a multi-pin plug, one of which was still
energized. Also this DCP was scheduled during a period of higher
risk since, at the time of the event, RHR Train A was in the
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service for maintenance. Two SGs were available as a heat sink in
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accordance with TS 3.4.1.4.1.
The tie between the DCP work and
the single train of decay heat removal was not considered.
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Based on this review, URI 50-424/93-04-02, Review Causes of Loss
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of Decay Heat Removal Event, is closed. The loss of decay heat
removal which resulted from a failure to take adequate corrective
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action when troubleshooting an illuminated bistable light on the
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main control board is identified as violation 50-424/93-07-05,
Failure To Take Adequate Corrective Action Resulting In loss Of
De.ay Heat Removal .
c.
(0 pen) LER 50-424/91-015, Rev.1, Valve Manufacturing Defect Leads
to Containment Isolation Valve Failing Open.
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The casting mark on valve 1-2401-U4-034 was removed, LLRT testing
was completed satisfactorily, and the valve was returned to
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service.
The licensee was unable to find any similar occurrences
of this type of event. Additionally, the licensee committed to
inspect, during refueling outages 1R4 and 2R3, a representative
sample of these valves which are not inspected as part of the IST
program or the valve disassembly program. During outage IR4, a
sample of seven valves were inspected out of a total of twenty-one
possible valves.
None of the seven valves tested exhibited the
casting mark discrepancy described in the LER which would cause
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the disk to bind. During refueling outage 2R3 later this year,
the licensee will again perform a random sample of these
identified valves which are not inspected in other inspection
programs.
The inspectors will review the results of those
inspections following 2R3.
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One violation was identified.
9.
Exit Meeting
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The inspection scope and findings were summarized on April 30,
1993, with those persons indicated in paragraph 1.
The inspector
described the areas inspected and discussed in detail the inspection
findings listed below.
No dissenting comments were received from the
licensee. The licensee did not identify as proprietary any of the
material provided to or reviewed by the inspectors during the-
inspection.
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item No.
Description and Reference
NCV 50-424/93-07-01
Failure to Implement ESFAS Test Procedure
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Results in the loss of Unit 1 Train
B IE Switchgear (paragraph 2d)
NCV 50-424/93-07-02
Failure to Implement low Power Physics
Test Procedure For Controlling Reactivity
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(paragraph 2f)
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Item No. (cont.)
Description and Reference (cont.)
URI 50-424,425/93-07-03
Licensing Basis Reviewed to Determine
10 CFR 50.59 Applicability
(paragraph 5b)
NCY 50-424/93-07-04
Failure to Document Approval of Excess
Overtime (paragraph 7)
VIO 50-424/93-07-05
Failure to Take Adequate Corrective Action
Results in Loss of Decay Heat Removal
(paragraph 8b)
10.
Abbreviations
ACOT
- Analog Channel Operational Test
- Auxiliary Feedwater System
ANII
- Authorized Nuclear Inservice Inspector
- Automatic Test Insertion
- Anticipated Transient Without Scram
CBD
- Control Rod Bank D
- Centrifugal Charging Pump
- Closed Cooling Water System
CFR
- Code of Federal Regulations
CR
- Control Room
- Containment Ventilation Isolation
- Deficiency Card
- Design Change Package
- Diesel Generator
- Data Processing Module
DRPI
- Digital Rod Position Indication
- Emergency Core Cooling Systems
E0P
- Emergency Operating Procedures
- Emergency Response Facilities-
- Engineered Safety Feature
- Engineered Safety Features Actuation System
F0SAR
- Foreign'0bject Search and Retrieval
- Final Safety Analysis Report
- Instrumentation and Control
- Institute for Nuclear Power Operations
IR
- Inspection Report
ISEG
- Independent Safety Engineering Group
- In-Service Test
LC0
- Limiting Condition for Operation
LDCR
- Licensing Document Change Request
LER
- Licensee Event Report
LOPAR
- Low Parasitic Fuel
- Loss Of Off-Site Power
- Motor Operated Valve
mR
- Milli-Roentgen
ms
- millisecond
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MWO
Maintenance Work Order
- Non-Cited Violation
NPF
- Nuclear Power Facility
NRC
- Nuclear Regulatory Commission
- Nuclear Service Cooling Water System
- Nuclear Steam Supply System
- Protected Area
pcm
- Percent MilliRho
PE0
- Plant Equipment Operator
PERMS
- Process and Effluent Radiological Monitoring System
- Pounds per Square Inch
- Reserve Auxiliary Transformer
RCCA
- Rod Control Cluster Assembly
R0
- Reactor Operator
- Refueling Water Storage Tank
SAER
- Safety Audit And Engineering Review
SDB
- Shut Down Rod Bank B
- Safety Evaluation Report
- Safety Injection
- Safety Parameter Display System
TS
- Technical Specifications
- Unresolved Item
USS
- Unit Shift Supervisor
- Ultrasonic Testing
V
- Volt
- Violation
WRT
- Work Request Tag
WWRB
- Waste Water Retention Basin
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