ML18011A489
| ML18011A489 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 06/03/1994 |
| From: | Christensen H, Tedrow J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18011A485 | List: |
| References | |
| 50-400-94-10, NUDOCS 9406270253 | |
| Download: ML18011A489 (31) | |
See also: IR 05000400/1994010
Text
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UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W., SUITE 2900
ATLANTA,GEORGIA 30323-0199
Report No.:
50-400/94-10
Licensee:
Carolina
Power
and Light Company
P.
O.
Box 1551
Raleigh,
NC 27602
Docket No.:
50-400
Facility Name:
Harris
1
Inspection
Conducted:
April 3 - Hay 6,
1994
Lead Inspector:
J.
e row,
Se
r Re
ent Inspector
Other Inspectors:
D. Roberts,
Resident
Inspector
R. Hoore,
Reactor Inspector
Approved by:
H.
h istensen,
Acting
Reactor Projects
Branch
Division of Reactor Projects
Licensee
No.:
ate Signed
e
ige
SUHHARY
Scope:
This routine inspection
was conducted
by two resident
inspectors
in the areas
of plant operations,
refueling activities, safety
system walkdowns, review of
nonconformance
reports,
evaluation of licensee self assessment,
maintenance
observation,
surveillance observation,
design
changes
and modifications,
system engineering,
plant housekeeping,
radiological controls, security, fire
protection,
review of licensee
event reports,
and licensee
action
on previous
inspection
items.
Numerous facility tours were conducted
and facility
operations
observed.
Some of these tours
and observations
were conducted
on
backshifts.
Results:
One violation was identified:
Failure to establish
and
implement procedures
for system fill and vent
and for procedure
changes,
paragraphs
2.a(2)(a),
and
4.a(3).
Additionally, two non-cited violations were identified:
Failure to implement
fire protection procedures
adequately,
paragraph
5.d.;
and failure to
adequately
post dress
requirements for a high contamination
area,
paragraph
S.b.
9406270253
940603
ADm:K 05000400
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Inspection of opposite trains for similar deficiencies
was strong,
paragraph
2.d.
Improvement
was noted in line organization self assessment,
paragraph
2.e.
Plant management
demonstrated
a commitment to resolving deficiencies
prior to accident
occurrence
by stopping all outage work, paragraph
3.a.
Pre-
evolution briefs for complex tests
were effective in avoiding plant upsets,
paragraph
3.b.
Excellent radiation protection efforts demonstrated
during
containment modification work, paragraph
5.b.
Weaknesses
were identified in configuration control of circuits for the fuel
manipulator,
paragraph
2.b; initial documentation
of an engineering
evaluation
for a loose object in the
"C" steam generator,
paragraph
4.a(4);
performance
and verification of new nuclear instrument intermediate
range currents,
paragraph
4.b; lack of attention to detail in radiological posting,
paragraph
5.b;
and fire protection activities,
paragraph
5.d.
~ ~
\\
REPORT
DETAILS
1.
PERSONS
CONTACTED
Licensee
Employees
D. Batton,
Hanager,
Work Control
D. Braund,
Hanager,
Security
- B. Christiansen,
Hanager,
Haintenance
J. Collins, Nanager,
Training
- J. Donahue,
General
Hanager,
Harris Plant
J.
Dobbs,
Nanager,
Outages
- H. Hamby,
Hanager,
Regulatory
Compliance
- H. Hill, Nanager,
Site Assessment
- D. HcCarthy,
Hanager,
Regulatory Affairs
- J. Nevill, Hanager,
Technical
Support
- R. Prunty,
Hanager,
Licensing
& Regulatory
Programs
- W. Robinson,
Vice President,
Harris Plant
W. Seyler,
Hanager,
Project
Hanagement
- H. Smith, Hanager,
Radwaste
Operation
D. Tibbitts, Hanager,
Operations
- B. White, Hanager,
Environmental
and Radiation Control
A. Williams, Hanager, Shift Operations
Other licensee
employees
contacted
included office, operations,
engineering,
maintenance,
chemistry/radiation
and corporate
personnel.
- Attended exit interview
2.
Acronyms and initialisms used throughout this report are listed in the
last paragraph.
Operations
a ~
Operational
Safety Verification (71707)
The plant began this inspection period in refueling operations
(Node 6).
Fuel
was off-loaded from the reactor vessel
at
8:27 a.m.
on April 4.
Refueling operations
were
recommenced
at
9:37 a.m.
on April l9 with the core being completely reloaded
on
April 21 at 11:30 p.m.
At 6:30 a.m.
on April 27 the reactor
vessel
head
studs
were tensioned
and the plant was placed in the
cold shutdown
(Node 5) condition.
On Hay 4 a plant heatup
was
commenced
and the hot standby
(Node 3) condition reached
at 5:07
p.m.
on Nay 5.
The plant remained in hot standby for the
remainder of this inspection period.
(1)
Shift Logs and Facility Records
The inspector
reviewed records
and discussed
various entries
with operations
personnel
to verify compliance with the
Technical Specifications
(TS)
and the licensee's
administrative
procedures.
The following records
were
reviewed:
shift supervisor's
log; outage shift manager'
log; control operator's
log; night order book; equipment
inoperable record; actiVe clearance
log; grounding device
log; temporary modification log; chemistry daily reports;
shift turnover checklist;
and selected
radwaste
logs.
In
addition, the inspector
independently verified clearance
order tagouts.
The inspectors
found the logs to be readable,
well
organized,
and provided sufficient information on plant
status
and events.
Clearance
tagouts
were found to be
properly implemented.
No violations or deviations
were
identified.
Facility Tours
and Observations
Throughout the inspection period, facility tours were
conducted to observe
operations,
surveillance,
and
maintenance activities in progress.
Some of these
observations
were conducted
during backshifts.
Also, during
this inspection period, licensee
meetings
were attended
by
the inspectors
to observe
planning
and management
activities.
The facility tours
and observations
encompassed
the following areas:
security perimeter fence; control
room; emergency diesel
generator building; reactor auxiliary
building; Reactor
Containment Building (RCB); waste
processing
building; turbine building; Fuel Handling
Building (FHB); battery rooms; electrical
switchgear
rooms;
and the technical
support center.
During these tours,
observations
were
made
on monitoring
instrumentation
which included equipment operating status,
area
atmospheric
and liquid radiation monitors, electrical
system lineup, reactor
operating
parameters,
and auxiliary
equipment operating parameters.
Indicated parameters
were
verified to be in accordance
with the
TS for the current
operational
mode.
The inspectors
also verified that
operating shift staffing was in accordance
with TS
requirements
and that control
room operations
were being
conducted
in an orderly and professional
manner.
In
addition, the inspector
observed shift turnovers
on various
occasions
to verify the continuity of plant status,
operational
problems,
and other pertinent plant information
during these turnovers.
(a)
On April 16,
1994,
the inspector
observed
control
room
operator activities associated
with filling and
venting the
RHR,
CVCS,
and SI systems.
These
systems
were being filled to establish
a flowpath of water to
the
RCS.
The
RCS was being filled via this process
to
a level which would allow continued filling using
General
Procedure
GP-009,
Refueling Cavity Fill,
Refueling
and Drain of the Refueling Cavity.
While observing the system filling activities, the
inspector noticed
a
memo in the control
room written
by an off-shift supervisor containing steps,
notes,
and caution statements
similar to those normally
contained
in operating
procedures.
The memo, dated
April 16,
1994,
contained specific sequential
instructions to operators
on which valves to
manipulate in order to fill the
ECCS systems.
The
inspector
asked
the on-shift
SCO if the
memo was being
referenced for the filling and venting then in
progress.
The
SCO responded
that the
memo was indeed
being used.
The inspector then inquired
as to whether
any other operating
procedures
were being
used
and if
not, should the steps
contained
in the
memo appear
in
such
a procedure.
The operator
responded
that the
memo was being
used to concurrently fill the
RHR,
and SI systems
because
plant operating
procedures
did
not cover the concurrent filling of these
systems.
While the majority of system filling was accomplished
using steps outlined in the
memo, the document did
reference
using procedure
OST-1107,
ECCS Flow Path
and
Piping Filled Verification Honthly Interval,
as
a
guide to vent the
Only venting
for the
CVCS system
was to be accomplished
by
manipulating valves
as specified in the
memo.
The
inspectors
did not note
any technical
problems with
the sequence
of actions
used
by the operators
to fill
and vent the systems
as prescribed
by the
memo.
However, Regulatory
Guide 1.33
recommends
procedures
be established
and implemented for filling and venting
the systems
noted
above.
The inspectors
considered
the licensee's
actions to fill and vent the
RHR,
CVCS,
and SI systems
without a formally approved
procedure
to be
a violation of the requirements
contained in TS 6.8.1.a.
Violation (400/94-10-01):
Failure to establish
and
implement procedures.
Refueling Activities (60710)
The inspectors
witnessed
several shifts of fuel handling
operations
and verified that the refueling was being performed in
accordance
with TS requirements
and approved
procedures.
Areas
inspected
included containment integrity, housekeeping
in the
refueling area, shift staffing during refueling, surveillance
testing,
and periodic monitoring of plant status
during refueling
operations.
In addition, partial
implementation of the following
procedures
was observed:
~ GP-009
Refueling Cavity Fill, Refueling,
and Draindown of the
Refueling Cavity
~ FHP-010
Core Mapping Following Fuel
Loading
~ FHP-014
Fuel
and Insert Shuffle Sequence
~ FHP-020
Fuel Handling Operations
~ HPP-625
Performance of Radiological
Surveys
The inspectors
viewed the video tape
made of the core verification
performed in accordance
with procedure
FHP-010.
In addition, the
inspectors verified that locked high radiation areas
were properly
posted during the spent fuel transfer
between the
FHB and reactor
cavity.
The licensee's
implementation of the refueling activity
was satisfactory with proper oversight of the contractor
performing the activity.
The inspectors
reviewed the manipulator crane
and auxiliary hoist
checkout during fuel movement
on April 21,
1994.
This activity
was controlled by procedure
FHP-020 attachment
3 and included
startup
checks to verify proper equipment operation prior to
latching onto fuel elements.
The inspector
noted that step
3. 1.4'.g
had not been initialed to signify that circuits seven
through ten were off in the
MCC distribution panel
on the
manipulator crane.
An operator
was requested
to check the status
of the circuits which were found to be
on instead of off.
The
circuits were labeled
as spares.
The inspector
requested
licensee
personnel
to explain the function of these circuits.
The
applicable
CWD's depicted
these circuits to have
no function so
the inspector
concluded that the circuits were spares
and served
no safety related function.
However, the inspector considered
the
licensee's
configuration control of the manipulator
MCC circuits
was weak.
No violations or deviations
were identified.
Safety
Systems
Walkdown (71707)(71711)
The inspectors
conducted
a walkdown of the passive
safety
injection system to verify that the lineup was in accordance
with
license requirements for system operability and that the system
drawing
and procedure correctly reflected "as-built" plant
conditions.
In addition, portions of the high head safety
injection, low head safety injection,
emergency
containment
spr ay,
systems
located inside
containment
were walked down.
The inspectors
also walked
down
containment
located inside the steam tunnel
area.
~ ~
d.
e.
The inspector
found the material condition of components
to be
generally good.
Hinor boric acid leaks were evident
on
a few
passive
safety injection vales
and flanges.
These
items were
corrected
by the licensee.
No violations or deviations
were identified.
Review of Nonconformance
Reports
(71707)
Adverse Condition Feedback
Reports
(ACFR) were reviewed to verify
the following;
TS were complied with, corrective actions
and
generic
items were identified and items were reported
as required
by 10 CFR 50.73.
ACFR 94-1194 identified that the stem for primary
disconnected
from the valve disk.
The licensee's
preliminary
investigation revealed that the roll pin which fastens
these
two
parts together
had sheared.
An inspection of the other two
revealed similar damage to valve
and
no damage to valve
was performed
by licensee
personnel
who determined that the valves remained
in this
condition because
system pressure
underneath
the valve disc would
force open the valve when the actuator
opened
the valve stem.
The
valves were subsequently
repaired.
Licensee
personnel
had noticed
that valve
1RC-114 exhibited seat
leakage during the previous
operating cycle.
In addition to the
PORVs, licensee
personnel
investigated
opposite
train components
when problems
were identified on the
CCW heat
exchangers
and the shaft failure of the "8" CSIP.
When the "A"
CSIP was disassembled
for inspection,
no evidence of similar
cracking
was found.
However, the disassembly
resulted
in damage
to threads
on the shaft for the balance
drum lock nut which
necessitated
the replacement of the rotating element.
The
practice
by licensee
personnel
to examine other train components
when problems
were identified is considered
to be
a strength.
No violations or deviations
were identified.
Evaluation of Licensee Self Assessment
(40500)
The inspectors
attended
selected
PNSC meetings to observe
committee activities
and verify TS requirements for committee
composition, duties
and responsibilities.
Heeting minutes
were
also reviewed to verify that activities were accurately
documented.
During a meeting
on Hay 6, preparations
for starting
the unit were discussed
as
was
a service advisory letter from the
nuclear
steam supplier which alerted the licensee to
nonconservatisms
in the overpower reactor trip setpoint
calculations for operation with inoperable
main steam safety
valves.
Technical specification 3.7. 1. 1 requires that the main steam
safety valves
be operable.
Action statement
3.7. 1. I.a specifies
that with a safety valve inoperable,
power operation
may proceed
provided that the overpower reactor trip setpoints
be reduced
per
Table 3.7-1.
The licensee's
nuclear
steam supplier vendor advised
that the calculations for the setpoints
specified
by this table
were nonconservative.
The
PNSC decided to implement
an internal
TS interpretation to require
a plant shutdown if a main steam
safety valve was inoperable.
This interpretation will remain in
effect until new setpoints
can
be developed.
The inspector
considered
the licensee's
action to be conservative.
The licensee recently implemented
procedure
PLP-614, Self-
Assessment
for Restart
Readiness
to Startup.
Highlights of this
new program included field walkdowns of 41 systems
on the primary
and secondary
side of the plant as well as electrical distribution
systems
by both senior reactor operators
and system engineers.
Walkdown deficiencies
were evaluated
to determine
the need for
correction before
commencement
of startup.
Along with the systems
readiness
element of this program,
organizational
readiness
was
also assessed.
Managers
were required to ensure staffing levels
were adequate,
personnel
training was conducted
as appropriate,
ACFRs/FBRs/CAP items were completed,
backlog items reviewed,
and
regulatory items were completed.
The program also contained
an
operational
readiness
assessment
from each shift supervisor
and
a
verification of the core configuration by engineering
support
organizations.
Predetermined
assessment
hold points were placed
in the outage
schedule prior to changing to Mode
4 and to Mode
2
to ensure
management
expectations
were met.
The
PNSC meeting
attended
by the inspector discussed
the readiness
for changing to
Mode 2.
The inspector
noted very detailed discussions
between
the
plant general
manager
and
PNSC members.
The inspector considered
this activity to be valuable to ensure items/deficiencies
received
appropriate
management
review prior to plant startup.
This
program was considered
to be
a substantial
improvement in licensee
self assessment
efforts.
Licensee Action on Previously Identified Operations
Inspection
Findings
(92901)
(I)
(Closed) Violation 400/93-12-02:
Failure to properly
implement plant procedures.
This item was previously discussed
in
NRC Inspection
Report
50-400/93-24.
The inspectors
reviewed
and verified
implementation of the corrective actions listed in the
licensee's
response letter dated July 30,
1993.
Training
was provided to operations
personnel
and lesson
plans for
continuing training revised.
Placards
were installed inside
the breaker cubicles to indicate the location for verifying
HOC alignment.
(2)
Maintenance
(Closed) Violation 400/93-25-03:
Failure to properly
establish
procedure
The inspectors
reviewed
and verified completion of the
corrective actions listed in the licensee's
response letter
dated July 30,
1993.
The licensee
has discussed
the
violation with shift supervisors
and counseled
involved
personnel.
The procedure
steps for reducing turbine load
have
been
removed
from procedure
revised
and
included into procedure
GP-006,
Normal Plant Shutdown
From
Power Operation
To Hot Standby.
'a ~
Maintenance
Observation
(62703)
The inspector observed/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.
Maintenance
was observed
and work packages
were reviewed for the following maintenance activities:
~
Rebuild main steam isolation valve
1HS-84 in accordance
with
procedures
CH-M0061, Hain Steam Isolation Valve Disassembly
and,Maintenance,
and
Operator.
Breaker replacement for the "A" service water booster
pump
and supply breaker for HCC-1A32-SA in accordance
with
modification PCR-6526.
Weld repair of "A" CCW heat exchanger
web cracks in
accordance
with procedure
HAP-04, Plant Management
Work
Control Procedure,
PLP-710,
Work Management
Process,
HHP-
006, Installation of Structural
Steel
and Electrical,
Instrumentation,
and
HVAC Supports,
and modification
PCR-
7238.
Replace
PSU-Ill connectors
in
DC HCCs DP-1B2-SB
and DP-1A2-
SA in accordance
with modification PCR-7167.
Place placards
inside the breaker cubicles for the "A", "B",
and
"C" RCPs.
Clean
and inspect the intake manifold to the right bank
intercooler
on
EDG-1A,
and replace
the intercooler in
accordance
with procedure
Emergency Diesel
Generator
Intercooler Inspection
and Cleaning.
Perform EDG-lA fuel injector nozzle inspection
and cleaning
in accordance
with procedure
HST-H0016,
Emergency Diesel
Generator
Fuel Injection Nozzle Inspection
& Cleaning.
~
~
~ ~
I
b.
~
Perform
EDG-1A and
EDG-1B Cold Compression
and
maximum
firing pressure
checks
in accordance
with procedure
HST-
H0014,
Emergency Diesel
Generator
Cold Compression
& Haximum
Firing Pressure
Checks.
~
Inspect motors for "A" and "B" CSIPs for oil buildup.
~
Perform
EDG-1B cylinder liner inspection in accordance
with
procedure
HST-H0010,
Emergency Diesel
Generator Cylinder
Liner Visual Inspection.
In general,
the performance of work was satisfactory with proper
documentation of removed
components
and independent verification
of the reinstallation.
During the inspection of the "A" CCW heat
exchanger for fouling, the interior channel
lower web was found to
be bent
and cracked.
In addition, the baffle partition plate
was
also found to be bent.
Hodification PCR-7238
was initiated to
repair this condition.
An inspection of the "B" CCW heat
exchanger
was subsequently
performed
and revealed similar damage
(See
paragraph 4.a.(1)).
The inspector attended
safety meetings
held for licensee
personnel
on April 12.
Outage
work was stopped for approximately
one hour
to conduct these
meetings.
The meetings
were held because
licensee
management
and
NAD noted several
instances
of personnel
safety equipment not being utilized properly and examples of near
misses
to potential
accident situations.
Examples
included
situations
where licensee
personnel
identified poor electrical
safety practices of using under rated extension
cords
and use of
multi-plug power strips without adequate
overload protection.
Also, personnel
were noted to be working without proper eye
protection
and without tie-offs while working in elevated
positions.
Housekeeping
problems
had also
been identified.
The
near misses
included
a fire extinguisher which was dropped
from a
high elevation in the
CST as well as several
tools which had
been
dropped
from elevations
inside the
RCB.
An electrocution
was
narrowly avoided
when electricians incorrectly removed
access
covers to the
22
KV isophase
buss duct instead of accessing
the
intended
deenergized
neutral
grounding enclosure.
The isophase
buss duct was energized for backfeeding electrical
power from the
switchyard to the plant electrical distribution system.
The
energized
buss
was discovered
before
any personnel
injury/equipment
damage
occurred.
The inspector considered
the
conduct of these
meetings to be appropriate
considering all the
events
discussed
above
and demonstrated
licensee
management's
commitment to resolving deficiencies prior to the occurrence of
events/accidents.
Surveillance
Observation
(61726)
Surveillance tests
were observed to verify that approved
procedures
were being used; qualified personnel
were conducting
=
the tests;
tests
were adequate
to verify equipment operability;
calibrated
equipment
was utilized; and
TS requirements
were
followed.
The following tests
were observed
and/or data reviewed:
~ OST-1046
Hain Steam Isolation Valve Operability Test quarterly
Interval
~ OST-1088
Low Head Safety Injection Valves ISI Test quarterly
Interval
~ OST-1801
ECCS Throttle Valve,
CSIP,
and
Verification 18 Honth Interval.
~ OST-1809
Switchover to Recirculation
ESF Response
Time
18 Honth Interval.
~ OST-1824
1B-SB Emergency Diesel
Generator Operability Test
18
Honth Interval.
~ OST-1835
MSIV Remote
Shutdown With MSIV And Bypass Isolation
Remote Position Indication Test
18 Honth Interval
~ OST-1841
ESF Response
Time Heasurement for 1C-SAB CSIP
18 Honth
Interval
~ HST-I0044 Calibration of Nuclear Instrumentation
System
Power
Range
N41
~ HST-I0045 Calibration of Nuclear Instrumentation
System
Power
Range
N42
~ HST-I0046 Calibration of Nuclear Instrumentation
System
Power
Range
N43
~ HST-I0047 Calibration of Nuclear Instrumentation
System
Power
Range
N44
~ HST- I0167 Excore Nuclear Instrumentation
System Intermediate
Range
N35 Operational
Test
~ HST-I0168 Excore Nuclear Instrumentation
System Intermediate
Range
N36 Operational
Test
The performance of these
procedures
was found to be satisfactory
with proper
use of calibrated test equipment,
necessary
communications
established,
notification/authorization of control
room personnel,
and knowledgeable
personnel
having performed the
tasks.
As discussed
in
NRC Inspection
Report 50-400/94-06,
licensee
personnel
continued the pre-evolution briefs for the
complex surveillance tests.
This practice
was effective in
avoiding potential plant upsets.
No violations or deviations
were
observed.
~ ~
l
10
C.
Licensee Action on Previously Identified Maintenance
Inspection
Findings
(92902)
(1)
(Closed) Violation 400/93-25-02:
Failure to promptly
correct
a deficiency in
MCC wiring and prevent recurrence.
The inspector
reviewed
and verified completion of the
corrective actions listed in the licensee's
response letter
dated
March 7,
1994.
The licensee
has completed
work
tickets to restrain electrical
cables with tie-wraps to
prevent
movement
and
has replaced
the terminal connections
with a non-separable
connection.
(2)
(Closed) Deviation 400/93-24-01:
Failure to request
NRC
extension for corrective actions
committed to.
Engineering
The inspectors
reviewed
and verified implementation of the
corrective actions listed in the licensee's
response letter
dated
February
14,
1994.
Licensee
personnel
have installed
visual aids inside the "A", "B" and
"C" RCP breaker cubicles
to indicate the location for conducting
MOC alignment
verifications.
In addition, the licensee
has revised the
process for controlling action items to improve visibility
a ~
Design
Changes
and Modifications (37828)
Plant
Change
Requests
(PCR) 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.
In addition,
PCR's
documenting engineering
evaluations
were also reviewed.
The
following modifications and/or testing in progress
was observed:
~ PCR-7167
Replacement
Terminal for DP-IA2-SA, DP-1B2-SB
~ PCR-7238
CCW Heat Exchanger Baffle Plate
~ PCR-0420
RTD Bypass Elimination
~ PCR-4888
Loose Parts in Steam Generators
A, B,
and
C.
~ PCR-3525
"A" and
"C" Steam Generator
Loose Parts
~ PCR-6526
Train "A" Load Centers
LK Breaker
Replacement
I
~ PCR-7171
Pressurization
of
B CSIP while under clearance
~ PCR-7274
CSIP Runout Technical Specification
Compliance
~ PCR-6831
1A-SA and
1B-SB CSIP Replacement
~ PCR-6925
Remove
~ PCR-7251
Foreign Object in
Secondary
Modification PCR-7238
was initiated following the
identification of cracks in the interior channel
lower web
of the "A" CCW heat exchanger.
The baffle partition plate
was also found to be bent.
Licensee
personnel
analyzed this
condition and believe that it was caused
by normal
system
operation in conjunction with inadequate
component
design
margin.
The modification reinforced the horizontal
partition web and baffle plate.
This modification was
performed
on both the "A" and
"B" CCW heat exchangers.
The inspector
found the engineering
evaluation
associated
with PCR-6831 to be very thorough
and included the effect of
rotating element
replacement
on the accident analysis,
systems
operation,
design basis
issues,
pump protection
concerns
and any electrical
impacts.
As
a result of the
pump rotating element replacement,
the licensee
plans to
startup using the
"C" CSIP in place of the "A" CSIP to
facilitate smooth operation of the normal charging flow
control valve.
In addition, the
TDAFW pump controller
setpoint
was adjusted to 28 percent to accommodate
less
margin in the steam generator
tube rupture analysis.
The
RTD bypass elimination modification removed
and capped
the
RTD bypass manifold piping and installed
new
thermowells
and fast acting
Company
RTDs.
In addition to the hardware
changes,
circuit changes
were
made to the Protection
Instrument Cabinets
(PIC-1, PIC-2,
PIC-3,
and PIC-8).
The instrumentation
portion of the
modification was reviewed to determine if adequate
instrument overlap
was incorporated
in post modification
testing to assure
appropriate verification of RTD instrument
protection
channel
performance.
Additionally, the
acceptance
criteria was reviewed to determine if the
instrument
response
time assumptions
assumed
in the Final
Safety Analysis Report
(FSAR) accident analysis
would be
verified by the testing.
The
RTD instrument loop was separated
into three sections
for response
time testing.
Section
one was from the
detector to the input of PIC-1, PIC-2,
and PIC-3 (thermal
response
time).
Procedures
HST-10635, Self Heating Test of
Installed Resistance
Thermometers,
and I0636, Insitu
a
12
Response
Time Testing of Installed
RTDs, provided guidance
for this test.
Section
two was from the
PIC input to the
output of the Train A and Train
B Solid State Protection
System
(SSPS) drivers (circuit response
time).
Procedures
HST I0644,
Group
1 of 3 Channel
RTS and
ESFAS Response
Time
Test,
I0645,
Group
2 of 3 Channel
RTS and
ESFAS Response
Time Test,
and I0646,
Group
3 of 3 Channel
RTS and
Response
Time Test,
provided guidance for this test.
Section three
was from the
SSPS drivers to trip breakers
A
and
B (breaker
response
time).
Procedures
HST I0647, Train
A UV Output Driver to Reactor Trip Breaker
Response
Time
Test,
and
10648, Train
B UV Output Driver to Reactor Trip
Breaker
Response
Time Test,
incorporated this test.
Although section three
was not impacted
by this
modification,
a surveillance test
was performed to determine
the breaker
response
time of trip breaker
A.
All procedures
used for response
time testing were existing procedures.
The inspector
concluded
the scope of post modification
testing incorporated
adequate
instrument loop overlap.
Section
one testing
had not been
completed at the end of
this inspection period.
The acceptance
criteria was
specified to be
4 seconds for the
RTD thermal
response.
Documentation for section
two testing
was included in WRs
94-ABYE1, 94-ABYHl, and 94-ABYKI.
Section
two response
time
was 0.3 seconds.
Section three tests
were documented
in
94-JI(001
and 92-JIPOOI.
The breaker
response
time was
0.0675 seconds.
The acceptance
criteria for the overall
trip circuit electronic delay which included the scram
control rod gripper release
time was specified to be 1.25
seconds.
The total. trip circuit electronic release
time was
0.3675
and substantially less
than the specified
1.25
seconds.
Completion of the section
one testing will
determine if the
RTD response
time was within the
6 seconds
assumed
by the
FSAR chapter
15 accident analysis.
The
inspector concluded that scheduled
and completed testing
was
adequate
to verify the
FSAR assumption
regarding
instrument loop response.
The inspectors will follow
section
one testing during future routine inspection
activities.
Other instrumentation
post modification testing included
average
temperature
and differential temperature
loop
calibration
and operational
tests of PIC control functions.
These tests
were accomplished
by existing surveillance
procedures.
Overall, the inspector
concluded that the scope
of post modification testing
was adequate
to demonstrate
instrumentation
performance
and verify the accident analysis
assumption.
The inspector verified the main control board
and simulator
instrumentation
hardware
changes
were implemented
and that
I ~
~
(4)
13
operator training had
been developed
to address
the
modification.
The hardware
changes
included removal of
meters,
and control
channel
defeat
switches
for low bypass
manifold flow, average
temperature,
and
control differential temperature.
These
changes
were
completed for both the control
board
and the simulator.
Lesson
Plan
LOR 94-1,
RTD Bypass
Loop Removal
& Reduced
T-
Hot Hods,
addressed
the impact of PCR-0420 to operators.
The inspector
concluded that the control
board
and simulator
instrumentation
had
been appropriately
updated consistent
with the plant modification.
Training adequately
addressed
the impact of the modification to operators.
The breaker
replacement
modification replaced existing 480
VAC Asea
Brown Boveri
(ABB) LK breakers with Siemens
RLN
type 480
VAC breakers.
The licensee
had experienced
chronic
failures of the
ABB LK breakers
over the preceding years.
The breaker failure issue
was addressed
in
NRC inspection
reports
50-400/93-15,
50-400/92-26,
and 50-400/89-13.
Unable to resolve the cause of the failures, the licensee
elected to install breakers of a different manufacturer with
equivalent operating characteristics
and rating.
This
modification replaced
the breakers
on 480
VAC emergency
buses
1A2-SA and
Additionally, the breaker cradles
were replaced to permit mounting of the Siemens
breakers
in
the existing switchgear cubicles.
The inspector
reviewed the scope of post modification
testing specified
by the
PCR to determine if testing
requirements
provided adequate verification of breaker
performance.
Completed test documentation for breakers
1A3-
SA-5B and lA3-SA-6B was reviewed to determine if the
required testing
was performed
and acceptance
criteria were
satisfied
by the test.
Additionally, the inspector reviewed
the licensee's
safety analysis
which evaluated
the
suitability of the replacement
breakers for the plant
application.
The following WR implemented post modification testing of
breaker
1A3-SA-5B: 93-AKABl performed
HST E0072,
480
VAC
Siemens
Type
RLN Load Center Breaker
and Cubicle Test,
and
CH-E0016, Current Transformer Ratio and Polarity Test.
Verification of indication and control functions
was
performed with WR 93-AKAB3 and procedure
EPT-615T,
Temporary
procedure for Testing Indication
and Control functions of
Breaker Supply Motor Control Center
1&4 A33-SA.
This
also verified the setpoint of the ground fault relay.
Post
modification testing identified an installation deficiency
related to missing jumpers
on breaker
1A3-SA-5B which was
corrected
by
WR 94-AEHK1.
Similar testing
was accomplished
for breaker
1A3-SA-6B on
WRs 93-AKDA1 and 93-AKAD3.
The
post modification testing
on 1A3-SA-6B identified
a breaker
indication deficiency caused
by an incorrect termination
which was corrected
by
WR 94-AENW1.
The inspector
concluded
that the scope of post modification testing required
by PCR-
6526 was adequate
to verify the equipment function.
Test
documentation
demonstrated
appropriate verification of
equipment control
and protection features.
The inspector
reviewed the licensee's
safety analysis
which
determined that the Siemens
RLN breakers
were
an acceptable
replacement for the previously installed
ABB LK breakers.
The analysis
addressed
breaker time-current characteristics
and over current devices to verify that proper coordination
was maintained with upstream
and downstream
breakers.
Applicable electrical calculations,
drawings,
and setpoint
documents
were referenced
in the analysis.
The Siemens
breaker current interrupt ratings of 65,000
amperes
instantaneous,
and 50,000
amperes
short term met or exceeded
the
ABB LK breaker ratings of 50,000
amperes
instantaneous
and short term.
The maximum available short circuit current
calculated
by the licensee's
electrical fault calculation
(E-6000)
was less than 50,000
amperes.
The inspector
concluded that the licensee's
analysis
adequately
evaluated
the suitability of the Siemens'reakers
for this
, application.
The inspectors
observed
some of the post modification
testing mentioned
above for the
new Siemens
breakers.
All
testing
was performed
by electricians
in accordance
with
various test procedures
that were developed
by a cognizant
test engineer.
To affect the cycling of the breakers,
certain sections of the procedures
required that control
power and auxiliary control
power fuses
be removed
and
reinstalled.
In some cases,
the test procedure
steps
specified the sizes of the fuses.
For example, test
procedure
EPT-609T,
Temporary Procedure for Testing
Indication of Control Functions of Breaker Supplying
HCC
IA35-SA, step 7.4.3, specified auxiliary control
power fuse
sizes of 35 and 30 amperes
to be installed in fuse locations
FU5 and
FU6, respectively.
The inspectors identified several
cases
where
pen
and ink
changes
were
made to the procedures
in order to complete the
tests.
The inspector
reviewed procedure
EPT-609T
and noted
that the specified fuse sizes of 35 and
30 amperes
had
been
lined through
and replaced with annotations
of 15 and
5
amperes,
respectively.
The inspector then noticed that
15
and
5 ampere
fuses
had actually been installed in the
breaker cubicle.
The inspector performed
a followup review
of various other test procedures
associated
with the
new
Siemens
breakers
and found three additional
examples
where
incorrect fuse sizes
were lined out and replaced
by pen
and
ink changes,
then initialed and dated
by the test engineer.
I
~ ~
~
15
The inspector discussed
the handwritten
changes
in these
procedures
with the test engineer
who indicated that the 35
and
30 ampere
fuse sizes
were
now incorrect
due to late
PCR-
6526 revisions.
The inspector
asked
why a temporary
procedure
change
was not initiated.
The engineer
stated
that the tests
were being performed in compliance with
applicable plant guidelines
on procedural
adherence
(AP-100,
Procedure
Use
and Adherence).
The engineer's
assumption
was
based
on the fact that the test procedures
contained
notes
directing technicians
to contact the on-shift test engineer
(or the maintenance
supervisor) for resolution of any
unexpected
results prior to continuing with the test.
The
engineer believed the notes
empowered
him to line through
the erroneous
fuse information and write in the proper
sizes.
(6)
During
a closeout
review of the other work packages
associated
with PCR-6526,
the inspector identified one
additional
example
where
a test procedure
had
been
completed
with erroneous
information lined through, corrected,
and
initialed by the test engineer.
Step 7.3.6.b of test
procedure
EPT-605T,
Temporary Procedure for Testing
Indication
and Control Functions of Breaker Supplying
Pump lA-SA, specified
a jumper with resistance
of 2.5k ohms
be installed
between
two terminations.
The 2.5k ohm value
had
been lined through
and annotated
with 1.25k ohms.
The
test
had
been
signed off as complete.
The inspectors
did not identify any technical
problems
associated
with the performance of these tests.
The test
engineer did reference
the correct data from PCR-6526 prior
to making the
pen
and ink changes.
However, Technical Specification 6.8. l.a and the licensee's
administrative
procedure
AP-100 require that procedure
changes
be
implemented
before continuing with a procedure
step which
would result in an incorrect action or inappropriate
response.
The licensee's
actions involving unofficial pen
and ink changes
to the various
480 volt breaker test
procedures
are contrary to the above requirements
and are
considered
to be another
example of the violation discussed
in paragraph
2.a(2)(a) of this report.
During sludge lancing of the secondary
side of the steam
generators
during this refueling outage,
licensee
personnel
found several
loose parts which could not be retrieved.
A
ball detent
was found in steam generator
B and two ball
detents
and
a small rod were found in steam generator
C.
The ball detents
had been previously analyzed to be
acceptable
for continued plant operation
by PCR-3525 for the
life of the plant.
Evaluation
PCR-3525 also analyzed
two
short weld rods which were left in the "A" steam generator
after the first refueling outage
and evaluation
PCR-4888
16
analyzed
a bolt shank which was discovered
in the "B" steam
generator
which could not be retrieved after the second
refueling outage.
Engineering
personnel
evaluated
the
foreign objects in the steam generators
found during this
refueling outage
and considered
them to be acceptable
for
continued operation
based
upon
PCR-3525
and
PCR-4888.
This
conclusion
was documented
in an internal licensee
memorandum.
The inspector reviewed the
and the
memorandum
and noted that the
PCRs contained
statements
that
the evaluations for the weld rods
and bolt shank were valid
for only the next cycle of operation after which the
adjacent
tubes
were to be examined
and further attempts
made
to retrieve.
The memorandum only discussed
the loose object
identified in "B" steam generator
and failed to evaluate
the
objects in "C" steam generator.
When this matter
was
brought to the attention of licensee
management,
evaluation
PCR-7251
was performed to formally document the
acceptability for leaving the objects in the "B" and
"C"
Although the inspector
agreed with the
conclusions
reached
by the engineering
evaluations,
the
documentation of the initial analysis
was considered
to be
weak.
b.
System Engineering
(71707)
The inspectors
reviewed the licensee's
actions to adjust the
excore nuclear instrumentation
current for the
new values
expected
from the low leakage
core.
The new core load reduced
the flux
present
at the reactor vessel
and the flux detected
by the excore
nuclear instrument detectors.
To account for this change the
intermediate
and power range nuclear instruments
were recalibrated
to adjust the old cycle instrument currents to predicted
new cycle
instrument currents.
The inspectors
reviewed calculations
by
licensee
personnel
of the predicted
values
performed in accordance
with procedures
EPT-008,
Intermediate
and
Power
Range Detector
Setpoint Determination,
and
EPT-009,
Intermediate
Range Detector
Setpoint Verification, and verified through reviews of work
packages
that these predicted setpoints
were implemented into the
nuclear instrumentation
channels.
The setpoints
were implemented
through calibration procedures
HST-I0044, HST-I0045, HST-I0046,
HST-I0047,
HST-I0167,
and HST-I0168.
The inspectors
found that
the methodology utilized was approved
by the fuel vendor.
During the review of procedure
EPT-009,
the inspector
noted that
step 7.2.5 directed the reactor engineer to select
acceptable
intermediate
range nuclear instrumentation trip currents to an
equivalent
value of less
than or equal
to 26 percent reactor
power.
The reactor engineer
selected
a current equal to
approximately
23 percent for conservatism.
Although it was
intended for this selected
current to be applied to the
calculations
contained
in procedure
EPT-008,
the inspector
noted
that
25 percent currents
were utilized instead.
The inspector
17
discussed
this observation with licensee
personnel
and
was
informed that although
a personnel
error had
been
made in the
calculations,
the
25 percent current actually selected
was within
the procedure
requirement of less than
26 percent
and was
therefore
considered
satisfactory.
Since the calculations of
these current adjustments
received
independent verification from
other licensee
personnel,
the error should
have
been identified by
the licensee.
Therefore the inspector considered
the performance
and verification of this calculation to be weak.
No violations or deviations
were identified.
Licensee Action on Previously Identified Engineering
Inspection
Findings
(92903)
(1)
(Closed)
Inspector
Followup Item 400/93-25-04:
Follow the
licensee's
activities to replace indicating bulbs
on the
TDAFW pump control panel
and identify similar bulb
applications
in other equipment.
Licensee
personnel
have replaced
the NB-120 type bulbs with
LED lamps in the control panel.
A search of the equipment
data
bases
and
CWDs identified two other applications of the
HB-120 bulbs consisting of the waste neutralization control
panel
and the containment
personnel
airlock control panel.
Licensee
personnel
concluded that the circuits associated
with these control panels
do not perform safety-related
functions
and therefore
no further action
was taken.
(2)
(Closed)
Inspector
Followup Item 400/93-07-01:
Follow the
licensee's
activities to prevent oil intrusion into charging
pump motors or establish
a periodic motor cleaning schedule.
The inspector
accompanied
licensee
personnel
on
a visual
inspection of the "A" and "B" CSIP motors to determine the
amount of oil accumulation
on the rotors
and stators.
The
motor windings for each
pump
had
a small
accumulation of oil
and dust, with the "B" motor exhibiting
a sl'ightly larger
buildup.
The licensee
discussed
this observation with the
equipment
vendor
who indicated that
a small accumulation of
oil and dust is
a normal,
expected
occurrence.
The vendor
recommended
that the licensee
continue to monitor the stator
temperatures
for each motor as
a means of determining
whether
any problems existed with cooling air flow as
a
result of the dirt buildup.
Recent stator temperature
data
for the three
CSIPs taken during the previous operating
cycle demonstrated
running temperatures
averaging
150
degrees
F.
The vendor stated that
150 degrees
was
acceptable
for these motors.
The licensee
plans to continue
monitoring the operating stator temperatures
and will base
any further actions
on future data.
(3)
18
(Closed) Violation 400/92-17-02:
Failure to correct
a
deficiency with the emergency diesel
generator starting air
system.
The inspector reviewed
and verified completion of the
corrective actions listed in the licensee's
response letter
dated
November 2,
1992.
This item was previously discussed
in
NRC Inspection
Report 50-400/93-08.
Licensee
personnel
have reviewed existing
and determined that
no
additional
ACFRs were required other than those already
written and submitted.
(4)
(5)
(Closed) Violation 400/93-21-02:
Failure to properly
review/approve
vendor procedures.
The inspector reviewed
and verified completion of the
corrective actions listed in the licensee's
response letter
dated
December
16,
1993.
The licensee
performed technical
and safety reviews of the vendor procedures
used in
safety-related
temporary leak repairs that were still in
place.
The licensee's
administrative
procedure
AP-032,
Procedure to Obtain
Non-Company
Labor and Services,
was
revised to clarify that vendor procedures
used for work on
safety-related
structures,
systems,
or components will be
formatted in accordance
with procedure
AP-005,
Procedures
Format
and Preparation,
and approved in accordance
with
procedure
AP-006,
Procedure
Review and Approval. If not
formatted
and approved in accordance
with procedures
AP-005
and AP-006, then the vendor procedure will be included in
the modification package
as
a work instruction
and will have
technical
and safety reviews along with the
PCR.
Designated
Contract Representatives
received training on the
new
requirements
of procedure
AP-032 prior to the start of the
current refueling outage.
(Closed) Violation 400/93-12-03:
Failure to establish
adequate
measures
to verify that designs
were technically
accurate with respect to the design basis for the
system.
The inspectors
reviewed
and verified completion of the
corrective actions listed in the licensee's
response letter
dated July 30,
1993.
The licensee
h'as completed
modification PCR-6925 to remove the motor operators for
valves
1AF-5 and
1AF-24 and replaced
them with manual
handwheels.
~
~
5.
Plant Support
a.
Plant Housekeeping
Conditions
(71707) - Storage of material
and
components,
and cleanliness
conditions of various areas
throughout
the facility were observed to determine whether safety and/or fire
~
~
~
19
b.
hazards
existed.
The inspectors
found plant housekeeping
and
material condition of components
to be satisfactory.
Radiological Protection
Program
(71707)
- Radiation protection
control activities were observed routinely to verify that these
activities were in conformance with the facility policies
and
procedures,
and in compliance with regulatory requirements.
The
inspectors
also reviewed selected
radiation work permits to verify
that controls were adequate.
Fuel sipping was performed during this outage to identify the
source of increased
RCS activity during the previous operating
cycles discussed
in
NRC Inspection
Report 50-400/93-04.
The
sipping process identified one fuel assembly,
HF04, which required
repair.
This assembly
was replaced
by another similar assembly
during the core re-load.
Extensive health physics
coverage
was provided to support the
implementation of PCR-0420,
RTD Bypass Elimination.
This coverage
included the use of cameras,
increased lighting, remote indicating
teledosimetry for workers located inside the containment bioshield
wall, and direct communications
between
health physics technicians
monitoring the job from an outside trailer and the workers inside
containment.
The inspector considered
overall performance
from
the health physics technicians
during this evolution to be good.
On one occasion
at the beginning of the
RTD bypass demolition for
RCS loop C,
a health physics supervisor actually stopped all work
associated
with the demolition when it was detected that the
workers were not coordinating work activities properly.
No work
was allowed to continue until
HP technicians
and workers regrouped
to resolve
problem areas.
This conservative
action
was taken
without regard to outage
schedule restraints,
and
was consistent
with the philosophy of ensuring that the job was done correctly
the first time with as few radiological or industrial incidents
as
possible.
In addition to the above activities, other controls
such
as the use of lead shielding helped
keep the total dose
recorded for the
RTD modification well below the industry average.
Approximately 60 man-rem
was expended for the modification.
Despite the overall
good performance
noted
above,
there
were
a few
minor examples of radiological control incidents which centered
around lack of attention to detail
and poor worker practices.
On
April 15, while observing inspections
on the "B" CSIP motor, the
inspector identified that the area
around the
pump baseplate
had
been
roped off and posted
as
a High Contamination
Area
(HCA).
The
dress
requirements for the area
were posted
as
shoe covers
and
gloves
as
a minimum.
Directly above the
HCA posting were three
signs posting the area
as
a Contamination
Area
(CA) with the
same
dress-out
requirements.
The inspector brought this dual posting
observation to the attention of HP technicians
who corrected
the
problem by removing the
CA signs
and upgrading the
HCA dress-out
requirements
to include overalls.
The technicians
noted that the
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c ~
d.
20
CA signs were remnants
from before the outage
when the area
was
still a contamination
area.
The inspector
noted that no work was
ongoing inside the area
posted
as
a
HCA and that the potential for
worker contamination
in this area
was minimal due to the fact that
HPs were then requiring lab coats,
gloves
and
shoe
covers
as
a
minimum for entry into the "B" CSIP room.
Other examples of
inadequate
radiological posting were identified by the licensee
throughout the outage.
In addition,
examples of poor worker
practices
were noted including
a case
where
a worker removed
a
breathing
apparatus
required
by his
RWP in order to communicate
with a fellow worker.
Also, seven
personnel
contamination
events
occurred early during the
RTD bypass, demolition which were
directly attributed to poor worker practices.
Plant program procedure
PLP-511,
Radiation Control
and Protection
Program,
Section 5.7.4.4
and Health Physics
Procedure
HPP-625,
Performance of Radiological
Surveys,
Section 10.4.2. 11 require
full protective clothing, including head cover, coveralls,
gloves,
shoe covers,
and rubbers
be worn as
a minimum for entry into HCAs.
The procedure further required that posting for HCAs include,
as
appropriate,
minimum dress
requirements
required for entry into
the area.
The inadequate
posting of dress
requirements
for the
HCA in the charging
pump room is contrary to the requirements
of
procedures
PLP-511
and
HPP-625
and is considered
to be
a
violation.
This violation is not being cited because
the
licensee's
efforts in correcting the violation meet the criteria
specified in Section VII.B of the Enforcement Policy.
Non-cited Violation (400/94-10-03):
Failure to adequately
post
dress
requirements for a High Contamination Area.
Except for the examples
noted
above involving lack of attention to
detail, the licensee's
performance
in the area of radiological
controls
was satisfactory.
Security Control
(71707)
- The performance of various shifts of
the security force was observed
in the conduct of daily activities
which included:
protected
and vital area
access
controls;
searching of personnel,
packages,
and vehicles;
badge
issuance
and
retrieval; escorting of visitors; patrols;
and compensatory
posts.
In addition, the inspector
observed
the operational
status of
closed circuit television monitors, the intrusion detection
system
in the central
and secondary
alarm stations,
protected
area
lighting, protected
and vital area barrier integrity,
and the
security organization interface with operations
and maintenance.
The licensee's
adherence
to security requirements
was found to be
satisfactory.
Fire Protection
(71707) - Fire protection activities, staffing and
equipment
were observed to verify that fire brigade staffing was
appropriate
and that fire alarms,
extinguishing equipment,
21
actuating controls, fire fighting equipment,
emergency
equipment,
and fire barriers
were operable.
A weakness
was identified in the licensee's fire protection
activities during RFO-5.
Fire-related
incidents occurred inside
the containment building and in the turbine building which
resulted
from less
than adequate
control of combustible materials
during hot work activities primarily performed
by contractor
personnel.
In addition,
one example of inadequate fire watch
activities was identified by the inspectors
and several
more
examples
were identified by licensee
personnel.
During
a two week period in April, four small fire-related
incidents occurred in the plant.
On April 4,
a small hand-held
dosimetry console
smoldered
inside containment.
This was due to
a
piece of welding slag from an upper elevation fell onto the
console
through
a small hole in the
protective covering used to
shield workers
and equipment
from just such
a hazard.
The
smoldering instrument
was extinguished within seconds,
but it was
not reported to the main control
room until several
days later
when the inspectors
and licensee
outage
management
began to
inquire about it.
On April 9 and
10, two small fire-related
incidents occurred in the turbine building.
Both involved poor
control of combustible material during welding activities.
In
each
case,
the fire was extinguished rapidly and the main control
room was properly notified.
On April 15,
a similar incident was
reported in the containment building.
A fire tech immediately
extinguished
a smoldering rag
and notified the main control
room.
None of the above incidents resulted in fires that were long
enough in duration
(10 minutes) to meet the licensee's
Unusual
Event declaration criteria.
Other examples of poor fire protection activities were also noted.
On April 6, the inspector
observed cutting/welding activities
associated
with the
AFW system pipe replacement
inside
containment.
The inspector
asked the associated
firewatch where
the Hot Work Permit was located.
The firewatch identified
a
permit posted
on
a nearby wall that
had expired the previous day.
It was later discovered that
a permit had
been
issued to allow the
work, but the copy required to be posted at the job site
had
been
lost.
The firewatch failed to identify the deficiency upon
assuming
his duties that afternoon.
In a separate
incident
on
April 8,
a roving firewatch in the containment building discovered
welding and grinding being performed
under
an expired hot work
permit.
Once again,
a firewatch was in place while the work was
ongoing,
but failed to identify the deficiency.
On April 15, licensee
personnel
discovered that there
was
no
firewatch personnel
monitoring the preheating of a certain section
of AFW piping inside containment.
It was determined
during the
licensee's
investigation that there
had
been
a firewatch posted
earlier,
but he was dismissed
when licensee
personnel
decided to
g
0
22
secure
the heating process.
Licensee
personnel
accidentally
shut
off power to
a different heating element leaving the subject
element
energized with no fire watch posted.
Upon discovery,
a
firewatch was immediately dispatched
to continue monitoring the
pre-welding activity on the heated
section of pipe.
Other
examples of deficiencies identified by the licensee
included low
pressure
gas containers
being
abandoned
at work sites following
completion of work.
Each
one of the items discussed
above
were documented
in ACFRs.
The inspectors
discussed
the above
problems with licensee
management
who acknowledged
the increased
trend in fire-related
incidents during the month.
Licensee
management
stated that
a
need existed to increase
the level of awareness
of potential fire
hazards
among firewatches
and the individuals performing the hot
work.
At the close of the inspection,
the licensee
was in the
process
of revising various fire protection procedures
and the
firewatch training lesson
plan to incorporate
more details
and
testing requirements.
Actions being considered
by the licensee
included required training for all personnel
performing hot work,
development of guidelines for containing hot work hazards,
and the
development of hot work permits requiring fire line supervision
signature signifying that all permit conditions
had
been
met prior
to starting work.
Fire Protection
Procedure
Control of Ignition Sources
Hot
Work Permit, Section 5.0, requires
maintenance
supervisors
to
ensure that all maintenance
and modification activities involving
hot work employ
a Hot Work Permit,
and that
a firewatch is
assigned.
It further states that the maximum duration of any Hot
Work Permit is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,
and if the work requiring the
HWP has not
been
completed
when the permit expires,
a new permit will be
issued
in accordance
with Section 8.0.
Section 8.0 states
that
where Hot Work Permits apply, associated
work will not commence
until the Firewatch
has inspected
the work site.
It further
states
that
a copy of the
HWP is then conspicuously
posted
in the
work area.
The examples of poor firewatch activities identified
above
were considered
to be in violation of the requirements
of
procedure
This violation will not be subject to
enforcement
action because
the licensee's
effort in identifying
and correcting the violation meet the criteria specified in
Section VII.B of the Enforcement Policy.
Non-cited Violation (400/94-10-02):
Failure to implement fire
protection procedures
adequately.
Exit Interview (30703)
The inspectors
met with licensee
representatives
(denoted
in paragraph
I) at the conclusion of the inspection
on Nay 6,
1994.
During this
meeting,
the inspectors
summarized
the scope
and findings of the
inspection
as they are detailed in this report, with particular emphasis
23
on the Violation 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.
No dissenting
comments
from the
licensee
were received.
Item Number
Descri tion and Reference
400/94-10-01
(VIO)
400/94-10-02
(NCV)
400/94-10-03
(NCV)
and Initialisms
Failure to establish
and implement
procedures,
paragraphs
2.a(2)(a)
and
4.a(3).
Failure to implement fire protection
procedures
adequately,
paragraph
5.d.
Failure to adequately
post dress
requirements
for a high contamination
area,
paragraph
5.b.
ACFR
CFR
CSIP
CWD
ESFAS-
FBR
FHB
KV
HSIV
NAD
NRC
PNSC
n System
tioning
Asea
Brown Boveri
Adverse Condition Feedback
Report
Corrective Action Program
Component Cooling Water
Code of Federal
Regulations
Charging Safety Injection
Pump
Chemical
and Volume Control
System
Control Wiring Diagram
Emergency
Core Cooling System
Emergency Diesel
Generator
Engineered
Safety Feature
Engineered
Safety Feature Actuatio
Feedback
Report
Fuel Handling Building
Final Safety Analysis Report
High Contamination
Area
Health Physics
Heating, Ventilation and Air Condi
Inservice Inspection
Kilovolt
Motor Control Center
Mechanism Operated
Cell
Motor Operated
Valve
Nuclear Assessment
Department
Nuclear Regulatory
Commission
Plant
Change
Request
Process
Instrument Cabinet
Plant Nuclear Safety Committee
Power Operated Relief Valve
RCB
SCO
SSPS
TDAFW-
TS
VAC
Reactor
Containment Building
Pump
Refueling Outage
Residual
Heat
Removal
Resistance
Temperature
Detector
Reactor Trip System
Radiation
Work Permit
Senior Control Operator
Safety Injection
Solid State Protection
System
Turbine Driven Auxiliary Feedwater
Technical Specification
Voltage Alternating Current
Work Request