ML17278A964
| ML17278A964 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 07/24/1986 |
| From: | Rebecca Barr, Dodds R, Johnson P, Toth A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML17278A962 | List: |
| References | |
| 50-397-86-21, NUDOCS 8608130328 | |
| Download: ML17278A964 (24) | |
See also: IR 05000397/1986021
Text
U. S.
NUCLEAR REGULATORY COMMISSION
REGION U
Report No: 50-397/86-21
Docket No: 50-397
Licensee:
Washington Public Power Supply System
P.
O. Box 968
Richland,
WA. 99352
Facility Name:
Washington Nuclear Project No.
2
(WNP-2)
\\ )
Inspection at:
WNP-2 Site near. Richland,
I
Inspection
Conducted:
June l - July 12,
1986
Inspectors:
R.
T
odd
Senior Resident Inspector
Date Signed
A.
D
oth, Reactor Inspector
June
- 13,
1986
Approved by:
P.
H. J
Reacto
nson,
Chief
rojects Section
3
R.
C.
arr, Resident Inspector
~
~
Date Signed
/ Md0
Date Signed
>(~ ri
Date Signed
Summary:
Ins ection on June
1 - Jul
12
1986 (50-397/86-21)
Areas Ins ected:
Routine inspection by the resident inspectors
and
a regional
inspector of control room operations,
engineered
safety feature
(ESF) status,
surveillance program,
maintenance
program, licensee
event. reports,
special
inspection topics,
and licensee action on previous inspection findings.
During this inspection,
Inspection Procedures
30703,
39701,
42700,
61707,
61726,
62703,
71707,
71710,
71711,
72700,
90712,
92700,
92701,
92702,
92703,
and 93702 were utilized for guidance.
Results:
Two violations or deviations
were identified pertaining to radiation
controls
and tool and equipment accountability during the repair of an
system valve (paragraph 6).
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DETAILS
Persons
Contacted
D.
J.
-C.
J.
-R.
R.
K.
J.
R.
"D
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p.
M.
~D
"D
'V
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""S.
-S.
L.
Mazur, Managing Director
Martin, Assistant
Managing Director for Operations
Powers,
Plant Manager
Baker, Assistant Plant Manager
Corcoran,
Operations
Manager
Beardsley,
Assistant Operations
Manager
Cowan, Technical Manager
Harmon, Maintenance
Manager
Graybeal, Health Physics
and Chemistry Manager
Feldman, Plant equality Assurance
Manager
Peters,
Administrative Manager
Powell, Licensing Manager
Wuesterfeld,
Reactor Engineering Supervisor
Walker, Outage
Manager
Williams, BPA, Nuclear Engineer
Shockley,
HP/Chemistry
Webring, Plant Technical
Hedges,
Maintenance
McKay, Assistant Operation Manager
Davison,
Compliance Engineer
Harrold, Engineering
Manager
>Personnel in attendance
at exit meeting July 11,
1986
An interim presentation
regarding follow-up of previously identified
inspection findings was conducted
June
13,
1986 in conjunction with an
exit meeting by a regional inspector (ref.
NRC Inspection Report 86-22).
The inspectors
also interviewed various control room operators, shift
supervisors
and shift managers,
engineering,
quality assurance,
and
management
personnel relative to activities in progress
and records.
General
Supply System
Management
continued the practice of assigning
an
experienced
licensed senior operator
as the Assistant Operations
Manager
(AOM).
Mr. S.
McKay recently replaced
Mr. R.'Beardsley in that position.
The
AOM position was established
to strengthen interaction between plant
operations
and its support organizations.
Also, the
AOM provides
an
additional technical resource
to the Technical Support Center staff
during an emergency.
'
1
At the beginning of the period the reactor 'was shutdown with repairs to
uninterruptible power supply IN-l,in progress.
The reactor
was restarted
on June 4,
1986.
Training for,prospective-'licensed
operators
was
conducted
on June
4 and 5.
Start-up testing
and surveilla'nces
were
performed from-June
4 through June '20. 'n',June
10,
1986 an unplanned
L
1
V
~'
V
(
)C
reactor
scram occurred while shifting digital electro-hydraulic
(DEH)
control modes of the main turbo-generator.
On June 21,
1986 the reactor
was shut
down when MK-V-53B was determined to be leaking by the seat in
excess
of Technical Specification allowed limits.
On June 29,
1986 the
reactor
was restarted
and achieved
100'/ power on July 4,1986.
On July 10,
while performing
a ground isolation, the reactor
scrammed
due to
de-energizing
a relay that indirectly provides input to the turbine
governor valve control system.
The reactor
was restarted that evening
and
on July 12 was at 82/ thermal power.
0 erations Verifications
The resident inspectors
reviewed the control room operator
and shift
manager log books
on a daily basis during this report period.
Reviews of
the Jumper/Lifted Lead Log, the Clearance
Order Iog and the the
Nonconformance
Log were also performed to verify operations
were being
conducted per Technical Specifications
and licensing requirements.
Events involving unusual
equipment conditions were discussed
with
on-shift control room operators
and were evaluated for safety
significance.
The licensee's
adherence
to Limiting Conditions for
Operations
(LCO's) was observed.
The inspectors
routinely took note of
activated annunciators
on control panels
and ascertained
that control
room operators
were familiar with the reason
and the significance of each
active annunciator.
The inspectors
observed
access
control, manning,
nuclear instrument operability and availability of on-site
and off-site
electrical power.
The inspectors
made regular tours of accessible
areas
to evaluate
equipment conditions,'adiological
controls,, s'ecurity, safety
and adherence
to regulatory requirements.,
n
Detailed reviews of Control Room..Operators'nd
Shift Managers'ogs
and
Clearance
Order Logs were conducted to evaluate
the effectiveness
of
management
actions directed at ~improving. the quality of these
documents.
The omission of significant off-normal conditions from the'ontrol
room
logs--e.g.
frequent venting of the "B" RHR"Loop'due.,to'leakage
from
RHR-V-53B--continue to persist.
Management actions ',to; eliminate
previously noted deficiencies in'he Equ'ipment Clearance
'and Tagging
Procedure
(PPM 1.3.8) (e.g. redundant'erifications
not performed
and
pink sheets
remaining in the active log subsequent
to hanging tags)
have
been effective.
(86-06-02,'Closed).
Future, ins'pections will focus on
licensee efforts to improve the quality of the control room logs
(86-06-04).
No violations or deviations
were identified.
Surveillance
Pro
ram Im lementation
The inspectors
ascertained
that surveillance 'of,safety-related
systems
or
components
was being conducted in accordance
with license
requirements.
In addition to witnessing
and verifying daily control panel instrument
checks,
the inspectors
observed portions of several surveillance tests
by
operators
and instrument
and control technicians.
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PPM 7.4.44.1
PPM 7.4.1.4.1.1
PPM 7.4.8.21.20
PPM 7.4.3.1.1.1
T.P. 8.2.100
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Reactor Coolant 'PH, Chloride and Conductivity
Analysis
Rod Worth Minimizer Pre-Critical Check
Weekly Battery Testing
Intermediate
Range Monitor-.Ch. A-CFT
Test
Late preparation
and questionable
quality of T..P.8.2.100
(RCIC Over'speed
Test)
caused
the Shift, Manager to reject the initial proposed test.
Additionally, the initial procedure
taken to the Shift Manager
had not
been approved
by the Plant Operating
Committee.
When performed,
the test
was conducted successfully.
No violations or deviations
were
identified.'onthl
Maintenance
Observation
Portions of selected
safety-related
systems
maintenance activities were
observed."
By direct observation
and review of records
the inspectors
evaluated
maintenance activities to be consistent with LCOs and proper
administrative controls.
Maintenance
Work, Request
(MWR) AU0746,
RHR-V-53B repair,
was examined in detail.
The plant was
shutdown
on June
21,
1986 to repair RHR-V-53B which had
seat
leakage in excess of technical specification allowed limits.
Work,
document
MWR AU 0746 established
the requirements
and conditions for
conduct of the repair.
The work instructions did not refer to, as they
should have,
the use of plant procedure
PPM 1.3.18,
Tool and Equipment
Accountability Around Open Plant Systems.
PPM 1.3.18 requires stringent
tool accountability
and work practices
when working on open reactor or
Emergency
Core Cooling Systems.
These controls were not being used in
the repair of MiR-V-53B.
Prior to completing the repair,
the potential
for entry of foreign objects into the system existed particularly when no
actual maintenance
was being performed.
Only a yellow polyethylene
bag
stuffed in the valve body served
as
a boundary.
The work area
was
extremely cluttered
and disorganized.
Tools, rags,
spray
cans
and
miscellaneous
debris were strewn throughout the work area.
The globe of
an overhead light had been broken during the repair and glass
fragments
were scattered
throughout the work area.
The globe had not been replaced
prior to recommencing repairs.
The exposed light bulb represented
an
electrical shock hazard to the workers.
Lanyards
were not being used
on
tools when working over the open valve.
Controls to prevent the entry of
foreign material in the reactor
and
RHR system were not in use.
These
conditions were identified as
an apparent violation of Techni'cal
Specification 6.8.1 (86-21-01).
Radiation Work Permit
(RWP) 286-00-280 established
health physics
and
contamination control requirements for the repair of RHR-V-53B.
The
required continuous health physics
coverage
when breaching
the system
boundary
and during grinding, welding, machining
and wire brushing,
activities.
Additionally, the
RWP required use'of
a face shield or
respirator during work when the system
was open. During grinding of
RHR-V-53B, continuous health physics;,coverage
was not provided nor was
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the required protective respirator/face
shield worn as required by the
RQP.
Good contamination control work practices
also were not being used during
the maintenance.
The potentially contaminated poly bag that was stuffed
in the valve body to prevent foreign objects
from entering the system
was
frequently removed from the valve and placed
on scaffolding, lagging or
piping during maintenance.
Potentially contaminated tools used during the
maintenance
were not segregated
from those that were not contaminated.
The containers for the discarded protective clothing (PC's)
were too far
from the step off pad requiring the throwing of PC's
6 to 8 feet from the
step off pad to the container.
The container for used PC's
was overfilled
and required emptying.
These conditions were identified as
an apparent
violation of Technical Specification 6.11.
(86-21-02)
Two apparent violations were identified.
7.
En ineered Safet
Feature Verification
This inspection
was performed
as part of the team inspection
and will be
discussed
in inspection report 86-11.
8.
Start-u
Testin /Refuelin
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Control Rod Drive Scram Time Tests
On June 8,
1986,
the inspectors
examined the computer printout of
control rod scram tests that were conducted at rated pressure
and
temperature
pursua'nt to P.P.M. 7.4.1.3.2.
Data were not collected
on two rods and,the
scram times for three other rods were longer
than normal.
These control rod hydraulic systems
were vented to
remove any gas in the hydraulic system.
On June )3,
1986, the
inspector
observed
a portion of the testing of these
rods. All but
one of the five rods
(34-51) met the acceptance
criteria of 3.497
seconds
from the fully withdrawn position to position 5 for the
average
scram insertion time of all operable control rods.
Rod 34-51 had
a measured
time of 3.569 seconds,
well under the
individual scram time limit of 7 seconds.
b.
Local Power Range Monitor (LPRM) Calibration
c ~
The inspector
observed portions of the initial calibration of LPRM
panel meters
and individual power supplies.
He also observed
the
initial traversing in-core probe (TIP) configuration verification
relative to fuel position and the subsequent
full-core flux mapping
(performed at 30/ percent power) that was used to determine the
gain adjustments.
The tests
were'"conducted
pursuant to P.P.M.
7.4.3.1.1.70,
LPRM Gain Calibration and Setpoint Check.
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Reactor Recirculation
Pump "B" (RRC-P-B) Testing
The inspector
observed
the initial 60Hz,'0% reactor power testing
of RRC-P-B.
Due to excessive vibration,
RRC-P-,B had been replaced
during the refueling outage.
The maximum pump shaft vibratio'n on
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June
13,
1986 was found to be 6.3 mils at
a Slow of 10,000
gpm.
The
shaft vibration decreased
to about 5.5 mils as
pump flow was
increased
to rated flow during power ascension.
These vibration
levels were within the manufacturer's
recommendations
and,the
pump
was returned to service.
d.
Shutdown Margin Determination
The inspectors
reviewed Surveillance
Procedure 7.4.1.1.,
Shutdown
Margin Determination
(SDM), which utilizes the In Sequence
Method,
and found the procedure technically adequate.
The inspectors
also
observed
the performance of and reviewed the results of the Shutdown
Margin Determination.
The calculations
were verified accurate
and
the surveillance results
were within Technical Specification limits.
The shutdown margin,
however, differed significantly (1'/ delta k)
from the fuel vendor's
(Exxon Nuclear
Company-ENC) calculated value.
ENC, after reviewing their calculations at the licensee's
request,
determined incorrect bias values
had been used in calculating the
After using the correct bias values,
the
calculated
shutdown margin closely agreed with the actual
shutdown
margin.
e.
Procedural
Controls
The licensee in Administrative Procedure
P.P.M. 1.2.3.,Plant
Procedure
Control, establishes
the requirements for procedural
use
and compliance.
Step-by-step
performance is used only when the
procedure
being used specifically states
that sequenced
step-by-step
performance is required.
While observing
two reactor start-ups
this reporting period,
the inspectors
observed that the steps within
the procedure
were not being performed in a defined
sequence
and by
not doing so contributed to the complexity and delayed the start-up.
In one instance
the steps for opening the main steam isolation
valves
and having an operating Digital Electro-Hydraulic Control
System were skipped
and not signed off as part of the start-up
checkoff list. Subsequently,
the "verify start-up checklist" block
of the Start-up Procedure,
P.P.M.3.1.2.,
was incorrectly signed off
as completed without a qualifying or explanatory statement.
When
the reactor
reached
190 degrees F., the start-up
was delayed
due to
the inoperability of the
DEH system.
The inspector noted that the
licensee
should reevaluate
the guidance provided in P.P.M. 1.2.3 for
using plant procedures
(86-21-03).
No violations or deviations
were identified.
9.
S stem Ba'ck-I,eaka
e
Following the refueling outage, trains A'and
B of the
RHR system were
pressurizing
due to back-leakage
from the primary system;
This.
necessitated
frequent venting of each train.
Examination, of local
leakage
rate test data performed during the outage
showed 'the isolation
valves to be well within the test acceptance
criteria.
By'June
19,
1986,
train B needed to'e vented
a'bout every
8 minutes'to,preclude
actuation
of the system pressure relief valves, whil'e trainA needed to be vented
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about every hour.
Accordingly, Temporary Procedure 8.3.4.2,
RHR System
Sack-leakage
Test,
was drafted
and approved by the the Plant Operations
Committee.
The inspector
observed part of the testing conducted
on
June 20,
1986.
The tests
showed the leakage
through RHR-V-53B to be 1.3
gpm, exceeding
the Technical Specification limit, of 1.0 gpm.
The
licensee promptly commenced
a reactor
shutdown in accordance
with
Technical Specification Action Statement 3.4.3.2.c,
achieving "hot
shutdown" within ll hours
and "cold shutdown" within the next
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
Tests of the valve following shutdown
showed the leakage rate to be
equivalent to that measured
at power.
The associated
"B" Train inside
containment isolation check valve was still well within leakage
rate
limits as were the corresponding
valves in train A.
RHR-V-53B was
subsequently
disassembled
and repaired.
Following return to power
operation both train A and
B RHR containment isolation valves exhiMted
minimal back-leakage.
No violations or deviations
were identified other than those associated
with the valve repair itself (see paragraph
6).
Iicensee
Event Re orts
The resident inspectors
reviewed the following reports
and supporting
information on site to verify that licensee
management
had reviewed the
events,
corrective action had been taken,
no unreviewed safety questions
were involved and violations of regulations
or Technical Specification
conditions
had been identified.
LER-85-23
(Open) - Cable Tray Cover Omissions -
As noted in NRC
inspection report 85-37, the licensee
had deferred installation of
missing covers pending evaluation of need via testing by Wylie
Iaboratories.
Hourly fire tours of susceptible
areas
have been
implemented
as compensatory interim measures.
The Hylic Laboratories
testing has
now been
completed
and documented in a test report dated
January
30,
1986.
The licensee
stated that identification of trays with
missing covers
and determination of which do not need covers
and which
still require covers is documented in design
change
package
DCP-85-352,
revisions
OA, OB, and
OC.
The licensee
provided the inspector with a
copy of the Wylie Laboratories test report for review by NRC.
In a January
31,
1986 letter to NRC Nuclear Reactor Regulation
(G02-86-121)
the licensee
advised that
a spray-on non-flammable coating
would be used in lieu of tray covers to meet requirements
of Regulatory
Guide 1.75.
Physical work was found to be in-progress in conjunction
with corrective actions for 10 CFR 50 Appendix R discrepancies
identified
during
a recent
NRC inspection.
The licensee
stated that July 31,
1986
was the target date for completion of all corrective work relative to the
cable tray cover omissions.
This item is considered
an open issue pending
NRC review of,the test
report,
examination of the design
changes
which i.dentify,which trays will
not be covered,
and completion of corrective actions planned by July 31,
1986.
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No violations or deviations
were identified.
11.
IE Bulletin Follow-u
For the IE Bulletin listed below, the inspector verified the bulletin was
received by licensee
management,
that
a review for applicability was
performed,
and that. if the bulletin were applicable to the facility,
appropriate
corrective actions
were taken.
IB-85-02
UV Trip Attachments of DB-50 RT Breakers
(The DB-50 Westinghouse electrical breaker is not used at WNP-2.
This
breaker is frequently used in PWR's
as
a scram breaker.)
12.
IE Circulars
and Notices
For the IE Circulars/Notices listed below, the inspectors verified that
the Circular/Notices
were received,
that an applicability review was
performed,
and that if applicable to the facility, appropriate
corrective
actions
were taken.
IC-80-21
Regulation of Refueling Crews
IC-81-14
Evaluate
MSIV Programs
and Propose
Changes
to
Minimize Problems
13.
WNP-2 has developed
and implemented
an effective computerized
system for
logging and tracking the status of IE Bul'letins, Circulars
and Notices.
However, if the item is found to be applicable to the facility,
appropriate corrective actions
are frequently slow to be identified and
implemented.
Recently, all open external,",commitments
were integrated
into a
common data base to aid management in tracking open commitments.
To date, little improvement has been observ'ed,in
the closing of long
standing
open information notices
(86'-21,-04).
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Xicensee Actions
On Previous
NRC Ins ection Findin s
The inspectors
reviewed records,
interviewed'personnel,
and i.nspected
plant conditions relative to licensee
actions
on previously identified
inspection findings:
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(Closed) Follow-up Item (84-35-01) - Omission of small diameter vent
and drain valves from containment integrity checklist.
NRC inspections
(reports
84-35
and 84-37) identified vents
and
drains which did not appear
on surveillance checklists.
This
matter, subject of a subsequent
notice of violation (report 85-11,
item 85-11-03)
was resolved
as described in'eport 85-36.
A
Technical Specification
change to section 4.6.1.1.b
(Amendment No.
22) clarified that small diameter vents
and drains
need not be
secured in place
and verified on 31 day cycles.
The previously
noted omissions of such valves from the surveillance checklists is
now a closed issue.
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(Closed) Follow-up Item (84-36-01) - Potent~al'deficiency
in
cable crimp connections
of battery chargers
(identified at another
,reactor facility).
Inspection report 85-12 (Paragraph
6.'a) recognized
the licensee's
efforts to assess
the performance of electrical power connectors in
equipment throughout the plant, using infra-red detectors.
The
licensee
has recently procured
such equipment for repeated
use
and
has
conducted additional surveys
and corrective actions.
The
electrical maintenance
supervisor stated that the battery charges
had been subject to maintenance
which involved tightening of the
connections,
and no problems
have been noted with the crimp
connectors.
(Closed) Follow-up Item (85-36-02) - Minor loss of control of
weld filler material by craft's.
The January
1-May 10,
1986 Quality Assurance
surveillance report
2-86-085
documented part of the refueling outage. It included field
checks of locked and unlocked tool boxes
and associated
work areas
in the reactor plant,
and determined that uncontrolled filler
material was not being retained
by crafts.
The audits apparently
had the effect of sensitizing the crafts to management
concern in
this area
and eliminated the practice of retaining miscellaneous
weld electrodes for use in weld joint fit-ups and other non-welding
purposes.
The inspectors
noted
no instances
of uncontrolled weld
material during routine tours of work areas.
(Open) Follow-up Item (85-12-01) - Improvement In Identifying
and Correcting Drifting Setpoints of Instruments.
Inspectors
noted in reports
85-12 and 85-36 that the licensee's
calibration activities had identified setpoint drift which in some
cases
exceeded
est:ablished
tolerances,
but for which corrective
actions
appeared
confined to simply adjusting the instrument.
It
appeared
advisable to establish
methods to identify and assess,,
instrument drift prior to exceeding
tolerances
permitted by
technical specifications
thus assuring that the instruments
would
not be out of specification for prolonged periods of time between
calibrations.
It was noted that corrective measures
could include
incre'asing
the frequency of calibration beyond that required by the
technical specifications
or repair/replacement
of the instruments.,
Subsequently,
the licensee
established
a simple computer program
data base
which will print, a graph of the setpoint calibration
history of each principal instrument
and display 'technical
specificat:ion
and more conservative administrative limits.
Currently, the ISC engineer,
who is responsible for review of
completed calibration data,
enters
the current calibration data into
the computer
and identifies items which exceed
the administrative
limit.
d'or such out-of-calibration items, he obtains a,printout of
the history for that instrument
and forwards it to the Technology
department instrument engineer for'evaJuation
and action.,b-
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For the specific items previously identified by the
NRC inspectors
(HCU and CIA pressure
switches)
the printouts demonstrated
that
corrective action was taken to increase
the calibration frequency to
correct the setpoints
before technical specification limits were
exceeded.
The licensee
program included Static 0-Ring differential pressure
transmitters
used for reactor vessel level measurement.
Printouts
showed that setpoint drift had not exceeded
technical specification
tolerances,
as determined
by the calibration technique
being used,
in those
areas
examined by the inspector
(WPPSS pressure/level
transmitter calibrations involve isolation of the instrument
and
application of fluid pressure
to the actual mechanical
components of
the sensors).
The licensee
program is still under development
and all instruments
have not yet been included in the data base.
Procedures
have not yet
been established
to define responsibilities,
interfaces,
action
thresholds,
and other management
expectations
such
as corrective
action alternatives.
This licensee
program will be subject to
further NRC review as it is developed.
(Closed) Violation (85-37-02) - Vent and Drain valves were
found unlocked without record in the locked valve checklist.
The licensee
corrective action was documented in letter to NRC dated
January
10,
1986. Additionally, the locked valve checklist procedure
(PPM-1.3.29)
has been revised to clarify control of deviations.
The
specific valves in question,
(RFW-45A and 458) have been deleted
from locked valve requirements,
consistent with the resolution of
item 84-35-01
(above).
(Open) Zollow-up Item (85-37-03) -
Licensee initiatives to reduce
backlogs of nonconformance
reports.
In November
1985 the inspector noted licensee efforts to reduce
and
control backlogs of nonconformance
reports.
Additional continuation
of this licensee effort was noted in plant internal memoranda
and
memoranda
to the corporate Assistant
Managing Director for
Operations,
during the September
through December
1985 period.
Initiatives and follow-up included directions to plant technical
supervisors
expressing
management
expectations
for periodic review
of outstanding
NCRs, including review for validity and alternative
corrective actions for each
NCR. Additional guidance
was included
for separating
long term design
improvement related corrective
measures
from the interim, short term actions necessary
to correct
the immediate nonconforming condition.
The Quality Assurance
department is currently the final approval authority for close-out
of nonconformance
reports
and interview of the Quality Assurance
Manager
showed that he appeared
to be, sensitive toassuring that
NCRs are not prematurely or improper'ly 'closed
.He had limited NCR
closure review and approval authority to only a few of his'enior
level staff members,
and had monitored their activities through
personal
and corporate office audits...Various plant management
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supervisors
stated that
a concerted effort had been
made prior to
the recent refueling outage to identify NCRs which should be
included for corrective action during the outage.
The inspector
examined the current Plant Tracking Log (PTL) for all
open
NCRs originating from time of issuance
of the operating license
through mid 1985
((283-0167,, through 285-372).
Of these
84 NCRs,
20
were examined to determine
the appropriateness
of their, current lack
of resolution.
Most of these
appeared
to-be of low:significance or
had other elements
which appea'red 'to'justify, deferral
due to
prioritization of resources.
However<" three of these
unresolved
NCRs appeared
overdue without good
cau'se
and reflected
need for
continued
management
emphasis, on ba'cklogs:
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285-020 QCII tubing installed in a QCI in'strument rack
(Originated January
15,
1985') ~,"'-'
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285-228 Improper installation of anchor bolts in spray pond header
drain piping supports
(Originated May 14,
1985)
5
285-267 Failure to obtain/record specific quarterly surveillance
information to meet,
ASME in-service te'st, requirements
(Originated April 22,
1985)
The inspector did not identify any long-standing unresolved
with significant safety implications.
However, the inability to
effect timely resolution of issues
appears
to be an indicator of
adequacy of technical support resources.
This aspect will be
considered
during future
NRC inspections
and the issue of backlogs
is considered
a continued
open item.
(Open) Follow-up Item (85-38-02) - Absence of management policy
regarding long-standing activated annunciators.
The Operations
Department
implemented
a procedure requiring
clearance
order tagging of malfunctioning annunciators.
The
temporary procedure
change
system permits
changing annunciator
response
procedures
to clarify operator actions for malfunctioning
According to the Operations
Manager, Shift Managers
should utilize this procedure
(e.g. specify frequency of repeated
verification of alternative information to assure that spuriously
activated annunciators
do not mask actual
abnormal conditions)
when
repeatedly malfunction.
This policy is not currently
documented
and such actions
have been
at, the discretion of
individual Shift Managers.
This will remain an open item until licensee's
management
expectations
are
documented
or sufficient inspection data are
compiled to support effective policy implementation.
(Open) Follow-up Item (85-38-04) - Lifted leads,
authorized
under
a
long suspended
work request,
were found on the spray pond level
transmitter barometric
compensators.
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When this finding was identified, the licensee
committed to an audit
of open,
long-standing
work requests
to determine if similar
jumpers or lifted leads existed.
During the current inspection,
the
licensee
maintenance
manager
said such
a review had been
conducted
but was not documented.
He also stated
a review was conducted of
the Jumper/Lifted Lead Logs and all unnecessary
jumpers
and lifted
were removed.
To improve future controls in this area
a
revision of the Jumper
8 Iifted I,ead administrative procedure
was
submitted to the Plant Operations
Committee for consideration.
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(Closed) Follow-up Item (86-04-02) - Chemistry Department
technicians failed to verify or record that they verified use of the
current revision.
The licensee
issued
nonconformance
reports
286-021
and
022 to
prescribe corrective action.
The
January
24,
1986 memorandum to
chemistry staff reiterated
the importance of surveillance procedure
verification prior to use.
This action is considered
appropriate to
alert technicians
and data reviewers to the requirement.
j.
(Open) Violation (86-06-01) - Improperly stored high pressure
nitrogen bottles
(May 23, 1986).
Licensee corrective actions
included:
1)
revising P.P.M. 1.3.19,
Housekeeping,
to clarify gas cylinder
storage
requirements
2)
committing to training plant and contractor personnel
on gas
cylinder storage
requirements
(not yet implemented)
During a tour, the inspector noted six cylinders in the reactor
building not secured at the base,
as prescribed
by the newly issued
procedure.
When brought to a
maintenance
supervisor's
attention he
took immediate corrective action.
This discrepancy is an element of
general
housekeeping
and is addressed
in more detail in inspection
report 86-22.
This item remains
open pending completion of the
licensee'raining definition and implementation
and reinspection of
plant conditions.
12.
Mana ement Meetin
During this time the inspectors
met periodically with plant management
and discussed
inspection findings.
On July 11,
1986 the inspectors
met
with the Plant Manager
and members of his staff,
as noted in paragraph
1,
to discuss
inspection f'indings.
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