ML17279A373
| ML17279A373 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 06/17/1987 |
| From: | Bosted C, Dodds R, Johnson P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML17279A371 | List: |
| References | |
| 50-397-87-09, 50-397-87-9, NUDOCS 8707080220 | |
| Download: ML17279A373 (16) | |
See also: IR 05000397/1987009
Text
U.S.
NUCLEAR REGULATORY COMMISSION
REGION
V
Report No:
Docket No:
Licensee:
50-397/87-09
50-397
I
Washington Public Power Supply System
P. 0.
Box 968
Richland,
WA 99352
Facility Name: Washington Nuclear Project No.
2. (WNP2)
Inspection at:
WNP-2 Site near Richland, Washington
Inspection
Conducted: April
1 - May 23,
1987
Inspectors:
. T.
Dod s, Senior
es> ent Inspector
Dat
Si
ned
Approved by:
oste
,
Ress
ent
nspector
P.
H. Johnson,
C ief
Reactor Projects
Section
3
a
e
signed
Da
e
signed
Summary:
Ins ection
on A ril
1 - Ma
23,
1987
50-397 87-09
~AI<<:
i
i
p
i
b
h
id
i
P
. room operations,
engineered
safety feature
(ESF) status,
surveillance
program,
maintenance
program,
licensee
event reports,
special
inspection topics,
spent
fuel pool
and refueling activi.ties, simulator upgrade,. initial control rod
drive installation, radiological protection,
management
involvement,
gA
program for MSTE, local leak rate testing,
BWR radiological controls for
drywells during spent fuel movements,
operator attentiveness
to duties,
control of overtime,
and licensee
action
on previous inspection findings.
During this inspection,
Inspection'rocedures
30703,
35701,
35750,
40700,
41400,
60710,
61701,
61710,
61720,
61726,
62702,
62703,
71707,
71709,
71710,
71881,
86700,
90712,
90713,
92701,
92702,
92703,
93702,
and 92700 were
covered.
Results:
Of the
23 areas
inspected,
three apparent violations were identified
pertaining to the control of gas bottles
(paragraph 3),
removal of equipment
from the fuel pool (paragraph 8), and source
check of a constant air monitor
(paragraph 8).
i
8707080220
870617
ADOCK 05000397
6
DETAILS
Persons
Contacted
J.
Shannon,
Deputy Managing Director
- C. Powers,
Plant Manager
"R. Glasscock,
Licensing
8 Assurance
Director
- J. Baker, Assistant Plant Manager
- H. McGilton, Manager,
Operational
Assurance
Program
"R. Corcoran,
Operations
Manager
"S.
McKay, Assistant Operations
Manager
~A. Hosier, Nuclear Safety Assurance
Group
(NSAG) Manager
"K. Cowan, Technical
Manager
"D. Walker,
Outage
Manager
"R. Graybeal,
Health Physics
and Chemistry Manager
D. Feldman,
Plant guality Assurance
Manager
"M. Bartlett, Operations guality Assurance
Supervisor
"J. Peters,
Administrative Manager
P.
Powell, Licensing Manager
M. Wuestefeld,
Reactor
Engineering Supervisor
"J.
Landon,
Maintenance
Manager
T. Stanley,
Principal Engineer
"S. Washington,
Sr.
Compliance
Engineer
- J. Arbuckle, Compliance
Engineer
- Personnel
in attendance
at 'exit meeting
The inspectors
also interviewed various control
room operators, shift
supervisors, shift managers,
and engineering,
quality assurance,
and
management
personnel
relative to activities in progress
and records.
Plant Status
The plant operated at approximately 71.5X power from April 1 until
shutdown
on April 10, 1987, for the annual refueling and maintenance
outage.
The outage
was expected to be completed
and the plant ready for
startup early in June.
The fuel shuffle was completed
and
20 control
drives exchanged,
ahead of schedule
on May 11,
1987.
Principal
maintenance activities included the removal
and refurbishing of the
internals of both reactor recirculation
pumps, including seal
and
stuffing box modifications to eliminate the type of mechanical
failures
experienced
in the past.
A containment
vessel
(drywell/wetwell)
integrated
leakage rate test will be conducted at the end of the outage.
No significant operating events
occurred during the reporting period.
0 erations Verifications
The resident
inspectors
reviewed the Control
Room Operator
and Shift
Manager log books
on a daily basis for this report period.
Reviews were
also
made of the Jumper/Lifted
Log and Nonconformance
Report
Log to
verify that there were
no conflicts with Technical Specifications
and
that the licensee
was actively pursuing corrections to conditions listed
in either log.
Events involving unusual
conditions of equipment
were
discussed
with the control
room personnel
available at the time of the
review and evaluated for potential safety significance.
The licensee's
adherence
to Limiting Conditions for Operation
(LCO's), particularly
those dealing with ESF and
ESF electrical
alignment,
were observed.
The
inspectors
routinely took note of activated annunciators
on the control
panels
and ascertained
that the control
room licensed personnel
on duty
at the time were familiar with the reason for each annunciator
and its
significance.
The inspectors
observed
access
control, control
room
manning, operability of nuclear instruments,
and 'availability of on site
and offsite electrical
power.
The inspectors
also
made regular tours of
accessible
areas
of the facility to assess
equipment conditions,
radiological controls, security, safety
and adherence
to regulatory
requirements.
During a tour of the Reactor Building on April 26, 1987, the inspector
observed that 39 apparently
empty gas bottles
had been attached
to
safety-related
cable tray support
PS-4362 in the railroad bay of the
reactor building.
The support held safety-related
cable trays
numbers
S-DIV-2-2200, C-DIV-2-2350, and P-DIV-2-2500.
This appeared
to be
a
violation of Plant Procedure
1.3. 19, "Housekeeping,"
which states
that
gas bottles
are not to be secured
to safety-related
equipment
such
as
conduit, pipes,
etc.
(Enforcement
Item 87-09-01)
Monthl
Surveillance Observation
The inspectors
ascertained
that surveillance of safety-related
systems
or
components
was being conducted in accordance
with license
requirements.
In addition to witnessing
and verifying daily control panel
instrument
checks,
the inspectors
observed portions of the following detailed
surveillance
tests
by operators
and technicians
on the dates
indicated.
PPM 7.4.6.5.3. 1, Standby
Gas Treatment
System Operability Test.
(May 9,
1987)
PPM 7.4.6. 1.2.4,
Containment Isolation Valve and Penetration
Leak Test.
(May 1, 4, 6, 1987)
PPM 7.4.3.7.5.49,
Accident Monitor Instrumentation Calibration for SM-4,
7,
and 8.
(April 22, 1987)
PPM 7.4.8. l. 1.2.6,
HPCS Diesel Generator
Loss of Power Test
(April 26,
27, 1987)"
PPM 7.4.3.3.2.27,
Loss of Power/Loss of Coolant Accident Test.
(April 27, 28, 1987)"
PPM 7.4.0.5. 14,
CAC Valve Operability Test
(May 9, 1987-Data
Review)
"Complex surveillance observation/verification following 3 year
mechanical
maintenance
inspection of the
HPCS diesel
engine.
No violations or deviations
were identified.
S ent Fuel
Pool
and Refuelin
Activities
The inspectors verified that prior to the handling of fuel in the core,
surveillance testing required
by technical specifications
and licensee
procedures
had been completed; verified that during the outage the
periodic testing of refueling related
equipment
was being performed
as
required
by technical
specifications
and that reactor building
ventilation and reactor pool/spent fuel pool conditions were maintained
within the prescribed
technical specification limits; observed
several
shifts of fuel handling operations;
verified that good housekeeping
was
being maintained in the refueling area,
and verified that staffing during
the refueling was in accordance
with technical specifications
and
approved procedures.
An accurate
record was maintained in both the
control
room and
on the refueling bridge of all fuel loading changes.
Fuel bundle location was verified periodically and minimum shutdown
margin checks
were appropriately performed during and after the fuel
shuffle.
Following completion of the fuel shuffle,
a total of 20 control rod
drives were
removed
and replaced with spare drives.
The associated
fuel
cell was unloaded to the storage
canal pri'or to each
exchange.
Appropriate nuclear instrumentation
was available
and periodically
checked.
The inspectors
also observed activities associated
with the removal
and/or the replacement of the reactor vessel
head,
steam separator
and
steam dryer.
No violations or deviations
were identified; however,
the licensee
informed the inspector that
a source
range monitor had been tested
simultaneously with .insertion of a fuel bundle into the core.
This item
was to be reported to the
NRC within thirty days pursuant to 10CFR50.73.
Initial Control
Rod Drive Installation - Code Waiver
Re uest
During the period of November
1982 to'January
1983, the Washington Public
Power Supply System installed
185 control rod drives
(CRDs) in the lower
reactor vessel
head.
At the time of installation, the Supply System did
not have
an
ASME Certificate of Authorization to use the applicable
N-type symbol or stamp permitting installation of the
CRDs.
A bolted
flange is used to connect
each
CRD to the vessel
head.
This deviation from the
ASME Section III Code was identified by guality
Assurance
when the
gA inspector tried to locate the
Code data
N-5 forms
from the initial installation.
An NCR was issued to document
and
disposition this discrepancy.
The State
Department of Labor and Industries
was informed of the
deviation
and asked to accept the installation as-is
based
upon. the
licensee's
gA program at the time of installation.
While the
CRDs were
not installed
by an
ASME certificate holder, the procedure
included all
the planned
and systematic
actions
necessary
to assure
that the
CRDs were
installed correctly and would perform satisfactorily in service.
The
State
was provided
a copy of relevant installation records to support
this contention.
It was the Supply System's
position that the cost
and risks of removing
and reinstalling the 185
CRDs would have
no real benefit and would not
demonstrate
any increased
safety relating to the actual
use of the
CRDs.
Removal, reinstallation
and repeat of the field hydrostatic pressure
test
has the potential for affecting the pressure
boundary integrity as well
as
undue radiation exposure to personnel.
Additionally, during annual
refueling outages,
CRDs will be systematically
removed for inspection
and
refurbishment
on a staggered
ten year maintenance
schedule.
The State
Department of Labor and Industries, with the Energy Facility
Siting Evaluation Council's
(EFSEC) concurrence,
granted the waiver and
accepted
the installation as-is during a Board meeting
on May 19,
1987.
The inspector
examined the licensee's
supporting documentation for the
CRD installation.
The examination of this documentation
indicated the
had been installed
and tested in conformance with the Licensee's
gA
program.
In addition, the initial hydrostatic pressure
test of the
installed
on January
29,
1984 had been
observed
by the
NRC
inspection staff.
Simulator
U
rade
A plant specific simulator
was purchased
and installed prior to plant
operation.
The implementation of simulator changes
identified through
plant preoperational
testing
and operation
lagged behind those in the
plant.
Management
has recognized that the simulator and plant were
diverging and
an upgrade
program for the simulator
has
been initiated.
A
review of the simulator upgrade
program was
made,
and direct observation
of the simulator was conducted
during formal licensed training.
The inspector
noted that the simulator reflects most of the current
hardware
changes
that have occurred in the control
room operating boards.
The remaining changes
appear to be scheduled for completion.
Some
simulations
were observed to be lacking; these
were especially evident in
low temperature,
low pressure
upset events.
The inspector
was
informed that these conditions
had been
noted
and were scheduled for
change
when the model
was revised.
The inspector
noted that the simulator
upgrade
program plan appeared
to
be well thought out and systematically
implemented.
The objective of the
plan is to have
a certifiable simulator available for licensed operator
training and examination within a four year time frame.
The plan is
designed to meet or exceed
the Electric Power Research Institute (EPRI)
standards
and will comply with ANSI/ANS 3.5-1985 "Nuclear Power Plant
Simulator for Use in Operator Training", the proposed revision of 10 CFR 55,
and Regulation
Guide 1. 149.
Overall, the program for matching the simulator to the plant appeared
to
be on schedule
and progressing
in a controlled and well thought out
manner.
Although the schedule
is not expected to be completed until mid
1990, the training department staff expressed
optimism that the changes
can
be effective before that time.
No violations or deviations
were identified.
8.
Radiolo ical Protection
The inspectors periodically observed radiological protection practices
to
determine whether the licensee's
program was being implemented in
conformance with facility policies
and procedures
and in compliance with
regulatory requirements.
The inspectors verified that health physics
supervisors
and professionals
conducted
frequent plant tours to observe
activities in progress
and were generally
aware of significant plant
activities, particularly those related to radiological conditions and/or
challenges.
ALARA consideration
was given'ach job that was
done during
the refueling outage
and was discussed
frequently at the daily Manager'
meeting, particularly with respect to the big jobs such
as the reactor
recirculation
pumps
and control rod drive replacements.
The various Radiation Work Permits
(RWPs) in use at principal entry
points such
as those for the refueling deck,
the dry well and the wet
well were examined
and found to consider the appropriate
elements.
The
RWPs generally referred the wor ker to the Health Physicist for the latest
specific activity levels in the area.
There appeared
to be sufficient
Health Physicists available to support the activities in progress.
Workers appeared
to have the proper monitoring equipment
and dosimeters
for their area of work, including high radiation areas.
High radiation
areas
were posted
and radiation warning lights were in use where
appropriate.
Personnel
exiting radiation/contamination
zones
were
observed
to conduct frisks in accordance
with the licensee's
procedure.
The inspector
questioned
the propriety of the following practices that
were observed
on the refueling deck.
a.
A radiation area monitor was not in use
on the refueling bridge.
The responsible
Health Physicist agreed that this was
a prudent
policy and promptly located
a portable monitor
on the bridge.
Later, licensee
management
stated that
a permanent
monitor will be
maintained
on the bridge
and that
a Plant Modification Request to
'ffect
this change
had been issued.
On April 26, 1987, at approximately
1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br />,
the Health Physicist
set his portable survey meter
on the refueling bridge deck and
assisted
operators
in the removal
and baggin'g of an underwater light
rather than surveying the light as it was being removed from the
pool.
A subsequent
survey, after the inspector questioned
the
activity, did not indicate
any significant radiation levels.
On May
7, 1987, at approximately '1400 hours0.0162 days <br />0.389 hours <br />0.00231 weeks <br />5.327e-4 months <br />,
Operations
personnel
removed
a
television
camera
from the pool that had been
used during control
rod unlatching operations without having it surveyed
as it was being
removed.
The Health Physicist subsequently
performed
a survey
following questioning
by the inspector
and determined that the
radiation level
was less
than
10 mrem/hr.
Failure to perform
surveys of equipment
upon removal
from the fuel pool appears
to be
a
violation of regulatory requirements
and the
licensee's
Radiation Work Permit 2-87-00138 for refueling bridge
activities.
(Enforcement
Item 87-09-02)
C.
At the time of the inspection
on April 26,
1987, portable monitors
AMS-3 RB4A (CAM) and RRA-RIS-1/AD-03 (ARM) had not been source
checked
since April 13,
1987.
The failure to source
check AMS-3
RB4A was identified to the licensee
as
a violation of Plant
Procedure
11.2.24. 1 which required the
CAM to be source
checked
weekly.
A subsequent
source
check
showed the monitors to be
functioning properly.
(NOTE: the
ARM need only be checked
by Plant
Procedure
when placed in operation,
even though it was being checked
weekly).
(Enforcement
Item 87-09-03)
9.
Monthl
Maintenance
Observation
and Post Maintenance
Restoration
Portions of selected
safety-related
systems
maintenance activities were
observed.
By direct observation
and review of records the inspector
determined
whether
these activities were consistent with LCOs;, that the
proper administrative controls
and tag-out procedures
were followed; open
system controls were in effect as appropriate;
and that equipment
was
properly tested
before return to service. It was specifically observed
that workers tended to clean
up after themselves
at the end'f the day
and were quite thorough in restoring the area at the end of the job.
The
inspector
also reviewed the outstanding job orders to determine if the
licensee
was giving priority to safety related
maintenance
and verify
that backlogs
which might affect system performance
were not developing.
The following maintenance activities were observed:
Repair of Reactor Building Security Door (AU5703)
I
'O'ing seal
replacement
on
FCV 'A'ydraulic packages
(AU8120)
Replacement
of reactor pressure
scram switch
(AU7839)
Preventive
maintenance
on Condensate
Pump 2-B 4160V breaker
Refurbishment of Main Turbine Governor Valve Hydraulic Actuators
(AU8423)
Preventative
Maintenance
on
RRC Flow Control Valve 'A'ryquil
system pressure
switch HY-PS-A2125
Installation of Emergency Utility for Auxiliary Building (AU8078)
Repairs
on Fuel Handling gripper light
(AV1114)
Preventative
maintenance
on 480 volt breaker per
PPM 10.25.2
Diesel generator
1 modification to ensure
engine to generator
alignment
(AU9506;
PMR 86-329-0)
Diesel generator airstart motor rebuild and replacement
(AU9759)
Diesel generator
18 month inspection
(PPM 7.4.8. 1.2. 14)
HPCS Diesel generator - replacement of brushes
(PPM 10.25.49)
HPCS Diesel generator - welding for shield grating
(AU9032)
HPCS Diesel Generator - 3 year maintenance
inspections'AU9322;
10.20.14)
HPCS Diesel generator - inspect main bearings
(AU9319)
RWCU High Differential Flow transistor failure (AV1146)
No violations or deviations
were identified.
10.
Mana ement Involvement
Examination of "Housekeeping
Reports"
by Area Coordinators for the months
of January-March
1987 indicate that assigned
actions
were being
accomplished
in a timely manner.
Explanation were provided for any
action that was outstanding for more than 4 weeks, =including an
indication of when corrective action was expected to be completed.
The inspectors
continued to notice the presence
of plant and corporate
management
in the plant evaluating 'plant conditions throughout the
refueling outage.
The inspector
also
accompanied
the Assistant Plant
Manager
on one of his tours.
He stated that it was his policy to tour a
different area at least weekly.
Log book entries
indicated that
operations
management
was frequently on site during the back shifts and
on weekends.
Overall, it is the inspectors'erception
that there
has
been
a
substantive
involvement of all levels of Corporate
and Plant management
at the
WNP-2 facility.
ll.
En ineered Safet
Feature Verification
The inspector verified the operability of the Standby
Gas Treatment,
Standby Liquid Control, High Press
(HPCS), Diesel Generator
2,
and Diesel Starting Air Systems
by performing a walkdown of the
accessible
portions of the systems.
The inspector, confirmed that the
licensee's
system lineup procedures
matched plant drawings
and the as-
built configuration,
and verified that valves were in the proper
position,
had power available,
and were locked as appropriate.
The
licensee's
procedures
were verified to be in accordance
with the
Technical Specifications
and the
FSAR.
No violations or deviations
were identified.
12.
ualit
Assurance
Pro
ram for Measurin
and Test
E ui ment
The inspectors
examined the licensee's
program for the control of
measuring
and test equipment
(M8TE). Plant records
include the
"Calibration Report" which lists test instruments
used
and certifies
them
to have calibrations traceable
to the National
Bureau of Standards
(NBS).
Plant Procedure
1.5.4,
"Control of Measuring
and Test Equipment-Transfer
Standards,"
defines the measures
established
to assure that tools,
instruments,
and other measuring
and test devices
used in
activities affecting quality will be properly controlled, calibrated,
and
adjusted at specified periods to maintain accuracy within the necessary
limits.
The proper
use
and control of M&TE was verified by the inspectors
during
the observations
of maintenance
and surveillance activities discussed
elsewhere
in this report.
The inspectors
frequently visited the Tool
Crib during the outage
and verified that equipment controls were being
implemented in accordance
with program requirements.
The licensee's
latest audit of this area (Surveillance
Report
No.
2-86-157,
conducted
January
16-23,
1987) identified eight deficiencies
that have
been appropriately corrected.
The Surveillance
Report was
made
mandatory reading for all maintenance
personnel.
The gA/gC Department's
observation
program includes
M&TE control
as
one of the attributes to be
inspected.
Of 50 activities observed
during the refueling outage,
only
two minor deficiencies
were identified in the area of M&TE control.
The
Plant Manager informed the inspectors
that
he had personally challenged
the Maintenance
Manager to do better by reducing the number of identified
deficiencies'he
observation
program appeared
to indicate
an improving
trend.
No violations or deviations
were identified.
Local
Leak Rate Testin
The containment local leak rate test
(LLRT) program was reviewed by the
inspector.
Through records
review, direct observation,
and independent
calculation,
the program implementation
was ascertained.
Records for
leak rate tests
on boundary penetrations
and valves were reviewed for
frequency
and acceptability,
and to insure that actions required
by the
Technical Specifications
were being met.
For selected
maintenance
outages,
and valves were reviewed'o verify that repairs
were preceded
and followed by LLRTs.
Test equipment
used during the
LLRTs was also reviewed to insure that the equipment was.within
calibration during the test.
During the refueling outage,
the inspector witnessed
the testing of
valves
RFW-V-32B and CIA-V-31B and penetrations
X-73e and X-lf.
A review of testing performed during the previous refueling outage
was
also reviewed.
The total leak rate
was calculated
from the results of
the individual leakage
rates
and compared with the data furnished
by the
licensee's
staff.
The inspector
concluded that the local leak rate
testing program was in conformance with the procedure.
No violations or deviations
were identified.
BWR Radiolo ical Controls for Dr
elis Durin
S ent Fuel
Movements
An inspection
was performed to evaluate
the radiological controls for
access
to and work in the drywell during spent fuel movement.
During
certain spent fuel movement conditions,
very high dose rates
can exist in
the drywell.
A transfer of a spent fuel bundle from the reactor pressure
vessel
(RPV) to the spent fuel pool requires
the fuel bundle to.be lifted
over the
RPV flange, over the drywell and through the "fuel cattle chute"
area,
into the spent fuel pool.
Normal fuel transfers
cause
a high
transitory dose rate in the drywell for short periods of time.
A loss of
power or malfunction could cause
higher dose rates in the drywell, and
a
dropped fuel bundle in the drywell region could cause
very large dose
rates in the upper drywell region.
The licensee's
written procedures,
training programs,
and organization interfaces
were examined.
The fuel
cattle chute
was also reviewed to determine
the adequacy
of the shielding
used to reduce the exposure
in the drywell.
The inspector
determined that the licensee
was aware of the potential
problem with personnel
exposures
during fuel handling and plant
procedures
administratively controlled access
above the 548'levation of
the drywell.
Fuel handling procedures
require that the radiation
protection health physics
group contact the control
room and refueling
supervisor
and that the upper drywell be cleared of personnel prior to
commencement
of fuel handling operations.
Installed radiation monitors in
the drywell above the 548'evel
provide an alarm,
and warning lights
activate,
on high radiation level.
Through conversations
with members of the engineering
department staff,
the inspector
determined that the fuel chute
was site fabricated
from
stainless
steel
and
has
a bottom thickness of approximately
9 j./2 inches.
The lower sides
are approximately
6 inches thick, and the
gamma reduction
factor is approximately
1000 to the drywell.
No weakness
were identified in the licensee's
approach to recognizing the
conditions or implementing administrative corrective measures.
The
installed hardware,
both radiation monitoring equipment
and the fuel
chute,
appeared
to be adequate
to meet the expected conditions.
Operation of the potential
high radiation area warning lights at the 548
foot elevation inside the drywell was verified by the inspector during
fuel handling operations.
0 erator Attentiveness
to Duties
The
NRC order to shut
down Peach
Bottom j. and
2 was discussed
with
licensee
management
and security personnel.
While the licensee did not
believe it would happen at WNP-2, all licensed
personnel
and others
were
informed via an interoffice memorandum
dated April 7, 1987, that if it
happened,
at the very least,
the offender would be removed
from licensed
duties.
The memorandum
contained
several
attachments
directly related to
the order.
Discussions
held with plant security personnel
indicated that the
security force generally alerted plant personnel
when corporate
management
or the
NRC came
on site during the off-shifts. It was not
clear whether the
NRC notification pertained to non-routine
NRC
inspectors
requiring Shift Manager approval
or others.
Nevertheless,
the
10
licensee
stated that this practice will be stopped for other than those
cases
where plant management
approval is required for entry.
According to the Plant Manager,
the plants in Region
V will be developing
a generic plan to assure
that management
and the
NRC will get
a realistic
review of operations
personnel
behavioral
practices.
The inspector
was
assured
by Corporate
Management that "underground" notification practices
would not be tolerated.
Security management
issued
a memorandum to this
effect to the security force.
Also, security issued
a memorandum
pertaining to "attentiveness
to job accomplishments".
These
memoranda
were considered
by the inspectors
to adequately
address
the subject
issues.
In addition, the Managing Director discussed
the Peach
Bottom event
and
its repercussions
on the nuclear industry .in the employee's
"Newsline"
tabloid.
The inspectors
performed
a number of off-shift inspections
during the
report period, including nights (both swingshift and graveyard)
and
weekends.
No buffoonery or other lack of attentiveness
to duty was
observed
in operations
or security.
16.
Overtime
During a review of employee
over time, the inspector
compared
the use of
"Deviation of Overtime Restrictions"
(Attachment
A of procedure
1.3.27,
"Overtime Control" ) to the requirements
of Technical Specification 6.2.2.f and procedure
1.3.27.
Twenty deviation sheets
generated
since January
1987 were reviewed.
Three
of the Deviation Approval Sheets
involved five individuals not being
~ approved to exceed
the overtime limitations of either working greater
than
16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period or working greater
than
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in a
48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period.
The inspector determined that on February
13,
1987
a
deviation approval
was signed
by the Shift Manager to allow three
mechanics
to exceed
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period while machining tools
needed for fuel channeling
inspection.
Two of the mechanics
apparently
worked
17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br />
and 17.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> in a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period without approval.
On
February
13,
a Shift Technical Advisor worked
16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
period.
On the next day, February 14,
he apparently
worked ll hours
providing technical
engineering
support to operators
performing
dechanneling
of spent fuel without authorization to exceed
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in a
48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period.
On March 23, two mechanics
were granted
approval
by the Shift Manager to
exceed
16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in a
24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period to complete welding on
a Reactor
Drain Line (RFH-P-B).
Both men worked 17.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />
on March 23,
and the next day March 24, worked
a normal eight hour period apparently
exceeding
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period without approval.
In addition, the inspector
noted that
on February 25, the Shift Manager
authorized
two mechanics
to exceed
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period to
11
provide startup mechanical
support,
which they .did.
The inspector
noted
that Technical Specification 6.2.2f.3 allows the Plant Manager,
his
deputy, or higher management
to approve
exceeding
the overtime limit for
safety-related
work, and procedure
1.3.27 allowed the Plant Manager,
the
Assistant Plant Manager,
the Operations
Manager,
the Assistant Operations
Manager, or the Shift Manager to approve deviations to the overtime
procedure.
The procedure
did not differentiate between safety or non
safety-related
work. In reviews of the workers'ime reports,
the
inspector
could not determine if the overtime
had
been
performed doing
safety related tasks.
The inspector also noted that the appendix to
1.3.27 ("Request for Deviation to Overtime Restriction" ) did not
recognize
the proviso of Technical Specification 6.2.2f.3
and procedure
1.3.27-A.3,
"A break of at least eight hours shall
be allowed between
work periods,
including shift turnover time."
From the time reports,
the
inspector
was unable to determine if this portion of the procedure
had
been violated; however, it was discussed
with management
along with the
above noted procedure deficiencies.
This issue is unresolved
pending
further evaluation during the next inspection period (87-09-04).
Licensee
Event
Re orts
The resident
inspectors
reviewed the following report and supporting
information on site to verify that licensee
management
had reviewed the
event, corrective action
had
been taken,
no unreviewed safety questions
were involved,
and
violations of regulations
or Technical Specification
conditions
had been identified.
Post Tri
Review - Licensee
Event
Re ort 87-02
0 en
On Narch 22,
1987, the reactor
was manually scramed following a trip of
both feedwater
pumps.
Post
scram operator error(s)
caused
the reactor
pressure
vessel
(RPV) to be overfilled on three occasions.
Details of
the trip were included in inspection report 397/87-04.
The
NRC was
notified of the event in a timely manner;
however, plant conditions
changed
(reactor vessel
level increased)
after the shift manager
assessed
the status of the plant and began his notification to the
NRC.
A later
review of the plant conditions
from the licensee
event report
(LER)
indicated that at the time of NRC notification the reactor vessel
level
was higher
than reported.
The resident
inspector interviewed the shift
manager
and confirmed that
he was
unaware of the change in plant
conditions
when the notifications were made
and later he did not perceive
the high
RPV levels to be
a "worsening plant condition" as discussed
in
The Shift Manager did not enter the site emergency
plan
because
he felt the situation did not meet the requirements
of the plan
implementing procedures.
The inspector later verified that the level
2
actuation for a short period
had
been identified by the licensee
as being
an expected
operational
result for a loss of feedwater
pump scram
and was
excluded
from the emergency
plan.
The inspector
was told that the licensee
is in the process
of preparing
a
report dealing with high
RPV levels
and flooded steam line conditions.
The report is to include
an analysis of liquid discharge
through the
including
a safety assessment
of off normal scenarios
involving these
plant conditions.
12
The inspectors
attended
the Followup Review Committee meetings
and the
final Plant Operations
Committee
(POC) meeting
on the trip for the
purpose of assessing
the effectiveness
of the post trip review.
The
Followup Review Committee
was
composed of the members specified
by Plant
Procedure
1.3.5,
"Reactor Trip and Recovery".
The Chairman of the
Committee
by procedure
is the cognizant Shift Engineer/Shift Technical
Advisor.
The Committee
appeared
to concentrate
on the time line for the
event
and the initial cause of the event with little interest in the
resolution of the .human errors that resulted in the overfilling of the
RPV.
The Committee did not appear to have
any questions of the operators
and Shift Management
who were in attendance.
Shift Management
made
some
explanations
of crew member actions,
but the Committee
members did not
appear to take any actions
towards understanding
the operators'ehavior
or understanding
of the situation.
Specifically, the Nuclear Safety
Assurance
Group
(NSAG) member did not appear to address
the operator
performance
problems,
nor did the guali.ty Assurance
member in attendance
take
an active roll in insuring that operator performance
issues
were addressed
by the Committee.
However, subsequently
the Plant Manager did directly
interview the operators
and ascertained
the pertinent factors pertaining
to personnel
errors.
It appeared
to the inspector that the Committee's
main objective
was the
composure
and refinement of the licensee
event report (LER) rather than
the determination of factors which affected operational
decisions.
However, the Committee's report did recommend
several
hardware
changes
and the
need for additional operator training.
Management
agreed,
during
the exit interview, to assess
overall
Committee effectiveness,
'including
leadership.
Items scheduled
to be implemented in the future include: determining
whether to deenergize
nonclass
loads fed from class I power only in the
event of a loss of power; evaluation of changing
some level
2 trips to
level 1; evaluation of an automatic trip of .the condensate
booster
pump
on a high
RPV level; and increased
operator training on feedwater
upsets
and high
RPV level conditions.
These
items will be evaluated
as part of
the
LER corrective action followup.
No violations or deviations
were identified.
18.
Licensee Actions
On Previous
NRC Ins ection Findin
s
The inspectors
reviewed 'records,
interviewed personnel,
and inspected
plant conditions relative to licensee
actions
on previously identified
inspection findings:
(Closed) Inspector
Followup Item 86-06-05:
Amber warning lights on
panel.
The independent
Nuclear Assurance
Group
(NSAG) identified several
warning
lights on the control panel that had the potential to distract or provide
confusion to the control
room operator.
NSAG noted that (1) the lights
should
be removed or (2) they should
be upgraded
and procedures
should
be
reviewed to include these
special
purpose lights.
The inspector
reviewed
NSAG correspondence
that showed these lights to have
been deactivated;
0
13
they have
been
scheduled
to be removed during refueling outage
3 and were
being tracked
on the Plant Tracking List (PTL).
Due to the removal
from
service, plant procedures
were not modified to include these lights.
This item is considered
closed.
(Closed)
Inspector
Followup Item 86-09-01:
Examine
NSAG identified
operational
event reports that should
be included in procedure
reviews.
The inclusion of lessons
learned
from industry operating experience
reports
appears
to be
a subject for consideration
in the review process.
In discussion with NSAG members,
the inspector learned that the operating
experience
reports
are reviewed
by NSAG and passed
on to the training and
operating groups,
and when conditions are identified that specifically
apply to this unit, procedures will be changed to reflect the learned
experience.
NSAG members
were concerned that the biannual
review of procedures
was
being performed without benefit of defined formal review criteria.
The
inspector
reviewed
a letter sent from the
NSAG group to plant management
expressing this lack of established
procedure
review criteria. This item
will be examined during a later inspection.
(87-09-05)
19.
Ph sical Securit
Security activities were observed for conformance with regulatory
requirements,
implementation of the site security plan,
and
administrative
procedures
including vehicle and personnel
access
screening,
personnel
badging, site security force manning,
and protected
and vital area integrity.
Exterior lighting was checked during backshift
inspections.
No violations or deviations
were identified.
20.
Review of Periodic
and
S ecial
Re orts
The licensee's
monthly operating reports for March and April 1987 were
examined
and found to reflect the inspector's
observations
for those
months.
The "1986 Radiological
Environmental Monitoring Program
Annual
Report"
was
examined.
No anomalies
were identified.
21.
Mana ement Meetin
The inspectors
met with the Plant Manager and/or his assistant
approximately weekly during this period, to discuss
inspection finding
status.
On May 21,
1987, the inspectors
met with the Plant Manager
and
members of his staff to discuss
the inspection findings during this
period.