ML17291A967
| ML17291A967 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 08/09/1995 |
| From: | Wong H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML17291A965 | List: |
| References | |
| 50-397-95-20, NUDOCS 9508220165 | |
| Download: ML17291A967 (25) | |
See also: IR 05000397/1995020
Text
ENCLOSURE
2
U.S.
NUCLEAR REGULATORY COMMISSION
REGION I V
NRC Inspection
Report:
50-397/95-20
License:
Licensee:
Washington Public
Power Supply System
3000 George
Way
P.O.
Box 968,
MD 1023
Richland,
Faci)ity Name:
Washington Nuclear Project-2
(WNP-2)
Inspection At:
WNP-2 site near Richland,
Inspection
Conducted:
June
4 through July 15,
1995
Inspectors:
R.
C. Barr, Senior Resident
Inspector
D. L. Proulx,
Resident
Inspector
Approved
H. J.
Wong,
C
'
rope
rane
Ins ection
Summar
Areas
Ins ected:
Routine,
announced
inspection
by resident
inspectors of
control
room operations,
licensee
action
on previous
inspection findings,
operational
safety verification, surveillance
program,
maintenance
program
and
licensee
event reports.
Results
and Assessments:
~0erations
Operators
demonstrated
weaknesses
in properly performing
and overseeing
control
rod manipulations.
An operator incorrectly positioned
a control
rod during surveillance testing.
The control
rod second verifier accepted
a phone call during rod manipulation contrary to management's
expectations.
Operators
made three errors
in documenting control rod
manipulations.
There
have
been similar problems
in the past in this area
(Section 3).
The plant startup
coming out of the refueling outage
was generally
performed conservatively,
in
a deliberate
manner,
and in accordance
with
licensee
procedures
(Section 3).
9508220i65
950816
ADOCK 05000397
8
~
Operators failed to restore
two nonsafety-related
circuit breakers
to
their proper position after removing clearance
orders.
Problems
in the
processing of clearance
orders
have
been
a problem in the past
(Section 3).
~
Operators
made
two administrative errors in the Limiting Condition for
Operation
(LCO) log (Section 3).
Maintenance
A safety significant
human performance error occurred
when maintenance
craftsmen
removed
an incorrect lead during surveillance testing
and caused
a reduction in reactor water level
(Section 2.2).
~
Due to inadequate
procedures,
the control
rod drive
(CRD) housing support
was not installed per design
requirements
even
though the support
had
been
accepted
as installed properly (Section
2. 1).
~
Surveillances
were otherwise
performed properly and in accordance
with
procedures
(Section 6).
~
Maintenance
tasks
were performed
and documented
properly (Section 7).
~E
~
A design
change
associated
with containment
pressure
indication was
generally well designed
and
implemented
(Section
4. 1).
~
The requirements
of 10 CFR 50.59
and licensee
procedures
to perform
a
safety evaluation
were not fully implemented for two temporary
modifications (Section 4.2).
Plant
Su
ort
~
Plant housekeeping
and cleanliness
were very good, with the exception of a
fan which was inadequately
restrained
near the safety-related
automatic
depressurization
system
gas bottles
(Sections
3 and 5.3).
Summar
of Ins ection Findin s:
~
Violation 397/9520-01
(Section
2. 1.5)
was opened.
~
Violation 397/9520-02
(Section 2.2.3)
was opened.
~
Violation 397/9520-03
(Section 4.2.3)
was opened.
~
Licensee
Event Report 397/95-01,
Revision
0 (Section
9. 1) was closed.
~
Licensee
Event Report 397/95-02,
Revision
0 (Section
9. 1)
was closed.
~
Licensee
Event Report 397/95-04,
Revision
0 (Section
9. 1)
was closed.
Licensee
Event Report 397/95-05,
Revision
0 (Section
9. 1) was closed.
~
Licensee
Event Report 397/95-06,
Revisions
0 and
1 (Section
9. 1) were
closed.
~
Licensee
Event Report 397/95-07,
Revision
0 (Section
9. 1)
was closed.
~
Unresolved
Item 397/93-18-02
(Section 8.1)
was reviewed
and closed.
Attachments:
Attachment
1 - Persons
Contacted
and Exit Meeting
~
Attachment
2 - Acronyms
DETAILS
1
PLANT STATUS
At the beginning of the inspection period,
the plant was in Mode
4 (cold
shutdown) during Refueling Outage
R10.
At the conclusion of the outage,
the
plant entered
Mode
2 (startup)
on June
9,
1995.
On June
13,
1995,
the plant
entered
Mode
1.
The main generator
was temporarily synchronized
to the grid.
The reactor
was shut
down to Mode
3 (hot shutdown)
on June
15,
1995, directed
by the Bonneville Power Administration due to
an excess
generation
of
electricity.
The licensee
used this period to conduct maintenance
that could
only be performed during cold shutdown.
The reactor entered
Mode
2 on
June
30,
1995.
The plant entered
Hode
1
on July 3,
1995.
One hundred percent
power was achieved
on July 14,
1995.
2
ONSITE FOLLOWUP TO EVENTS
(93702)
2. 1
De raded
CRD Housin
Su
ort
Background
On June
6,
1995,
the inspector
performed
an
end of refueling
outage closeout
walkdown of the drywell.
To serve
as the licensee's
final
drywell walkdown, the operations
manager
and the drywell coordinator
accompanied
the inspector.
The inspector
noted that the drywell was very
clean except for a few small pieces of polyethylene
bags
and wiring
insulation.
While looking under the reactor vessel,
the drywell coordinator
identified that
a jam nut was missing from the
CRD housing support.
2. 1. 1
CRD Housing Support Function
The
FSAR and the
Bases for-the Technical Specifications
(TS) provide
a
description of and the safety analysis for the
CRD housing support.
FSAR Section 4.6.2.
1 states,
"The
CRD housing supports
prevent
any significant
nuclear transient
in the event
a drive housing breaks
or separates
from the
bottom of the reactor vessel."
Section 4.6. 1.2.2 states,
"The
CRD housing supports
meet the following safety
design
bases:
a) following a postulated
CRD housing failure, control rod
downward motion shall
be limited so that
any resulting nuclear transient
could
not
be sufficient to cause
fuel damage;
b) the clearance
between
the
housings
and the supports
shall
be sufficient to prevent vertical contact
stresses
caused
by thermal
expansion
during plant operation."
TS 3.1.3.8 requires
the
CRD housing
support to be in place in operational
Modes
1,
2,
and 3.
TS 4.1.3.8 states
"the
CRD housing support shall
be
verified to be in place
by
a visual
inspection prior to startup
any time it
has
been disassembled
or when maintenance
has
been
performed
in the
housing support area."
I
At normal operating
temperature,
a gap of approximately I/4 inch exists
between
the
CRD housing
and the supports.
At lower temperatures
the gap is
greater.
2. 1.2
Sequence
of Events
On June
1,
1995,
the shift manager
contacted
the duty maintenance
manager
and
requested
that
a maintenance
supervisor
sign off Step
29 of Plant Procedures
Manual
(PPM) 3. 1. 1, "Master Startup Checklist."
Step
29 oF
PPH 3. 1. 1 verified
that the
CRD housing support
was in place per
TS 4. 1.3.8.
The duty
maintenance
manager
contacted
an instrument
and controls
( I&C) supervisor
(who
was knowledgeable
of the housing support structure)
and requested
he initial
Step
29.
The
I&C supervisor
had previously observed
work undervessel
and
noted that the
CRD housing support
was in place.
He initialed off Step
29.
The
I&C supervisor
and the shift manager
had noted that Step
29 of PPM 3. 1. I
required
a mechanical
maintenance
supervisor's initials; however,
they
considered
the
ILC supervisor to be technically qualified for this
verification.
During the drywell inspection with the inspector
on June
6,
1995,
the drywell
coordinator noted that
a jam nut on the
CRD housing support
was missing.
Subsequent
licensee
reinspection
of the housing support structure identified
the gaps
between
the
CRD housings
and the supports
were not within design
tolerances.
On June 8,
1994,
at the
NRC's request,
the licensee participated
in a
teleconference
to discuss
the
sequence
of events
and the safety significance
oF this event.
The licensee
described
the information contained
in
Problem Evaluation
Request
(PER)
295-0718 concerning
the as-found condition of
the
CRD housing support.
The licensee
also presented
GE's conclusions
concerning
the operability of the system.
During this teleconference,
the
NRC questioned
the identification of the
deficiency during
a drywell tour rather
than during the performance of TS
surveillance.
The licensee
stated that they had not yet performed this
surveillance.
They indicated the discrepancies
would have
been
found during
the surveillance.
The
NRC representatives
requested
that the licensee verify
this information.
Later on June
8,
1995,
the licensee
discovered
that the surveillance
(PPH
3 . 1. 1, "Master Startup Checklist, " Step
29) that veriFied the correct
installation of the
CRD support
housing
had
been
signed off.
Furthermore,
the
licensee
found that
an
I&C supervisor initialed the step rather than
a
mechanical
maintenance
supervisor,
as required
by
PPM 3. 1. 1, "Master Startup
Checklist," Step
29.
The licensee
found that the signoff for the surveillance
had already occurred.
The licensee notified the
NRC of these findings.
On June
9,
1995,
the licensee
conducted
an Incident Review Board
(IRB) to
determine
the circumstances
of the event
and corrected
the material
discrepancies.
2. 1.3
Licensee
Evaluation
and Corrective Action
t
PER 295-0718 discussed
the operability of the housing support.
The
PER
explained that the licensee
contacted
General
Electric concerning
the proper
configuration of the
CRD housing supports.
GE stated that the
1 a
. 12 inch
specification
was the standard for most housing supports.
However,
the
smaller dimension
(7/8 inch)
was found to be more critical than the
maximum of
1 1/8 inch.
GE provided
an analysis that would allow a maximum gap of
1 1/2 inch.
Because all the gap discrepancies
were in the 7/8 -
1 1/2 inch
band,
the licensee
considered
the
The licensee
corrected
the material deficiencies
and verified the installation of the
housing support.
The
IRB and
PER 295-0730 discussed
the improper signature
made
by the
IKC
supervisor
in
PPM 3. 1. 1.
The
IRB recommended
that
a checklist
be developed
to
verify the design
requirements
of the
CRD housing support.
PPH 3. 1. 1 was
revised to allow for flexibility in designating
a qualified individual to sign
Step
29.
The licensee
conducted
a lessons-learned
session
with maintenance
personnel.
2. 1.4
Previous History
On June
21,
1994,
craftsmen
completed installing and verifying correct
configuration of the
CRD housing support during Refueling Outage
R9.
Following the reinstallation of the housing support,
the craftsmen
documented
that two jam nuts
had
been missing since Refueling Outage
R7.
Therefore,
the
CRD housing support
may have
been in a degraded
condition for two operating
cycles.
Following the craft verification of the
CRD housing support,
the
system engineer
performed
a walkdown of the
CRD housing support.
The system
engineer
not only noted that two jam nuts
had
been missing,
but that the
housing support
was improperly spaced.
The licensee initiated
PER 294-0698 to
document this occurrence.
The licensee initiated
a work order to correct
these deficiencies.
The corrective actions for PER 294-0698 included:
(1) performing
an
evaluation to determine if additional training was required
on
CRD housing
support installation (the evaluation
concluded additional training should
performed to be completed
by Refueling Outage
R11);
(2) providing training for
maintenance
supervision
on work order paokage
closeouts;
and
(3) counseling
the applicable craft supervision
on maintaining
a "questioning attitude."
These corrective actions
were completed
in December
1994.
2. 1.5
Inspector
Findings
The inspector
found that it had
been
the licensee's
practice
in previous
verifications of the installation of the
CRD housing support,
a checklist or
a
procedure
was not used to identify the specific criteria to be verified.
The
inspector
considered
that without
a detailed checklist or procedure,
the
critical parameters
of the
CRD housing support could not
be assured.
TS 6.S'il.d requires
a procedure for each surveillance listed in the
TS and the
procedure
be implemented.
In addition,
FSAR Section 4.6.3.2.
1 states,
"When
the support structure
is reinstalled, it is inspected for correct
assembly
with particular attention to maintaining the correct
gap between
the
flange lower contact surface
and the grid."
The licensee failed to have
a
procedure with appropriate criteria for inspection of the
CRD housing support
which reflected the commitments
in the
FSAR.
The failure to have
a procedure
appropriate
to the circumstances
with appropriate quantitative or qualitative
acceptance criteria for inspection of the
CRD housing support is
a violation
of 10 CFR Part 50, Appendix B, Criterion
V (Violation 397/9520-01).
2. 1.6
Safety Significance
Because
the as-found
gap in the
CRD housing support would not result in
excessive
stresses,
a failure of the housing support would not result in rod
ejection of greater
than
3 inches
and
one jam nut missing would not affect the
operability of the housing support, this event
had low safety significance.
2. 1.7
Conclusions
The licensee
had not established
appropriate
controls to assure
correct
configuration of the
CRD housing support resulting in
a violation of
Appendix B, Criterion V, of 10 CFR Part 50.
The licensee's
previous
corrective actions for improper installation of the
CRD housing support were
not adequate
to prevent the problems
encountered
during Refueling Outage
R10.
The licensee's
investigation
was not thorough
because it did not address
the
missed opportunities
to correct this problem
and did not identify that
personnel
who performed the verification of the
CRD housing installation did
not use procedures
appropriate
to the circumstance.
2.2
Reactor
Vessel
Level Transient
Due to Wron
Lead Lifted
2.2. 1
Unplanned
Level Transient
On July 12,
1995, during performance of surveillance
PPM 7.4.3.7.5. 18,
Revision 6, "Accident Monitoring Instrumentation
Reactor
Core Isolation
Cooling
(RCIC) Flow Indication Channel
Check," reactor
pressure
vessel
(RPV)
level decreased
from +36 inches to +23 inches
due to
IKC technicians lifting
of an incorrect lead.
The licensee initiated
PER 295-0845
and conducted
an
IRB.
2.2.2
Licensee
Investigation
and Corrective Actions
The licensee's
investigation
found that
PPM 7.4.3.7.5. 18,
Steps
7.3.2
and
7.3.3 required the technicians
to identify and lift leads for Terminal
Block E51A-SRU-1 in the rear of Cabinet
H13-P612.
(Cabinet
H13-P612 is
divided into two sections,
one for the
RCIC system
and
one for the feedwater
level control system.)
The two technicians
assigned
to perform the
surveillance
worked in the feedwater level control
system section of
Cabinet
H13-P612
instead of the
RCIC system section.
The technicians
located
Terminal Block C34A-SRU-1 instead of E51A-SRU-1
and lifted the wrong leads.
The two technicians
indicated that they believed the
"SRU-1" portion of the
label
was the
key indicator
and
removed the leads;
however,
the procedure
was
particularly clear in requiring the field wires associated
with E51A-SRU-1
be
lifted.
The leads that the technicians lifted provided
one of the four inputs into the
total
steam flow portion of the feedwater control system.
With one of the
inputs
removed,
the circuit sensed
a decrease
in steam flow.
This created
a
feed
and
steam flow mismatch.
The feedwater control
system sent
a signal for the feed
pumps to decrease
flow.
This decrease
in feed flow caused
RPV water level to decrease.
When
operators
received
the
RPV low level annunciator
at +30.5 inches,
they
directed the technicians
to restore
any leads that
had
been lifted.
When the
technicians
restored
the leads,
the level transient
stopped..
The surveillance
was stopped until after the event investigation
was completed.
The licensee's
corrective actions for this event included:
counseling
the
personnel
involved, providing self-checking training for the individuals,
and
using operations
personnel
for second verification of selected
surveillances.
In addition,
although the operators
responded
well to the event,
the licensee
provided additional training on feedwater control
system failures (e.g.,
taking the system to manual
or single element control).
2.2.3
Conclusions
The failure of the technicians
to implement
PPH 7.4.3.7.5.18,
Steps
7.3.2
and
7.3.3 that required the identification and lifting of leads for Terminal
Block E51A-SRU-1
was
a violation of TS 6.8. l.d (Violation 397/9520-02).
This
violation had safety significance
in that the event resulted
in
a level
transient that
had the potential to result in a reactor
on low reactor
water level.
The licensee's
corrective actions
were adequate.
This event
represented
an additional
human performance error that resulted
from the
failure of individuals to self-check
and inadequate field supervision.
3
PLANT OPERATIONS
(71707,
92901)
3. 1
Plant Tours
The inspectors
toured the following plant areas:
Reactor Building
Control
Room
Diesel
Generator
Building
Radwaste
Building
Service Water Buildings
Technical
Support Center
~
Turbine Generator
Building
~
Yard Area and Perimeter
3.2
Ins ector Observations
3.2.
1
Control
Rod Manipulations
The inspectors
observed
portions of the plant startup
coming out of Refueling
Outage
R10.
The inspectors
noted that operators
generally performed in
a
conservative
and deliberate
manner.
The following instances
indicate that operator manipulation of control rods
requires
improvement.
3.2. 1. 1
Background
Licensee
PPM 1.3. 1,
"Department Policies
and Practices,"
Section 4.6.2.f,
states
"the second
person at H13-P603 must.
.
.
have
no other concurrent
duties while control rod manipulations
are in progress."
" Licensee
management
stated that the intent of this step
was to ensure that personnel
would not
be
distracted
during rod movement.
3.2. 1.2
Operator Distractions
On July 4,
1995, operators
were moving control rods to increase
reactor
power.
A control
room operator
(CRO) m'anipulated
the control
rods while the shift
technical
advisor
(STA) verified the correct
movements.
The inspector
observed
that while
a control rod was being
moved the lead
CRO received
a
phone call
and
passed
the telephone
to the
STA.
When rod movement
was ceased
for that control rod, the
CRO stopped control rod movement until the
completed his phone call.
The inspector contacted
the shift manager
and explained his observation.
The
shift manager
discussed
this event with the
STA and informed the inspector
that this instance
was not safety significant because
no rod movement errors
were
made
and the
CRO only had two notches left until the rod was full out
when the
STA accepted
the phone call.
The operations
manager
indicated to the
inspector that the
STA did not meet his expectations.
3.2. 1.3
Administrative Errors Associated with Control
Rod Manipulations
The inspector
reviewed the rod manipulation
sheets
to determine if the
sequence
of control rod movements
was correct
and the movements
were properly
documented.
The inspector
found three
instances
in which operators correctly
manipulated
but failed to sign the rod pull sheets.
The
inspector notified the shift manager,
who directed
the
CROs involved to sign
for the rod movements.
-10-
3.2. 1.4
Rod Control Error During Surveillance
On July 13,
1995,
the licensee initiated
PER 295-0846 to document
an error
during control
rod manipulation.
During performance of PPH 7.4. 1.3. 1.2,
"Control
Rod Exercising," the
CRO inserted
a control
rod an extra notch.
PPH 7.4.1.3. 1.2 requires
operators
to insert control
rods
one notch from
position
48 to 46 and then return to 48 (control rod exercising).
On July 13,
1995, while moving Control Rod 06-23 from position
48 to'46,
the operators
received
an annunciator
on the stator water cooling system.
The second
verifier left the rod console to address
the alarm, while CROI waited at the
panel.
After completing the annunciator
alarm response
actions,
the second
verifier returned to complete the control rod exercises.
Operators
mistakenly
inserted
Control Rod 06-23 again,
such that the control rod was then located
at position 44.
The operators
informed the control
room supervisor
and
initiated
a
PER.
The licensee
determined that the root cause of this error was the failure to
self-check.
A contributing cause of the error was that operators
were
distracted
by the alarm.
As corrective action,
the operations
manager
counselled
the involved personnel.
3.2. 1.5
Safety Significance
The inspectors
determined that this event
was not technically safety-
significant.
The control rod was mispositioned
one notch,
which did not
impact reactor
power,
thermal limits, or the preconditioning envelope.
3.2. 1.6
Inspector Conclusions
NRC Inspection
Reports
50-397/93-24,
94-27,
94-32,
and 94-33 all discuss
concerns with control rod operations'hile
these
discrepancies
had minor
safety significance,
they represent
a continuing lack of sensitivity for the
importance of deliberate
and thoughtful manipulation of control rods.
3.2.2
Operating
Logs
and Records
The inspectors
reviewed operating
logs
and records
against
TS and
administrative control procedure
requirements.
3.2.2.
1
LCO and Inoperable
Log
On June
13,
1995,
the inspector
reviewed the
LCO log and identified two
administrative errors.
The entry for the inoperability of Sample
Point
8 (SP-8),
an inline conductivity monitor, discussed
actions to be taken
if chemistry
was out of specification,
rather than actions to be taken if both
sample points were inoperable.
The entry for the inoperability of Temperature
Element
SW-TE-2B discussed
actions to be taken for Hodes
4 and 5, yet the
-11-
plant was in Mode 1.
These errors
had the potential
to mislead other
operators.
The inspector notified the shift manager
and the entries
were
corrected.
3,2.3
Monitoring Instrumentation
The inspectors
observed
process
instruments for correlation
between
channels
and for conformance with TS requirements,
and
no discrepancies
were
identified.
3.2.4
Shift Manning
The inspectors
observed
control
room and shift manning for conformance with
TS,
and administrative
procedures.
The inspectors
also
observed
the attentiveness
of the operators
in the execution of their duties.
The inspectors
concluded that shift manning
was in conformance with the
applicable
requirements
and operators
were generally attentive to duties.
The
control
room was observed
to be free of distractions
such
as nonwork-related
radios
and reading materials.
3.2.5
Equipment
Lineups
The inspectors verified that valves
and electrical
breakers
were in the
position or condition required
by TS and administrative
procedures
for the
applicable plant mode.
This verification included routine control
board
indication reviews
and conduct of partial
system lineups.
Appropriate entry
into TS
LCO was verified by direct observation.
The inspectors
performed verification that selected
breakers
were in the
proper position following the licensee's
completion of the breaker lineups per
PPM 2.7. 1B,
"480 Volt and Below AC Electrical
Power Distribution System,"
prior to plant startup.
3.2.5.1
Inspectors
Observations
On June
5,
1995, the inspectors
found Breaker
PP7BC-13
in the
"ON" position
rather than
"OFF" as required
by
PPM 2.7. 1B.
Breaker
PP-7BC-13
was
a spare
breaker.
The licensee initiated
PER 295-0708 to document this issue.
The
inspector
noted that Breaker
PP7BC-13
had
been repositioned
to support
a
clearance
order following completion of PPM 2.7. 1B, but had not been restored
to its proper position following removal of the clearance
order.
The licensee
repositioned
the breaker
and verified its position'.
On June
7,
1995,
the inspector
found Breaker
PP-7AE-38 in the
"ON" position
rather than the
"OFF" position,
as required
by
PPM 2.7. 1.B.
This breaker
interrupted
power to
a system
no longer in use.
The licensee initiated
PER 295-0722 to document this issue.
The inspector
noted that Breaker
PP-7AE-
38 had
been repositioned
to support
a clearance
order, following completion of
-12-
PPH 2.7. IB, but had not been restored
to its proper position following removal
of the clearance
order.
The licensee
repositioned
the breaker
and verified
its position.
The inspector considered
the licensee's
corrective actions for these
two
findings adequate
and noted that these
problems
were not safety
significant'he
failures to properly position these
breakers
constitute
a violation of
minor significance
and are being treated
as
a noncited violation, consistent
with Section
IV of the
These
problems
represent
additional
human performance
errors associated
with
the clearance
order process.
3.2.6
Equipment
Tagging
The inspectors
observed
selected
equipment,
for which tagging requests
had
been initiated, to verify that tags
were in place
and the equipment
was in the
condition specified.
The inspectors
noted the following issues
with respect
to clearance
tagging.
3.2.7
General
Plant
Equipment Conditions
The inspectors
observed
plant equipment for indications of system leakage,
improper lubrication, or other conditions that would 'prevent the system
from
fulfillingits functional requirements.
were observed
to
ascertain their status
and operability.
3.2.7.
1
Improperly Restrained
Equipment
During
a tour of the
441 foot elevation in the reactor building, the inspector
noted that
a piece of equipment did not appear to be properly secured.
A
reactor building ventilation fan (approximately
4 foot diameter)
was tied down
onto
a maintenance
cart that
was only 2 feet wide.
While the configuration
met licensee
procedure
requirements, it appeared
to be seismically unstable
and
was very close to the automatic depressurization
system
gas bottles.
The
inspector notified the shift manager,
who had the fan
and cart moved.
3.3
En ineered
Safet
Features
Walkdown
The inspectors
walked
down selected
engineered
safety features
(and systems
important to safety) to confirm that the systems
were aligned in accordance
with plant procedures.
During the walkdown of the systems,
items
such
as
hangers,
supports,
electrical
power supplies,
cabinets,
and cables
were
inspected
to determine that they were operable
and in
a condition to perform
their required functions.
Proper lubrication and cooling of major components
were also observed for adequacy.
The inspectors
also verified that certain
system valves were in the required position
by both local
and remote position
indication,
as applicable'
-13-
The inspectors
walked
down selected
portions of the following systems:
Diesel
Generator,
Divisions
1,
2,
and
3
Low Pressure
Coolant Injection, Trains A, B,
and
C
Low Pressure
(LPCS)
High Pressure
Residual
Heat
Removal Trains
A and
B
Standby
Gas Treatment
(SGT)
125-Vdc Electrical Distribution, Divisions
1 and
2
250-Vdc Electrical Distribution
The inspectors
noted that the engineered
safety features
systems
were
generally in good material condition
and were aligned in accordance
with
applicable licensee
procedures
for the portions walked down.
4
ONSITE ENGINEERING
(37551,
92903)
The inspectors
performed
inspections
of the following onsite engineering
related activities during this inspection period:
4. 1
Desi
n Chan
e Packa
e Review
The inspector
reviewed Plant Modification Record
(PMR) 91-0338-OA which
installed tubing
and gauges
to indicate drywell and wetwell pressures
locally.
This
PMR installed
CMS-PI-9 and CMS-PI-10.
The inspector
reviewed the
PHR package,
the applicable
procedures,
the safety evaluation,
the top tier
drawings,
and the implementing work orders to determine if the modifications
were performed in accordance
with licensee
procedures.
The inspector
performed
a walkdown of the
new equipment to ensure that the applicable
requirements
were met.
The inspector
determined that the design
and
implementation of PHR 91-0338 were performed in a thorough
and conservative
manner.
4.2
Review of Tem orar
Modification Re uest
THR
Lo
The inspector
reviewed selected
entries
in the licensee's
THR log to determine
and lifted leads
and jumpers
were installed in
accordance
with NRC and license requirements,
The inspector
noted the
following concern with the
TMR log.
4.2.1
Removal of Valve Position Indication
The inspector
reviewed
TMR 95-030,
which removed position indication from the
disc of Valve RCIC-V-66,
a testable
and
a containment isolation
valve.
The licensee
implemented
the
THR because
the position indication
lights for RCIC-V-66 provided the operators
with false indication that the
valve was closed
when it was actually open.
No
10 CFR 50.59 evaluation
had
been
performed to support
the removal of the position indication.
The
-14-
inspector
reviewed the
TMR sheets,
the licensee's
evaluation,
the applicable
procedures
and drawings,
and
PER 295-0641.
4.2.2
Requirements
10 CFR 50.59(b)(1) states,
"The licensee
shall maintain records of changes
in
the facility.
.
.
made pursuant
to this section,
to the extent that these
changes
constitute
a change to the facility and described
in the safety
analysis report.
These
changes
must include
a written safety evaluation
which
provides
a basis for determination that the change,
test,
or experiment
does
not involve an unreviewed safety question."
PPH 1.3.43,
Revision 6, Section
5. 1. I.b states
a
10 CFR 50.59 review and, if
required,
a safety evaluation will be required only for temporary
modifications not covered
by
a previously evaluated
and approved
document
such
as
a plant procedure
or
a
PHR.
Additionally, Attachment
7.3 states
a change
to the design of the plant
as described
in the licensing basis
document will
require safety evaluation.
4.2.3
Licensee
Assessment
of the Requirement for a Safety Evaluation
The licensee's
screening
evaluation of the issue
concluded that
a
evaluation for this
TMR was not required for the following two reasons:
(1) there were
no licensing basis
documents
that required the position
indication to be in place,
and
(2)
10 CFR 50.59 evaluations
are not required
for the interim disposition of problems if the timeliness for corrective
action completion is such that it can
be demonstrated
for a particular case
that the degraded
or nonconforming condition has not or will not become
a
defacto
change.
4.2.4
NRC Inspection
The inspector
reviewed the
FSAR and found that Section 5.4.6.2.4 described
Valve RCIC-V-66 stating,
"valve test provisions
are provided including limit
switches to indicate disc movement."
Based
on the inspector's
finding, the
licensee initiated
PER 295-0826 to document this problem.
The licensee
determined that
an inadequate
search of the licensing basis
documents
had
been
performed.
The system engineer
had performed
a key word search
on
an
electronic version of the
FSAR and
had not used the appropriate
key word.
As
corrective actions,
the licensee
performed
a
10 CFR 50.59 safety evaluation
and concluded
no unreviewed safety questions
were identified.
On July 13,
1995,
the licensee initiated
PER 295-0864 which indicated that
TMR 95-050
had
also
been
implemented without the required
10 CFR 50.59 safety evaluation.
The inspector
reviewed
PPH 1.3. 12A.
Attachment 8.2 of this procedure
states
a
10 CFR 50.59 evaluation is not required for an interim disposition of a
problem if the timeliness for corrective action completion is such that it can
be demonstrated- for a particular case that the degraded
or nonconforming
condition has not or will not become
a defacto
change.
While it appears
that
the change
was not intended to be permanent,
based
on the location of
-15-
Valve RCIC-V-66 (inside containment),
the
TMR would likely be in place for
approximately
I year
and, therefore,
this change
was essentially
a long-term
change.
4.2.5
Conclusions
The failure to perform
a written safety evaluation for the implementation of
THR 95-030 is
a violation of 10 CFR 50.59 {Violation 397/9520-03).
While the
violation itself had little safety significance,
the
NRC is concerned
that the
inadequate
electronic
search
performed for the safety evaluation creates
the
potential for a more significant error to occurr.
In addition,
the licensee
identified another
case
in which
a safety evaluation
had not been
performed.
5
PLANT SUPPORT ACTIVITIES
(71750)
The inspectors
evaluated
plant support activities based
on observation of work
activities,
review of records,
and facility tours.
The inspectors
noted the
following during this evaluation.
5. 1
Fire Protection
The inspectors
observed fire fighting equipment
and controls for conformance
with administrative
procedures.
The inspectors
noted that
a high number of
fire impairments existed for which fire tours were being conducted
because
of
concerns with Thermo-Lag
and fire seals,
and the number of propped
open fire
doors to support work.
5.2
Radiation Protection Controls
The inspectors
periodically observed
radiological protection practices
to
determine whether the licensee's
program
was being
implemented
in conformance
with facility policies
and procedures
and in compliance with regulatory
requirements.
The inspectors
also observed
compliance with radiation work
permits,
proper wearing of protective
equipment
and personnel
monitoring
devices,
and personnel
frisking practices.
Radiation monitoring equipment
was
frequently monitored to verify operability and adherence
to calibration
frequency.
5.3
Plant
Housekee
in
The inspectors
observed
plant conditions
and material
and equipment
storage
to
determine
the general
state of cleanliness
and housekeeping.
Housekeeping
in
the radiologically controlled area
was evaluated
with respect
to controlling
the spread of surface
and airborne contamination.
Housekeeping
was observed
to be generally
good during the inspection period.
S.4
~Securit
The inspectors periodically observed
security practices
to ascertain
that the
licensee's
implementation of the security plan was in accordance
with site
procedures,
that the search
equipment at the access
control points
was
operational,
that the vital area portals
were kept locked
and alarmed,
that
personnel
allowed access
to the protected
area
were
badged
and monitored,
and
that the monitoring equipment
was functional.
No problems
were noted during
these observations.
5.5
Emer enc
Plannin
The inspectors
toured the
Emergency Operations Facility, the Operations
Support Center,
and the Technical
Support Center
and ensured
that these
emergency facilities were in
a state of readiness.
Housekeeping
was noted to
be very good
and all necessary
equipment
appeared
to be functional.
5.6
Plant Chemistr
The inspectors
reviewed chemical
analyses
and trend results for conformance
with TS and administrative control procedures.
Plant chemistry
was
satisfactory during this inspection period.
5.7
Conclusions
Plant support
performance
was generally
good during this inspection period.
6
SURVEILLANCE TESTING
{61726)
The inspectors
reviewed
TS surveillance tests
on
a sampling basis to verify
that:
~
a technically adequate
procedure
existed for performance of the
surveillance tests;
~
the surveillance tests
had
been
performed at the frequency specified in
the
TS and in accordance
with the
TS surveillance
requirements;
and
~
test results satisfied
acceptance
criteria or were properly dispositioned.
The inspectors
witnessed
portions of the following surveillance tests:
Procedure
Descri tion
7.4.4.3.2.6
7.4.1.4.1.3
HPCS/Reactor
Coolant
System
Interface
Valve and Keepfill
System
Checks
Rod Worth Minimizer Control
Rod Sequence
Verification
-17-
7. 4. 1. 3. 2
7.0.0
Control
Rod Scram
Time Testing
Shift and Daily Instrument
Checks
Overall, surveillance testing
was performed
and documented
properly.
7
MAINTENANCE OBSERVATIONS
(62703)
During this period,
the inspectors
observed
and reviewed documentation
associated
with maintenance
and problem investigation activities to verify
compliance with regulatory requirements
and with administrative
and
maintenance
procedures,
required quality assurance/quality
control
involvement,
proper
use of clearance
tags,
proper equipment
alignment
and
use
of jumpers,
personnel
qualifications,
and proper retesting.
7. 1
Control
Rod Multi lexer Card
Re lacement
On July 4,
1995,
the inspector
observed
Work Order Task VN59-01,
"Replace
Multiplexer Card for Control Rod 18-11."
This work required deenergization
of
the entire rod position indication system.
The deactivation of the rod
position indication system required
the plant to enter
a 1-hour
LCO action
statement.
The inspector
observed
that the licensee
performed
thorough
planning
and
a good preevolution briefing.
The inspector
noted that during the work the craftsmen
did not use
a grounding
strap.
The inspector
had observed
replacement
of these
cards previously
and
on those
occasions
craftsmen
used
a grounding strap.
When the inspector
notified the craftsmen of this concern,
the craftsmen
stated that the
grounding strap
was not required
by procedure.
The craftsmen
completed
the
work and restored
the system satisfactorily.
The operators
successfully
completed
.the surveillance for Control Rod 18-11,
and the control
rod was
declared
The inspector
considered
the failure to use
a grounding strap while working on
sensitive electronic
components
a poor work practice.
7.2
Other Maintenance
Observations
The inspectors
also observed
JIS-95-3596
"Replacement
of Relay RFW-CIS-CNB,"
"Troubleshoot
and Repair MSLC-FT-3D," and "Replace
Scram Solenoid Pilot Valve
for Control Rod 46-15."
The inspectors
determined that these
maintenance
tasks
were performed
and
documented
properly.
8
FOLLOWUP
(92901,
92902,
92903,
92904)
The inspectors
reviewed records,
interviewed personnel,
and inspected
plant
conditions relative to licensee
actions
in response
to previous
open
items.
-18-
8. 1
Closed
Unresolved
Item 397 93-18-02:
"Reactor Vessel
Drainin
without
Secondar
Containment or Standb
Gas Treatment
S stem
"
To lower reactor
vessel
water level to the normal
range at the
end of a
refueling outage,
operators
drain the reactor vessel.
This drain path
was
through the residual
heat
removal
system which tapped off the
RPV below the
top of the active fuel.
The licensee
did not establish
secondary
containment
and the
was inoperable
at the time.
The
WNP-2
TS require that the
licensee
secure
"operations with a potential for draining the reactor
vessel
(OPDRV)" when secondary
containment
and
SGTS are inoperable.
The
inspector
was concerned
that the licensee
was not meeting
the TS.
The
licensee
stated that lowering of RPV level through the normal drain paths
was
not
an
OPDRV as long
as the system isolations
and emergency
core cooling
system actuation trips were operable.
The inspector
performed followup of the licensee's
definition of an
OPDRV and
found it to be consistent
with the industry definition and satisfactory for
safe operation.
9
LICENSEE EVENT REPORT
REVIEWS
(90712,
92700)
9. 1
In-Office Review
The inspectors
reviewed the following Licensee
Event Reports
(LERs)
based
on
in-office review.
The inspectors
considered
that the licensee
identified the
root causes
and corrective action to prevent recurrence.
LER Number
Title
95-01,
Revision
0
95-02,
Revision
0
95-04,
Revision
0
Inability to Satisfy Single failure
Criteria for Containment Isolation
Function
Due to Missing Electrical
Separation
Plate in Control
Panel
Reactor
Due to Personnel
Error
Reactor
Due to Hain Turbine Control
System Malfunction
95-05,
Revision
0
95-06,
Revisions
0 and
1
95-07,
Revision
0
TS Action Statement for Intermediate
Range
Monitoring Instrumentation
Not Met
Reactor
Scram During Surveillance Testing
Due to Protective
System
Relay Failure
Emergency
Diesel Start
Due to Voltage
on
BPA Grid
ATTACHMENT 1
1
PERSONS
CONTACTED
Washin ton Public Power
Su
1
S stem
V. Parrish,
Vice President
Nuclear Operations
- J. Burn, Engineering Director
- G. Smith, equality Assurance
Director
- P. Bemis, Regulatory
and Industry Affairs Director
- R. Webring, Support Services Director
- J. Swailes,
Plant General
Hanager
- G. Gelhaus,
WNP-2 Projects
Manager
- C. Schwarz,
Operations
Manager
- T. Love, Chemistry Manager
- J. McDonald, Technical
Services
Manager
- J. Albers, Radiation Protection
Manager
- R. Barbee,
System Engineering
Manager
- H. Manopoli, Maintenance
Manager
- G. Gelhaus,
Projects
Manager
- D. Atkinson, Reactor
and Fuels
Engineering
Manager
- D. Swank,
Licensing Manager
- W. Sawyer,
Operations
Services
Manager
- J. Huth, Plant Assessments
Manager
- D. Coleman,
Regulatory Services
Manager
- P. Inserra,
equality Assurance
Manager
G. Sanford,
Planning,
Scheduling,
Outage
Manager
- B. Hugo,
Compliance
Engineer
- H. Brant, Operations
Contractor
U.S. Nuclear
Re ulator
Commission
- H. Wong, Chief, Project
Branch
E
- R. Barr, Senior Resident
Inspector
- D. Proulx, Resident
Inspector
The inspectors
also interviewed various
CROs, shift supervisors,
shift
managers,
and maintenance,
engineering,
quality assurance,
and management
personnel.
- Attended the exit meeting
on August 4,
1995.
2
EXIT MEETING
An exit meeting
was conducted
on August 4,
1995.
During this meeting,
the
inspectors
reviewed the scope
and findings of the report.
The licensee
acknowledged
the inspectors'indings.
The licensee
did not identify that
any
proprietary information was provided to or reviewed
by the inspectors.
ATTACHMENT 2
CRO
IKC
IRB
LCO
LER
NRC
OPORV
PER
PMR
ppM
TMR
TS
WNP-2
control
rod drive
control
room operator
Final Safety Analysis Report
instrument
and controls
Incident Review Board
limiting condition for operation
licensee
event report
U.S. Nuclear Regulatory
Commission
operations with a potential for draining the reactor
vessel
problem evaluation request
plant modification record
Plant Operating
Committee
plant procedures
manual
reactor core isolation cooling
reactor pressure
vessel
shift technical
advisor
standby
gas treatment
system
temporary modification request
Technical Specifications
Nuclear Project
2