ML18011A957
| ML18011A957 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 07/07/1995 |
| From: | Elrod S, Darrell Roberts, Verrelli D NRC Office of Inspection & Enforcement (IE Region II) |
| To: | |
| Shared Package | |
| ML18011A955 | List: |
| References | |
| 50-400-95-10, NUDOCS 9507140315 | |
| Download: ML18011A957 (23) | |
See also: IR 05000400/1995010
Text
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UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W., SUITE 2900
ATLANTA,GEORGIA 303234199
Inspectors:
.Ero
enior
e ident Inspector
b rts
Resident
Inspector
Approved by:
D.
er elli, Chief
Reactor Projects
Branch lA
Division of Reactor Projects
Report No.:
50-400/95-10
Licensee:
Carolina
Power and Light Company
P. 0.
Box 1551
Raleigh,
NC 27602
Docket No.:
50-400
Facility Name:
Harris
1
Inspection
Cond
c
e
May 7
J ne 10,
1995
License No.:
D
e Signed
Da e
igned
7 i'll~
ate Signed
SUMMARY
Scope:
This routine inspection
was conducted
in the areas of operational
safety,
maintenance,
surveillance,
engineering activities,
and plant support.
Numerous facility tours were conducted
and facility operations
observed.
Backshift tours
and observations
were conducted
on May 10, ll, 14,
15,
16,
25,
27; 29,
and June 6, 8,
and 9,
1995.
Results:
0 erational
Safet
Operations
were generally performed well.
One violation involving .the power
for a,power operated relief valve demonstrated
the need for vigilance during
operations,
paragraph 3.b.(2)
Maintenance
Work performance
was satisfactory with proper documentation of removed
components
and independent verification of the reinstallation.
The inspectors
identified one violation involving maintenance
implementation of the site-wide
chemistry control program,
paragraph 4.a.(2).
9507140315
950707
ADOCK 05000400
9
En ineerin
Activities
Engineering activities were performed adequately.
~PP
1
Plant housekeeping
and component material condition was satisfactory.
The
licensee's
adherence
to radiological controls, security controls, fire
protection requirements,
emergency
preparedness
requirements
and
TS
requirements
in the plant support
area
was satisfactory.
REPORT DETAILS
PERSONS
CONTACTED
Licensee
Employees
D. Batton,
Manager,
Work Control
D. Braund,
Manager,
Security
J. Collins, Manager, Training
J.
Dobbs,
Manager,
Outages
- J. Donahue,
General
Manager,
Harris Plant
R. Duncan,
Manager,
Technical
Support
H.
Hamby,
Manager,
Regulatory Compliance
- H. Hill, Manager,
Nuclear Assessment
- R. Prunty,
Manager,
Licensing
& Regulatory
Programs
- W. Robinson,
Vice President,
Harris Plant
- G. Rolfson,
Manager,
Harris Engineering
Support Services
- C. Rose, Acting Hanager,
Maintenance
H. Smith,
Manager,
Radwaste
Operation
T. Walt, Manager,
Regulatory Affairs
B. White, Manager,
Environmental
and Radiation Control
"A. Williams, Manager,
Operations
Other licensee
employees
contacted
included:
office, operations,
engineering,
maintenance,
chemistry/radiation control,
and corporate
personnel.
NRC Personnel
T. Decker, Chief, Radiological Effluents and Chemistry Section,
RII
R. Carrion, Radiation Specialist,
RII
- S. Elrod, Senior Resident
Inspector,
Harris Plant
- D. Roberts,
Resident
Inspector,
Harris Plant
"Attended exit interview
and initialisms used throughout this report are listed in the
last paragraph.
PLANT STATUS AND ACTIVITIES
'a ~
Operating Status of the Plant Over the Inspection
Period.
The plant continued in power operation
(Mode I) for the duration
of this inspection period.
The licensee
reduced unit power to 97
percent
on Hay 14 for moderator temperature coefficient testing
and to 80 percent
on June
9 for turbine valve testing.
The unit
ended the period in day 213 of power operation
since startup
on
November 8,
1994.
b.
Other
NRC Inspections
or Meetings at the Site.
R. Carrion, Radiation Specialist,
NRC RII, was
on site from
May 15-19 conducting
an inspection in the area of radiological
effluents
and chemistry.
He was accompanied
during part of the
inspection
by T. Decker, Section Chief, Radiological Effluents
and
Chemistry Section,
NRC RII.
They conducted
an exit meeting
on
May 19 and their findings will be documented
in IR 400/95-09.
OPERATIONAL SAFETY
'a ~
Plant Operations
(71707)
(1)
Shift Logs
and Facility Records
The inspector
reviewed
numerous
records
and discussed
various entries with operations
personnel
to verify
compliance with the
TS and the licensee's
administrative
procedures.
In addition, the inspector
independently
verified clearance
order tagouts.
The inspectors
found the logs to be legible and well
organized,
and to provide sufficient information on plant
status
and events.
The inspectors
found clearance
tagouts
to be properly implemented.
The inspectors identified no
violations or deviations
in the shift logs
and facility
records
area.
(2)
Facility Tours
and Observations
Throughout the inspection period, the inspectors
toured the
facility to observe activities in progress,
and attended
several
licensee
meetings to observe
planning
and management
activities.
Inspectors
made
some of these
observations
during backshifts.
During these tours, the inspectors
observed
monitoring
instrumentation
and equipment operation.
The inspectors
also verified that operating shift staffing met
TS
requirements
and that the licensee
was conducting control
room operations
in an orderly and professional
manner.
The
inspectors additionally observed
several shift turnovers to
verify continuity of plant status,
operational
problems,
and
other pertinent plant information.
Licensee
performance
in
these
areas
was satisfactory.
During
a facility tour on May 26, the inspector learned of a
condition which had rendered
the "C" SG
PORV inoperable the
previous day.
Discussions with licensee staff members
indicated
a lack of clarity concerning
when the valve
became
and whether or not
a TS
LCO had
been violated.
The inspector
and the licensee. pursued this subject,
which
is discussed
further in paragraph 3.b.(2).
A violation identified in the facility tours area is
discussed
in paragraph 3.b.(2).
Effectiveness
of Licensee
Control in Identifying, Resolving,
and
Preventing
Problems
(40500)
Adverse Condition and
Feedback
Reports
(ACFRs) were reviewed to
verify TS compliance,
that corrective actions
and generic
items
were identified,
and that items were reported
as required
by
10
CFR 50.73.
Inspectors
reviewed
ACFRs documenting
incidents
involving RCA entry without proper dosimetry
and containment
isolation valves which were determined
by plant personnel
to have
been inoperable
beyond the four hours allowed by TS 3.6.3:
(1)
ACFR 95-1359
was generated
when
an individual entered
the
RCA without proper dosimetry.
The inspectors
reviewed the
licensee's
corrective actions following this incident
and
found them appropriate.
This. subject is discussed
further
in the plant support section,
report paragraph
6.b.
(2)
ACFR 95-1428 discussed
hydraulically operated
containment
isolation valve 1HS-62, the "C" SG
PORV.
On Hay 25,
was determined to have
been inoperable for about
30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />
when plant personnel
discovered
a blown fuse in its
hydraulic
pump circuit.
Events leading to this condition
are discussed
in the following paragraphs.
At approximately 9:30 a.m.
on Hay 24,
a non-licensed
operator noticed that there were no indicating lights
illuminated at
HCC breaker cubicle IA31-SA-14D, from which
the
1HS-62 hydraulic oil pump was powered.
The day before,
the operator
had observed
the green light to be
illuminated - indicating that the hydraulic oil pump control
power was then available
and the
pump was off.
The operator
replaced
the green
and red light bulbs,
however, neither
light energized.
The operator then attempted to replace the
blue indicator bulb but found it broken off at the base.
He
then initiated
a work request for maintenance
to extract the
bulb base
from the socket
and replace
the bulb.
An on-shift
SRO in the work control center reviewed
and
approved the work request.
He determined,
from lack of
associated
alarms in the main control
room, that 1HS-62's
operability was unaffected
by the
HCC indicating light
condition.
He initially treated the deficiency as
a light
bulb problem,
and assigned
a low maintenance priority to the
work request.
A day later,
on Hay 25,
a different
SRO in the work control
center determined that the light bulb deficiency might
indicate
an operability problem
and requested
a maintenance
investigation.
A few hours later,
maintenance
personnel
found that the control
power fuse
was blown.
They replaced
it and the blue light bulb within four hours
(approximately
3:30 p.m.).
Licensee
personnel
blamed the blown fuse
on the
broken light bulb, which probably overheated.
The blown
fuse
had deenergized
the hydraulic oil pump.
This pump was
necessary
to maintain hydraulic accumulator oil pressure
high enough to ensure that the valve could be closed from
the main control
room.
Following additional review, the
licensee
concluded that,
due to the loss of hydraulic
pump
power, valve IHS-62 had
been
from Hay 24,
when
the extinguished
HCC lights were initially discovered,
until
approximately
30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> later on Hay 25 when the control
power fuse
was replaced.
Technical Specification 3.6.3 required that each containment
isolation valve specified in plant procedure
PLP-106,
Technical Specification
Equipment List Program,
shall
be
operable with isolation times less than or equal to required
isolation times.
It further required that with one or more
of the containment isolation valves inoperable,
maintain at
least
one isolation valve operable
in each affected
penetration 'that was open
and complete
one of the following
four actions:
~
Restore
the inoperable valve(s) to operable status
within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />,
or
~
Isolate
each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
by
use of at least
one deactivated
automatic valve
secured
in the isolation position, or
~
Isolate
each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
by
use of at least
one closed
manual
valve or blind
flange, or
~
Be in at least
Hot Standby within the next
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
and
in Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
The closed
system in which the steam generator
PORVs were
located
(the main steam
system)
was considered
to be the
operable isolation valve for purposes
of complying with the
first action statement
to "maintain at least
one isolation
valve Operable..."
Although IHS-62 remained
shut during the
Hay 24-25 incident, it was not considered
"deactivated"
as
required
by the second of the four subsequent
LCO action
statements.
This was because
even with no control
power
available to the hydraulic oil pump, oil pressure
remaining
in the valve accumulator
could continue to operate
the valve
in either direction with power available to the solenoids,
as
was the case
on Hay 24-25.
Additionally, none of the
remaining
TS actions
were taken within allowed timeframes.
Because
1HS-62 was inoperable for greater
than
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />,
and
only one of the two required
LCO action requirements
were
satisfied,
the licensee's
actions
on Hay 24-25 were contrary
to the requirements
of TS 3.6.3.
This is Violation 400/95-10-01,
Inoperable for Greater than
Four Hours.
Subsequent
to the violation, engineers
on
an event review
team discovered
a construction-era
design deficiency which
rendered
the main control
room annunciator
inoperable for
loss of 1HS-62 hydraulic control power.
This deficiency was
not readily recognizable
by plant operators
and strongly
contributed to the licensed
SRO's initial determination that
the valve was operable
on Hay 24.
The design deficiency,
which did not affect the other two
PORVs, is further
discussed
in the engineering
section,
report paragraph
5.b.
On June
1,
because
of the design error, the licensee
again
declared
the valve inoperable
and took appropriate
TS
compensatory
actions.
At the close of the inspection
period, the
TS 3.6.3 action statement
requirements
were
being met by the
"C" SG
PORV block valve,
being
shut.
The licensee
planned to correct the design deficiency
before returning
The inspectors
concluded that the design
issue discussed
in
paragraph
5.b of this report
was the root cause of the
violation.
However, the inspectors
also concluded that
a
more questioning attitude
on Hay 24 by plant operators
might
have prevented
the violation.
The licensee's
Nuclear Assessment
Section
completed three
assessments
this month:
~
H-SC-95-01,
HNP Security Assessment,
~
H-HA-95-01, Work Request
& Authorization
(WR&A) Review
Process,
and
H-SP-95-02,
Harris Material Receipt Inspection Activities.
The licensee's
Site Management
Team completed
a self-assessment
of
independent verification this month.
The inspectors
reviewed the assessments
and concluded that the
assessments
were thorough
and resulted
in substantive
findings.
The inspectors identified no violations or deviations .in the
Nuclear Assessment
area.
One violation was identified in the operational
safety area.
HAINTfNANCE
a ~
Haintenance
Observation
(62703)
The inspector
observed
or reviewed maintenance activities to
verify that correct equipment clearances
were in effect; work
requests
and fire prevention work permits were issued
and
TS
requirements
were being followed.
Haintenance
was observed
and
work packages
were reviewed for the following maintenance
activities:
(1)
WR/JO 95-AFRKl, Replace
Locking Pin Cylinder on Personnel
Air Lock Door.
WR/JO 95-AFRLl, Troubleshoot
and Replace
Control
Power
Fuse
for PAL Door.
On Hay ll, when
a containment
PAL door operator closed the
outer door following a routine containment entry,
a locking
(load-bearing)
pin and its attached cylinder broke off from
the door's hydraulic control unit and fell to the floor.
The locking cylinder, which measured
about
5 inches
long and
1 inch in diameter,
did not physically provide
a door
locking function by design,
but was
a key component in two
other aspects
of the
PAL door's operation.
With the outer
door closed, this locking cylinder would engage
a limit
switch which would then provide the permissive for the inner
door to be opened.
Additionally, the engaged limit switch
would send
a signal to the door's electro-hydraulic control
system indicating that the door closing sequence
was
complete.
This signal
would shutoff the associated
hydraulic pump.
The broken locking cylinder alone did not render the
PAL
door inoperable.
However,
because
the cylinder had not
engaged
the limit switch, the hydraulic
pump continued to
operate
in the absence
of the shutoff signal, eventually
blowing a fuse in the hydraulic control circuitry.
This
removed
PAL door power and the ability to verify the
presence
of an interlock designed to prevent
both
PAL doors
from opening simultaneously.
Since the interlock
verification was
a TS surveillance requirement,
operators
appropriately declared
the door inoperable
and entered
a 24
hour shutdown action statement.
Three maintenance
work
tickets were subsequently
generated
to repair the door.
The inspector
observed
two of the three maintenance
jobs
associated
with returning the
PAL door to operability.
These
included replacing the hydraulic cylinder unit
(complete with a new locking pin and cylinder)
and replacing
the blown control
power fuse.
The third job, for which the
inspector reviewed
a completed
work ticket, involved
adjusting another limit switch.
This switch caused
the
locking cylinder to engage
before another locking device
could clear
a path for it during the earlier door closing
sequence.
When the locking cylinder impacted this device
during that sequence,
the load-bearing
pin yielded under
excessive
stresses.
The licensee
completed the work, successfully
tested
the
door interlock,
and exited the
LCO action statement within a
few hours.
The inspector
observed
good workmanship during
the jobs
he observed.
Plant personnel
adequately
documented
all maintenance
on the work tickets.
WR/JO 94-AHHN1 Change Trim in 1CS-231
Per
PCR-6796.
purpose
was to vary charging flow to maintain
programmed pressurizer level.
The original valve internals
were oversize
- resulting in poor control response
to
pressurizer
level transients
and inability to maintain
a
steady charging flowrate below about
85 gpm.
Operators
had
routinely operated with the valve in manual control.
The inspector
observed
the mechanics
disassemble,
modify,
and reassemble
valve 1CS-231.
Worksite activities inspected
included worksite preparation,
radiological control
practices,
conduct of maintenance,
and
use of calibrated
tools when called for.
Since this valve routinely had
reactor coolant flowing through it, the worksite was
prepared
as
a highly contaminated
area with a roped-off
boundary, stepoff pad, containers for used
anti-contamination
clothes,
and covered floor area.
The
area
was adequately
sized for work and included laydown
space for both old and
new parts.
The valve components
were
handled with care.
Workmanship, radioactive work practices,
and torquing techniques
were excellent.
The inspector
determined that both torque wrenches
used
were current in
the calibration program.
The inspector
observed
a quality
control person performing
a thorough post-maintenance
visual inspection of boundary leakage.
No leakage
was
found.
WR/JO 94-AHHNl required that the shop erect
enclosure
over radiation monitor 3502A located
under this
valve to prevent
something
being dropped
on it and causing
entry into a
LCO.
The inspector identified that the
scaffolding was not built.
Inquiry found that the
supervisor
had evaluated this requirement
and decided that
a
scaffold was not necessary
because
the deck grating floor
between the two items, in conjunction with the temporary
covering placed
on the deck grating floor for this job,
performed the
same protective function.
The inspector
concluded that the supervisor's
actions
were within the
allowed scope
and provided adequate
protection,
but that the
work package
was neither changed
nor annotated
to show what
actually happened.
The licensee
subsequently
corrected the
work package
and
added this documentation
problem to their
trending data
base
used to find organizational
and
programmatic deficiencies.
The inspector
concluded that
this resolution
was appropriate.
The inspector
reviewed the use of consumable
materials
during this job.
The review included observation of the
material
as it was being used,
and resolution of actual
practice with statements
in the applicable
Chemical
and
Consumables
Fact Sheets.
Haterials
reviewed included:
~
Gas leak detection fluid [part no. 731-778-00],
~
Spray cleaner
and degreaser
[part no. 737-832-76],
and
~
Nickel-based antisieze
compound [part no. 737-263-66].
Though the inspector
found no problems with the use of the
gas leak detection fluid or the spray cleaner,
the antisieze
compound being used to lubricate the valve packing was
contrary to conditions specified in the Chemical
and
Consumables
Fact Sheet.
Chemical
and Consumables
Fact Sheet
AP-501-00565,
Rev 4, dated
Harch 6,
1994, for antisieze part
no. 737-263-66,
specified that:
This compound
was to be used
as
a lubricant for
metal-to-metal
contact in primary or secondary
system
components,
and
~
This compound
was not permitted to contact reactor
coolant.
Part of the basis for this internal specification
was
an
unsupported
statement,
mentioning
a simila} material,
in an
external proprietary document.
, In contrast,
applicable
maintenance
procedure
HHH-017,
Rev I/2, Valve Packing
Reference
Hanual, specified throughout the procedure to coat
the valve stem,
the inside of the stuffing box,
and the
packing rings with the subject antisieze
compound.
The inspector
found that:
~
Lubrication of packing
was not
a "metal-to-metal
contact"
use,
and
~
Any antisieze
compound entering the valve bonnet from
the packing gland would certainly contact reactor
coolant
and
some would be transported
into the main
coolant loops.
~
Haintenance
procedure
HNN-017 did not adequately
implement the site chemical
and consumable
material
control program described
in AP-501,
Rev 8, Plant
Chemical
and Consumable
Controls.
Procedure
AP-501,
Rev 8,
and associated
fact sheet
AP-501-
00565,
Rev 4,
implemented
TS 6.8. l.a and
Rev.
2,
February
1978, Section
10, which required chemical
and
radiochemical
control procedures
prescribing the limitations
on concentrations
of agents that may cause corrosive attack
or fouling of heat transfer surfaces
or that
may become
sources of radiation hazards
due to activation.
Failure to adequately
implement Procedure
AP-501 is
Violation 400/95-10-02.
In general,
the performance of work was satisfactory with proper
documentation of removed
components
and independent verification
of the reinstallation.
The inspectors identified one violation
involving maintenance
implementation of the site-wide chemistry
control program.
Surveillance Observation
(61726)
The inspector
observed
several
surveillance tests to verify that
approved
procedures
were being used; qualified personnel
were
conducting the tests; tests
were adequate
to verify equipment
operability; calibrated
equipment
was used;
and
TS requirements
were followed.
Test observation
and data review included:
(1)
OST-1119,
Rev 6/2, Containment
Spray
Oper ability - Train B,
quarterly Interval.
The inspector
observed plant personnel
perform this
procedure,
which included
ASME Code Section
XI ISI testing
requirements for the "B" containment
spray
pump
and various
system valves
as required
by TS 4.0.5.
Operators
measur ed
or calculated
various
pump performance
indicators including
differential pressure,
recirculation flow and
pump
10
vibration.
All parameters
were within specified
acceptance
criteria and were appropriately
documented
on test data
sheets.
The inspector verified that operators
were using
an
approved
procedure
and calibrated test equipment.
Licensee
personnel
performed this test satisfactorily.
Acceptance test
EPT-815T,
approved
May, 8,
1995,
Temporary
Procedure
for Acceptance
Test for ESR-00196.
The acceptance
test focused
on response
to pressurizer
level
at both normal
and low charging rates.
This test
collected data at I-minute intervals using the
ERFIS
computer monitoring system.
It also collected stroke time
data.
The inspector
observed
portions of the test
performance.
Test control
and data collection were
effective from the control
room.
The valve system generally
performed well, however the test for normal
(high) flow from
pressurizer
level
above the setpoint
was terminated
because
of high letdown heat exchanger outlet temperature.
The
valve was stable but control
was
somewhat
slow.
Subsequently,
the licensee
adjusted
the valve control
response
speed.
Rev 3, 480/240
VAC Molded Case Circuit Breaker
Test,
and MST-E0007,
Rev 2/3,
120/208
VAC Molded Case
Circuit Breaker Test.
TS 4.8.4. 1 required that each containment penetration
conductor overcurrent protective device shall
be
demonstrated
operable at least
once per
18 months
by
selecting
and functionally testing
a representative
sample
of at least ten percent of each type of lower voltage (less
than
6900 volts) circuit breaker.
This functional test
consisted of removing the breakers
to the shop,
applying
substantial
electrical currents to the breaker using
a test
machine,
and verifying various thermal trip functions per
the subject procedures.
Following recent industry issues
concerning
molded-case
circuit breaker testing,
licensee
personnel
questioned
whether or not practices
contained
in their surveillance
procedures
were acceptable.
Specifically, the procedures
directed plant personnel
to manually cycle the breakers
open
and closed several
times prior to performing the thermal
trip tests.
This practice,
enhancing
personnel
(and
equipment)
safety, verified that the breakers
operated
smoothly and with a fast snap action.
The practice could
have
been considered
as "preconditioning" the breakers prior
to testing
them, thus eliminating potential failure
11
mechanisms
which might have
been present.
The licensee's
initial analysis
was that its testing
program was acceptable
and
was supported
by various industry standards
and previous
NRC correspondence.
The inspector consulted
the previous
NRC correspondence
on
this subject,
Nonconforming Molded-Case
Circuit Breakers.
This bulletin proposed
a test program for
this type of breaker
and cited such industry standards
as
NEHA AB-1, Molded-Case Circuit Breakers;
NEHA AB-2,
Procedures
for Field Inspection
and Performance Verification
of Molded-Case Circuit Breakers
Used in Commercial
and
Industrial Applications; UL-489, Molded Case Circuit
Breakers
and Circuit Breaker Enclosures;
and
NETA STD ATS-
1987, National Electrical Testing Association,
Acceptance
Testing Specifications.
The test
sequence
outlined in the
bulletin included manually operating the breakers
open
and
closed
"a minimum of five times to ensure that the latching
surfaces
a'e free of any binding" prior to an electrical
current test.
The inspectors
concluded
(and informed
licensee
management)
that the licensee's
testing practice
related to cycling these
breakers,
as
implemented in
HST-E0006
and HST-E0007,
was acceptable.
The inspector also concluded that, since mechanical
cycling
of the breaker
was part of the licensee's
test program,
any
related failures represented
a failed functional test.
TS 4.8.4. la.2 and the licensee's
surveillance
procedures
required that another ten percent
breaker
sample
be tested
if any of the first breakers failed functional testing.
Hence,
as
a followup to the licensee's
question,
the
inspectors
reviewed the latest procedure revisions to verify
that mechanical
breaker failures were being treated
as valid
test failures.
Based
on the procedure
review, the inspector
concluded that
a newer procedure
adequately
considered
mechanical
breaker failures,
but the licensee's
procedures
historically did not address
mechanical
problems
as valid
test failures.
Specifically,
a newer revision to procedure
HST-E0006
(Rev 3 effective February
24,
1995) contained
a
specific data sheet entry for mechanically cycling the
breaker,
while a current but older revision to procedure
HST-E0007
(Rev
2 approved April 13,
1992) did not.
Licensee
personnel
indicated that they were considering
enhancements
to procedure
HST-E0007 before the next scheduled
surveillance test.
The inspectors
reviewed results of the last performance of
procedure
and found no indications of problems
with either manually or thermally cycling the breakers.
12
The inspectors
found satisfactory surveillance
procedure
performance with proper
use of calibrated test equipment,
necessary
communications
established,
notification/authorization
of control
room personnel,
and knowledgeable
personnel
having
performed the tasks.
The inspectors
observed
no violations or
deviations in this area.
ENGINEERING ACTIVITIES
a ~
Design/Installation/Testing
of Hodifications (37551)
ESRs involving the installation of new or modified systems
were
reviewed to verify that the changes
were reviewed
and approved
in
accordance
with 10
CFR 50.59, that the changes
were performed in
accordance
with technically adequate
and approved
procedures,
that
subsequent
testing
and test results
met approved
acceptance
criteria or deviations
were resolved in an acceptable
manner,
and
that appropriate
drawings
and facility procedures
were revised
as
necessary.
ESRs documenting
engineering
evaluations
were also
reviewed.
The following engineering
evaluations
were inspected.
(1)
PCR-6796 evaluated
the flow control valve
1CS-231 flow
range.
This valve's
purpose
was to vary charging flow to
maintain
programmed pressurizer level.
The original valve
internals
were designed
oversize resulting in poor response
to pressurizer
level transients
and inability to maintain
a
steady charging flowrate below about
85 gpm.
Operators
had
routinely operated with the valve in manual control.
The
PCR-6796 evaluation
found that the original operating point
was
16 percent of total travel
from "closed"
(a non-linear
range)
and that
a small
change
in plug position resulted
in
a large flow change.
The new design featured linear control
at about
43 percent of the total travel from "closed".
The inspector
reviewed
PCR-6796 while inspecting the on-site
installation of the design
change.
The
PCR evaluated
seismic conditions
and also valve body thrust loads
caused
by using
a stronger spring.
The licensee's
10
CFR 50.59
review,
and the inspector,
concluded that there were
no
unreviewed safety questions.
The design
package
included
an
equipment data
base
change
request for the valve trim parts,
and
a change to applicable
vendor manual
"BJT."
The testing specified in the design
package
focused
on
demonstrating
the valve's response
in two ways:
Pressurizer
level transient
response
demonstrated
by
manually varying pressurizer
level
and then letting
the automatic control return it to normal,
and
System ability to maintain level at lower than normal
flow rates of about
40 gpm.
13
The inspectors
concluded that this
PCR was satisfactory
and
identified no violations or deviations in the
design/installation/testing
of modifications area.
(2)
ESR 9500572 evaluated
the use of anti-sieze
agents
in
contact with reactor coolant,
such
as
when used
as valve
packing lubricant.
One valve packing vendor
had specified
a
specific
compound to use during valve packing installation
while another type vendor
had instructed the plant not to
use similar materials
in contact with reactor coolant.
Plant procedure
AP-501, Plant Chemical
and
Consumable
Controls,
had implemented the requirement
not to use the
material while maintenance
procedure
HHM-017,
Rev 1, Valve
Packing Reference
Manual,
implemented the direction to use
the material.
Following NRC identification,
CR 95-01452
requested
engineering resolution.
This resolution focused
on licensee
standards
and directed certain
shop practices
such
as applying the material
only to certain surfaces-
which would actually negate its effectiveness
as
a valve
packing lubricant.
The resolution also did not resolve the
one vendor's instruction not to use the material
in contact
with reactor coolant.
Following discussions
with the
inspector
and maintenance
department,
the engineering
division reconsidered
the
ESR response,
consulted with
vendors,
and issued
replacement
ESR 9500582.
This second
ESR had
a clearer
evaluation - including reference to vendor
discussions,
and better documented
the reasons
why use of
these materials,
in the amounts
normally used for packing
valves,
was satisfactory without specific restrictions.
The inspector concluded that the engineering division produced
adequate
products,
with one exception found.
Onsite
System Engineering
(37551)
The inspectors
reviewed engineering activities associated
with the
root cause determination for the violation discussed
in paragraph
3.b.(2) of this report.
As discussed
in that paragraph,
an error
in the design of the "C" SG
PORV control
room trouble annunciator
made it unavailable
when the valve's hydraulic control
power was
lost for about
30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />
on Hay 24-25.
The following paragraphs
discuss this design
problem, which was not readily recognizable
by
plant operators
on Hay 24.
A design
change during plant construction
involved relocating
a
ventilation fan's
power supply from HCC 1A31-SA to
a non-safety
HCC, leaving
a spare cubicle
and associated
control
components
in
the lA31 HCC.
The spare cubicle had extra components
to account
for the original fan being controllable from either the main
control
board in the control
room or the auxiliary control panel
if the main control
room was uninhabitable.
14
Later, the designers
(via DCN 251-595,
approved
August 22,
1985)
powered the
1MS-62 valve hydraulic
pump from this spare location.
The AE, by practice,
did not include unused or extra components
in
CWDs.
Additionally, the construction practice
was to not label
components
such
as fuses
and relays in MCCs.
Therefore,
CWD 2166-
B-401,
Sheet
1257 did not show the second
fuse
and second relay
actually located in the cubicle,
and the second
fuse
and second
relay were physically not labeled.
The control board alarm terminals referenced
on the
CWD were
actually attached
to a relay,
but the
CWD referenced
the second
[wrong] relay.
If one tested
the motor controller by opening the
supply breaker, all the correct alarms would sound.
If the
operative control
power fuse blew, the
pump would die with no
alarm.
On May 24,
1995,
when the control
power fuse blew, operators
were
misled
by the absence
of an alarm in the control
room.
Plant
operators
were used to seeing this alarm for the other two
PORVs for similar problems.
The only indication of a blown fuse
was the extinguished light bulbs at the cubicle, which operators
initially treated
as
a bulb problem.
The licensee initiated an event review team after the violation on
May 24-25.
The team
s activities were still continuing at the end
of the inspection period.
Engineers
on the team conducted
a
rigorous search
on the modification history associated
with MCC
cubicle lA31-SA-14D.
The engineers
also conducted
a verification
walkdown of safety-related
cubicles,
containing similar design
features,
which may have
been susceptible
to the
same errors.
No
similar design errors were found.
The inspector inquired about field verification walkdowns to
support the
1985 design
change.
Licensee
personnel
have found no
record of any particular field verification walkdowns related to
the modification.
The inspectors
concluded that
an adequate
verification walkdown during that time would have identified the
drawing/field discrepancy.
The inspectors
also concluded that the
licensee's'efforts
to identify the root cause of the
inoperability were thorough.
However, while the root cause
stemmed
from an old design issue,
current practices to not label
fuses
and relays could lead to future problems related to working
inside
MCC cubicles.
The inspectors identified no violations or
deviations in the systems
engineering
area.
The:inspectors
concluded that the engineering activities were performed
adequately.
The inspectors
identified no violations or deviations in
the engineering activities area.
g
6.
PLANT SUPPORT
15
a ~
b.
Plant Housekeeping
Conditions
(71707)
- The inspectors
reviewed
storage of material
and components,
and observed
cleanliness
conditions of various areas
throughout the facility to determine
whether safety or fire hazards
existed.
The inspectors
found
plant housekeeping
and material condition of components
to be
satisfactory.
Radiological Protection
Program
(71750)
- The inspectors
reviewed
radiation protection control activities to verify that these
activities were in conformance with facility policies
and
procedures,
and in compliance with regulatory requirements.
The
inspectors
also verified that selected
doors which controlled
access
to very high radiation areas
were appropriately locked.
Radiological postings
were likewise spot checked for adequacy.
As mentioned in report paragraph 3.b.(l),
ACFR 95-1359
was
generated
when
an individual entered
the
RCA without proper
dosimetry.
Specifically, the individual logged into the
licensee's
computerized radiation exposure tracking system with
one electronic dosimeter
and erroneously
used another
"non-activated" dosimeter to enter the
RCA, leaving the
"activated" dosimeter at the computer log-in station.
In a
separate
incident,
another individual entered
the
RCA without a
dosimeter after leaving his at the computer log-in station.
Both
individuals reported their own actions.
Self reporting
was
commended
by both the inspectors
and licensee
management.
The main entrance
to the
RCA was provided with three computer log-
in terminals
and
a storage
rack for electronic dosimeters.
Individuals wishing to enter the
RCA would acquire
a personal
electronic dosimeter
and log it into the licensee's
exposure
tracking system using
a computerized
reader.
Once the dosimeter
was logged in and "activated", the individual would use it to
track personal
dose while in the
RCA.
Upon returning from the
RCA, the individual would log out using the computerized
reader
- which would assign
the person's
new dose to a permanent
record.
The individual would then return the newly "de-activated"
dosimeter to one of the storage
racks.
Because
the storage
racks were located
about waist high and just
below the log-in terminal, plant personnel
could just as easily,
due to distraction or lack of attention,
pick up the wrong
dosimeter prior to entering the
RCA.
To correct this
human
factors situation,
licensee
personnel
moved the dosimeter storage
racks to a location
a few feet to the side of the log-in
terminals.
This action should reduce the recurrence
rate of this
problem since,
to have the problem,
personnel
would have to
deliberately walk .over and get
a second
dosimeter after logging
one dosimeter into the computer.
16
The inspectors
were provided with data showing that, in 1995,
seven
ACFRs were related to improper log-in on electronic
dosimetry.
Of these,
four involved entering the
RCA with a
"deactivated"
dosimeter.
The other three involved entering the
RCA with no dosimeter.
For the period between
January I and
Hay 31, plant personnel
made about 44,000 separate
entries into
the
RCA.
The inspector
concluded that seven
incidents represented
a very small trend in this area.
The inspectors
concluded that, while preventing individuals from
"forgetting to take
a dosimeter with them" would be harder to
solve,
the licensee's
attempts to reduce the occurrences
of
individuals entering the
RCA with the wrong dosimeter
were
commendable.
Security Control
(71750)
- During this period, the inspectors
toured the protected
area
and noted that the perimeter fence
was
intact and not compromised
by erosion or disrepair.
The fence
fabric was secured
and the barbed wire was angled.
Isolation
zones
were maintained
on both sides of the barrier
and were free
of objects
which could shield or conceal
an individual.
The
inspectors
observed
various security force shifts perform daily
activities, including searching
personnel
and packages
entering
the protected
area
by special
purpose detectors
or by a physical
patdown for firearms,
explosives
and contraband.
Other activities
included vehicles
being searched,
escorted
and secured;
escorting
of visitors; patrols;
and compensatory
posts.
In conclusion,
the
inspectors
found that selected
security program functions
and
equipment
met requirements.
Fire Protection
(71750)
- The inspectors
observed fire protection
activities, staffing and equipment to verify that fire brigade
staffing was appropriate
and that fire alarms,
extinguishing
equipment,
actuating controls, fire fighting equipment,
emergency
equipment,
and fire barriers
were operable.
During plant tours,
areas
were inspected
to ensure fire hazards
did not exist.
The
licensee's
adherence
to fire protection requirements
was
satisfactory.
Emergency
Preparedness
(71750)
- The inspectors
toured
emergency
response facilities to verify availability for emergency
operation.
Duty rosters
were reviewed to verify appropriate
staffing levels were maintained.
As applicable,
the inspectors
observed
emergency
preparedness
exercises
and drills to verify
response
personnel
were adequately trained.
The licensee
conducted
an augmentation
response drill on June
9,
beginning at 4:50 a.m.
The shift supervisor declared
a "Site Area
Emergency" at 4:58 a.m.,
and initiated
a staff callout.
The
emergency
response facilities were adequately
staffed
and declared
activated at 5:50 a.m.
This response
easily met the emergency
plan requirement of 75 minutes to activate the facilities.
17
The inspectors
found plant housekeeping
and material condition of
components
to be satisfactory.
The licensee's
adherence
to radiological
controls, security controls, fire protection requirements,
emergency
preparedness
requirements
and
TS requirements
in these
areas
was
satisfactory.
The inspectors identified no violations or deviations in
the plant support area.
EXIT INTERVIEW
F
The inspectors
met with licensee
representatives
(denoted in
paragraph
1) at the conclusion of the inspection
on June 9,
1995.
During this meeting,
the inspectors
summarized
the scope
and findings of
the inspection
as they are detailed in this report, with particular
emphasis
on the Violations addressed
below.
The licensee
representatives
acknowledged
the inspector's
comments
and did not
identify as proprietary
any of the materials
provided to or reviewed
by
the inspectors
during this inspection.
The inspector recognized that
one document
reviewed
was marked proprietary,
but it was not used
as
a
basis for concluding acceptability of items inspected.
No dissenting
comments
from the licensee
were received.
Item Number
,
Status
. Description
and Reference
400/95-10-01
400/95-10-02
Open
Open
PORV for
greater
than
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />,
paragraph
3.b.(2).
Failure to adequately
implement
Procedure
AP-501, paragraph 4.a.(2).
ACRONYHS AND INITIALISHS
ACFR
AE
ASHE
CFR
CR
CWD
DCN
encls-
EPT
ERFIS-
GPM
IR
LCO
HCC
Adverse Condition and Feedback
Report
Architect/Engineer
[Design
and Construction
Firm]
American Society of Mechanical
Engineers
- Acceptance Testing Specifications
Code of Federal
Regulations
Condition Report
Control Wiring Diagram
Design
Change Notice
Enclosures
Engineering
Performance
Test
Emergency
Response
Facility Information System
Engineering Service
Request
flow control valve
Final Safety Analysis Report
Gallons
Per Minute
Harris Nuclear Plant
[NRC] Inspection
Report
Inservice Inspection
Limiting Condition for Operation
Motor Control Center
18
MMM
NETA
NPF
NRC
OST
RII
STD
TS
UL
VAC
WR/JO-
Maintenance
Management
Manual
National Electrical Manufacturer's
Association
National Electrical Testing Association
Nuclear Production Facility [a type of license]
Nuclear Regulatory
Commission
Operations Surveillance Test
Plant
Change
Request
Power Operated Relief Valve
NRC Region II
Senior Reactor Operator
Standard
Technical Specification [Part of the Facility License]
Underwriters Laboratories,
Incorporated
'Volts - Alternating Current
Violation [of NRC Requirements]
Work Request/Job
Order