ML17305A852
| ML17305A852 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 05/21/1990 |
| From: | Wong H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML17305A848 | List: |
| References | |
| 50-528-90-12, 50-529-90-12, 50-530-90-12, NUDOCS 9006050134 | |
| Download: ML17305A852 (40) | |
See also: IR 05000528/1990012
Text
U.
S.
NUCLEAR REGULATORY COMMISSION
Re ort Nos.
Docket Nos.
License
Nos.
Licensee:
Faci lit
Name:
REGION V
50-528/90-12,
50-529/90-12
and 50-530/90-12
50-528,
50-529,
50-530
Arizona Public Service
Company
P.
0.
Box 52034
Phoenix,
AZ. 85072-2034
Palo Verde Nuclear Generating Station Units
1, 28
3
Ins ection Conducted:
March 4 through April 14,
1990
Inspectors:
Approved By:
D.
Coe,
Senior
Resident
Inspector
C. Myers,
Resident
Inspector
(Rancho
Seco)
P. Quails,
Resident
Inspector
(Rancho
Seco)
F.
Ringwald,
Resident
Inspector
J.
Sloan,
Resident
Inspector
5
I Po
ong,
se
Reactor Projects
Branch,
Section II
Ins ection
Summar
a
e
sgne
Ins ection
on March 4 throu
h A ri 1
14
1990
{Re ort Numbers
an
Areas Ins ected;
Routine, onsite,
regular
and backshift inspection
by
e
ree
res> ent inspectors,
plus two inspectors
from the Region
V
staff.
Areas inspected
included: previously identified items;
review of
plant activities; engineered
safety feature
system walkdowns;. monthly
surveillance testing;
monthly plant maintenance;
(RCS) leak during valve repacking-Unit 1; poor freeze seal
work
implementation-Unit 1; containment pre-closeout
inspection-Unit 1; eating
and smoking debris inside the radiological control area
(RCA)-Unit 1;
deenergizing
four channels
of log safety excore nuclear
instrumentation-Unit
1; fives on the instrumentation
and control (IEC)
shop roof-Unit 1; snubber visual inspections
performed without an
available
procedure-Unit
2; cleanliness
zone III controls-Unit 2; plant
housekeepin~-Units
1 and 3; stuck shut essential
ventilation exhaust
damper for
D" battery-Unit 3; auxiliary spray pressurizer
spray nozzle
usage factor-Unit 3; design characteristics
(AFM)
motor operated
gate valves-Units 1,
2 and 3; reactor trip due to dvopped
control element
assembly
(CEA)-Unit 3; review of licensee
event
reports-Units
1,
2 and 3; and review of periodic and special
reports-Units
1,
2 and 3.
During this inspection the following Inspection
Procedures
were utilized:
30702,
30703,
37828,
41400,
61726,
62703,
71707,
71710,
90712,
90713,
92703
and 93702.
Results:
Of the
22 areas
inspected,
3 violations were identified and are
5e>ng cited.
The violations pertain to Fire Protection
Program
implementation,
compliance with Technical Specification Action Statement
requirements,
and adequacy of engineering analysis.
The two non-cited
violations involve performance of a work order without all appropriate
sign-offs completed
and control of material exclusion boundaries
inside
containment
near the open reactor
vessel.
General
Conclusions
and
S ecific Findin
s
Si nificant Safet
Matters:
None
Summar
of Violations:
3 Cited and
2 Non-Cited
Violations
Summar
of Deviations:
0 en Items
Summar
None
4 items closed,
1 item left open,
and
1 new item opened.
DETAILS
Persons
Contacted:
The below listed technical
and supervisory personnel
were
among
. those contacted:
Arizona Public Service
Com an
(APS)
"R. Adney,
J. Allen,
"R. Badsgard,
J. Bailey,
B. Ballard,
F.
Buckingham,
- H. Bieling,
- T. Bradish,
P. Brandjes,
P. Caudill,
- D. Crozier,
K. Davis,
"W. Conway,
D. Fasnacht,
E. Firth,
- S. Gross,
"D. Heinicke,
"P.
Hughes,
"W. Ide,
F. Larkin,
"J.
Levine,
~J.
LoCicero,
W. Marsh,
D. Mauldin,
"G. Overbeck,
C.
Rogers,
- C. Russo,
G. Shell,
W. Simko,
"E. Simpson,
"RE Snell,
- G. Sowers,
- D. Stover,
"C. Teeter,
S. Terrigino,
P. Wiley,
R. Younger,
Plant Manager,
Unit 3
Engineering
L Construction Director
Engineering
and Construction Supervisor
Vice President,
Nuclear Safety
8 Licensing
equality Assurance
Director
Operations
Manager,
Unit 2
Emergency Plan/Fire Prevention
Manager
Compliance
Manager
Central
Maintenance
Manager
Site Services Director
Fire Department Supervisor
Human Resources
Manager
Executive Vice President - Nuclear
Nuclear-Construction
Manager
Nuclear Training General
Manager
El Paso Electric Engineer
Plant Manager,
Unit 2
Radiation Protection/Chemistry
Manager
Plant Manager, Unit 1
Security Manager
Vice President,
Nuclear
Power Production
Independent
Safety Engineering
Manager
Plant Operations
8 Maintenance Director
Outage
Planning
8 Management
Manager
Technical
Support Director
Licensing Manager
equality Control Manager
equality
Systems
Manager
Maintenance
Manager,
Unit 2
Vice-President of Engineering
8 Construction
Chairman Arizona Public Service
Engineering Evaluations
Manager
Nuclear Safety Manager
Senior Information Coord.
Management
Services
Management
Services
Supervisor
Work Control Manager,
Unit 2
Plant Standards
and Control Manager
.The inspectors
also talked with other licensee
and contractor
personnel
during the course of the inspection.
"Attended the Exit meeting held with NRC Resident
Inspectors
on
April 19,
1990.
2.
Previousl
Identified Items - Units 1
2
and
3 (92701
92702
an
a.
b.
C.
(0 en) Enforcement
Item (528/89-36-01):
"Inadvertent
ra> nln
o
e
en
ue
oo
-
ni
This item involved a failure to follow instructions in that
sufficient valves were not verified to be appropriately
~
~
~
ositioned to encompass
the desired flow path to that intended.
his subsequently
resulted in an unintended
loss of level in
the Spent
Fuel
Pool to below the Technical Specification
minimum required level.
The long term corrective action was to
rewrite procedure
"Fuel
Pool Cooling and Cleanup"
to separate
out the various spent fuel pool evolutions
by
January
10,
1990.
The rewrite has not yet been
complete but is
being tracked
by the Commitment Action Tracking System with a
requested
due date extension to September
4, 1990.
This
extension
was requested partly due to an increase
in work
scope.
The inspector concluded that the expanded
work scope
and extension of time to complete the revision appear
appropriate.
This item will remain
open until the revision is
complete.
(Closed)
Enforcement
Item (528/89-50-01):
"Work Order
e
s
ot
1
ne
s
e
or
as
er orme
-
nit l.
This item involved work being performed without the work steps
being signed off concurrently
and the Quality Control
(QC)
Inspector not addressing this issue.
The licensee
stopped
the
work, amended
the work order to repeat
the work and properly
document it, inspected
the work performed finding no
discrepancies,
disciplined the workers
and
QC Inspector,
briefed maintenance
personnel
on work step documentation,
interviewed
QC Inspectors,
reviewed
and enhanced
QC Inspector
training,
and reemphasized
the importance of procedural
adherence
to plant workers.
All corrective action is complete.
This item is closed.
(Closed)
Enforcement
Item (528/89-50-02):
"Erroneous
ecsa
uc ear
a erma
rans er
orm
-
ni
1.
This item involved a transcriptional error on the Special
Nuclear Material Transfer
Form for reloading the Unit 1 core
which resulted in placing
a fuel assembly in an unanalyzed
core
location.
The licensee
discovered
the error,
stopped
the fuel
movement,
reviewed
and corrected the transfer
form, moved the
incorrect assembly to a storage
location and continued
refueling without informing plant management.
The licensee
subsequently=revised
procedure
72AC-9NF01, "Control of SNM
Transfer
and Inventory" to require 100'ndependent
verification of the completed transfer
form and clarify the
approvals
required.
The licensee
also analyzed the incorrect
fuel assembly
placement
and determined that adequate
shutdown
margin existed,
The Unit 1 Plant Manager discussed
the
incident with Unit 1 managers
to emphasize
the need to promptly
advise senior management
of significant events.
This item is
closed.
(Closed)
Enforcement
Item (529/89-43-02):
"Procedure
Ste
s
ne
s
ou
e)n
er orme
-
ns
This item involved a procedure
step which was signed off but
not actually performed during
a Reactor Startup.
As discussed
in the December
21,
1989 reply to the Notice of Violation,
several
administrative actions
and crew briefings occur red,
and
procedure
"Conduct of Shift Operations,"
was
revised to reinforce the management policy that procedures
are
to be performed
as written.
The inspector
concluded that these
corrective actions
appear
adequate.
This item is closed.
(Closed)
Enforcement
Item (529/89-36-01):
"Excess
ow
ec
a ves
nl
or
mer enc
lese
enerators
This item related to an apparent failure to incorporate
conclusions
from a design calculation into applicable
procedures,
which resulted in the
(EDG) air start system not being aligned
as required to perform
its design function under all conditions
assumed
in the Final
Safety Analysis Report
(FSAR).
Design calculation
13-MC-ZZ-704
and Engineering Evaluation
Request
(EER) 86-XM-046 had
determ)ned that the
XCV isolation valves
needed to be closed to
meet these
design criteria.
The licensee's
immediate
corrective action was to align the
XCV isolation valves per
86-XM-046.
Further review of the circumstances
by the licensee
determined
that the results of the calculation
and additional information
derived from field testing
had been incorporated into
appropriate
procedures,
but that
a misinterpretation of the
design criteria had occurred.
The licensee's
engineering staff
incorrectly assumed that the system's
two air receivers
together,
rather than each,
were to have sufficient capacity to
start the
EDG five times.
This error resulted in erroneous
conclusions
being incorporated into the procedures.
A major effort which should identify such deficiencies
in all
safety
systems
is the design basis reconstitution.
To prevent
recurrence,
all Nuclear Engineering
Departments
and Engineering
Evaluations
Department
engineers
and supervisors
have
been
instructed via memoranda of the necessity of ensuring design
criteria are fully understood
in dispositioning applicable
documents
and that disposition information is correctly
incorporated within applicable
implementing documents.
The inspector
concluded that licensee
actions
were
satisfactory.
This item is closed.
3.
Review of Plant Activities (71707
and 93702)
Unit 1
Unit 1 remained in Mode
5 throughout this inspection period.
Activities related to completing the refueling outage
included
the reactor coolant system fill, vent and pressurization
to
approximately
150 psig,
and control element
assembly testing.
b.
Unit 2
C.
Unit 2 entered
the reporting period in Mode
5 in the midst of a
refueling outage.
Mode
6 was entered
on March 11, 1990,
and
the reactor
was defueled
on March 22,
1990.
The plant remained
in a defueled condition for the remainder of the reporting
period.
Unit 3
Unit 3 began this report period operating at 98 percent
power.
Upon resolution of a difference
between
Cycle
1 and Cycle 2
electrical output and completion of Moderator Temperature
Coefficient (MTC) testing,
Unit 3 was brought to 100 percent
power,
where it remained until a reactor trip event occurred
on
April 14, 1990,
due to a dropped Control Element Assembly
(CEA)
during monthly
CEA testing.
Unit 3 was in Mode
3 at the end of
this report period.
d.
Plant Tours
The following plant areas
at Units 1,
2 and
3 were toured by
the inspectors
during the inspection:
Auxiliary Building
Containment Building
Control
Complex Building
Diesel Generator Building
Radwaste
Building
Technical
Support Center
Turbine Building
Yard Area and Perimeter
The following areas
were observed
during the tours:
1.
0 eratin
Lo s and Records - Records
were reviewed against
ec naca
peel
ica sons
and administrative control
procedure
requirements.
2.
Monitorin
Instrumentation - Process
instruments
were
o serve
or corre
a ion
etween
channels
and for
conformance with Technical Specifications
requirements.
Shift Staffin
Control
room and shift staffing were
o serve
or conformance with 10 CFR Part 50.54. (k),
Technical Specifications,
and .administrative
procedures.
E ui ment Lineu
s - Various valves
and electrical
breakers
were veri ie
o be in the position or condition required
by Technical Specifications
and administrative
procedures
for, the applicable plant mode.
E ui ment Ta
in
- Selected
equipment, for which tagging
reques
s
a
een initiated,
was observed to verify that
tags
were in place
and the equipment
was in the condition
specified.
General
Plant
E ui ment Conditions - Plant equipment
was
o serve
or in ica ions
o
sys
em leakage,
improper
lubrication, or other conditions that would prevent the
systems
from fulfillingtheir functional requirements.
Fire Protection - Fire fighting equipment
and controls
f
fttf tf
Specifications
and administrative procedures.
Plant Chemistr
- Chemical analysis results
were reviewed
or con ormance with Technical Specifications
and
administrative control procedures.
Securit
- Activities observed for conformance with
regu
a ory requirements,
implementation of the site
security plan,
and administrative procedures
included
vehicle
and personnel
access,
and protected
and vital area
integrity.
Plant Housekee
in
- Plant conditions
and
ma eria
equipmen
storage
were observed to determine
the
general
state of cleanliness
and housekeeping.
Housekeeping
in areas
where maintenance
was in progress,
but where work was not being performed
when toured by the
inspectors
was observed to be poor.
Examples
included
disorderly material
around the Unit 2 "B" diesel
generator
and tools and pipe caps
around the "A" auxiliary feedwater
pump (See
Paragraph
15).
Radiation Protection Controls - Areas observed
included
con ro
poin
opera ion, records of licensee's
surveys
within the radiological controlled areas,
postinq of
radiation
and high radiation areas,
compliance with
Radiation
Exposure
Permits,
personnel
monitoring devices
being properly worn,
and personnel frisking practices.
Shift Turnover - The inspectors
observed shift turnover in
d tdf
t f
manner.
The on-coming shift operators
each walked
down
the plant Control
Room panels
then walked down the panels
as
a group with the on-shift crew answering questions.
On-coming shift personnel
properly reviewed the logs.
During the shift turnover process
formal Control
Room
decorum
was maintained.
The inspectors
interviewed reactor operators
and auxiliary
operators
in all three units.
The operators
and plant staff
were knowledgeable
about plant conditions
and ongoing
maintenance
and activities in their respective
units.
The inspectors
reviewed control
room operator
response
in
Unit 3 to the steam generator
master controller failing high on
March 3, 1990.
The operators
shifted control to manual
and
maintained level control to mitigate the plant transient.
The
inspectors
observed that the licensee
stationed
a licensed
operator at the feedwater control console to maintain proper
level while the electronic problem was corrected.
The
inspector questioned
the operator
on proper placement of the
manual-auto
selection in the event of a reactor trip.
Operator
response
and knowledge appeared
adequate.
No violations of NRC requirements
or deviations
were identified.
4.
En ineered Safet
Feature
S stem Walkdowns - Units 1
2 and
3
Selected
engineered
safety feature
systems
(and systems
important to
safety)
were walked down by the inspector to confirm that the
systems
were aligned in accordance
with plant procedures.
During this inspection period the inspectors
walked
down accessible
portions of the following systems.
Unit 1
o
DC Batteries
and Buses
Unit 2
o
DC Batteries
and Buses
Unit 3
o
DC Batteries
and Buses
No violations of NRC requirements
or deviations
were identified.
5.
Monthl
Surveillance
Testin
- Units j.
2 and
3 (61726)
a.
Selected
surveillance tests
required to be performed
by the
Technical Specifications
(T/S) were reviewed
on
a sampling
basis to verify that:
1) the surveillance tests
were correctly
included
on the facility schedule;
2)
a technically adequate
procedure
existed for performance
of the surveillance tests;
3)
the surveillance tests
had been performed at the frequency
specified in the T/S;
and 4) test results satisfied
acceptance
criteria or were properly dispositioned.
b.
Specifically, portions of the followinq surveillances
were
observed
by the inspector during this inspection period:
Unit 1
Procedure
Descri tion
Unit 2
Engineered
Safety Features
Actuation System
Train "A" Subgroup
Relay 62 Dry Cold Shutdown
Functional Test
Engineered
Safety Features
Subgroup
Relay
Time Response
Test
Section XI Valve Stroke Timing and Position
Indication Verification
Mode 1 thru 4 Steam
Generator
No.
2 Containment Isolation Valves
Excore Startup
Channel
Functional
Test
Section
XI Check Valve Operability and
Position Indication Verification - Mode 1 thr u 4
HP "B" Train
P
d
~ll
gati
Snubber Visual Examination
Diesel Generator
"A" Test
The inspector
reviewed the results of snubber testing in
Unit 2, conducted to meet the requirements
of Technical
Specification (T/S) 3/4.7.9.
Procedure
73ST-9ZZ22 was
used
by
the licensee to test
39 snubbers.
One snubber,
failed.
The licensee
then performed the action statement
required
by T/S 3/4.7.9
and tested
an additional
19 snubbers.
All additional
were satisfactory
and the engineering
evaluation for system operability with the failed snubber
was
satisfactory.
Unit 3
Procedure
Descri tion
CEA Operability Check
ADV 185 Stroke Test
No violations of NRC requirements
or deviations
were
identified.
H
6.
Monthl
Plant Maintenance - Units 1
2 and
3 (62703)
a.
b.
During the inspection period, the inspector
observed
and
reviewed selected
documentation
associated
with maintenance
and
problem investigation activities listed below to verify
compliance with regulatory requirements,
compliance with
administrative
and maintenance
procedures,
required
equality
Assurance/equality
Control involvement, proper
use of safety
tags,
proper equipment alignment and
use of jumpers,
personnel
qualifications,
and proper retesting.
The inspector verified
that reportability for these activities was correct.
Specifically, the inspector witnessed portions of the following
maintenance activities:
Unit 1
Descri tion
Pre-welding examination fit-up planning
SGA-UV-174
As-built measurements
on SG-PSV-555
Potter-Brumfield washer,
lock washer,
and screw
replacement
"A'rain Containment
Hydrogen Monitor Cell Replacement
Auxiliary Feedwater Terry Turbine Overspeed Trip
Adjustment using Auxiliary Steam
The inspector identified a Foreign Material Exclusion
(FME)
cover missing from an open instrumentation line on SGA-UV-174,
the steam generator
number
one economizer
upstream
isolation valve.
Step 4.2 of the Work Order
(WO) required placing an
FME cover
on the line and it had been
signed off on the
WO.
A mechanic
was evaluating
upcoming work on this valve and placed
an
cover on the open line after the inspector brought it to his
attention.
Unit .2
Descri tion
4160 Volt Circuit Breaker Maintenance
Core Offload
Essential
Spray
Pump "B" Discharge
Check Valve Inspection
Modification of High Radiation Area Access
Control
Door
"B" Emergency
Diesel Generator Air Intercooler
Refurbishment
Potter Brumfield Relay Retests
During observation of Unit 2 defueling activities from the
refuelinq machine,
the inspector
noted that
a grapple limit
switch dsd not consistently function to activate indication of
grapple
engagement.
However, operators visually verified
grapple position by confirming the position of the manual
operator
and the inspector
concluded that this provided
adequate
assurance
of grapple position.
The inspectors
observed
maintenance
on the Unit 2 emergency
diesel
generator
on Harch 8, 1990, in which mechanical
maintenance
personnel
were repairing
a minor fuel oil leak on
an injector.
When loosening
a set screw,
the screw fell out
and fuel oil flowed to the floor of the room.
The on-shift
personnel
took proper actions to limit the oil spill to an area
of about four square feet and they properly cleaned
up the
s pi 1 l.
The inspector
observed
performance of maintenance
procedure
32NT-9Z234 on
a spare
4160 Volt circuit breaker.
The inspector
verified that properly calibrated test equipment
was
used, that
the electricians
had
a current and properly approved procedure
and that they were following the procedure
in a step-by-step
fashion.
The inspector
observed
the refurbishment of the "B" diesel
generator
"A" air intercooler.
Portions of the cooler
.including the
end plate seating surfaces,
had been coated with
a ceramic material.
Because this materia'I
was applied unevenly
on what is required to be
a machined surface,
substantial
rework was required.
Additionally, the licensee
is
reevaluating
the
use of this material
and the method of
application,
based
on this less
than satisfactory
experience.
Unit 3
Oescri tion
o
Valve RC-207 Packing
Leak
o
MSIV 4-way Valve Replacement
The inspector
reviewed the licensee's
analysis to,install
two
oversized
packing gland nuts
and spacers
over the existing nuts
on a 3/4 inch Borg Marner globe valve used
as
an instrument
isolation valve on the pressurizer.
The licensee
had
previously used injected material to stop
a packing leak, but
had not been successful
and had further determined that one
existing gland nut was galled and could not be taken
down any
further.
The licensee
chose the oversized
gland nut and spacer
combination
as
a means to further compress. the packing and
injected material.
Prior to performing this work,
NRC inspectors
questioned
the
adequacy of the
MNCR 90-RC-0009 Conditional
Release
engineering
analysis
on two points.
The first was the lack of an analysis
of the stress
on the packing gland retainer
due to the spacers.
Subsequently,
the licensee
determined that the retainer could
withstand this stress.
Second
was the lack of a stress
analysis
on the pin which secures
the packing gland stud to the
10
valve body.
In this case,
the licensee
was preparing to use
a
planned
nut torque of 75 ft-lbs which would have
exceeded
the
allowable stress
on the pin.
Following the inspector's
questions,
the licensee
modified the work order to reduce the
maximum allowed torque value to 30 ft-lbs, which was consistent
with the manufacturer's
recommendations.
The use of the
75
ft-lbs value
was based
on apparent
misunderstanding
of the
analysis
which appeared
to allow the
use of this
maximum value.
The inspector concluded that the licensee
was preparing to
proceed with this work order
(WO 00419787) without an adequate
technical
basis for the use of 75 ft-lbs of torque.
The
MNCR
analysis
which appeared
to authorize the use of 75 ft-lbs of
was approved
by Engineering,
Operations,
and equality
Control.
Following performance of the work order,
the packing
leak was significantly reduced.
The failure of the
MNCR
analysis to specify appropriate
acceptance
criteria for the
maximum torque allowed appears
to be
a violation of NRC
requirements
(530/90-12-01).
On March 20,
1990, the inspector
observed
a portion of a
corrective maintenance
Work Order
on Main Steam
Isolation Valve (MSIV)-180 in Unit 3 and noted that the
WO had
not been
signed
by the Shift Supervisor
(SS).
Thus the
signature certification for approval for the work and
verification of redundant train operability was not present
as
was required
by the licensee's
work planning and implementation
procedures.
When the inspector questioned this, the
maintenance
foreman
on the job immediately stopped all work and
returned to the Control
Room and obtained the
SS signature.
The inspector
determined that the
had conducted
a pre-job
briefing with the maintenance
personnel,
had authorized
the
work to be done,
and
had verified redundant train operability.
Both the
and the foreman accepted partial responsibility for
failing to ensure
the authorization signatures
were properly
affixed.
Licensee
management
counseled
the involved
individuals and communicated
a caution to all other foreman
and
Shift Supervisors.
The inspector
noted that although the
omission of the
SS signature
was
a violation of licensee
procedures
required
by Technical Specifications,
the safety
significance of the omission
was negligible and licensee's
corrective action was prompt.
Thus, the violation is not being
cited because
the criteria specified in Section
V.A. of the
Enforcement Policy of 10 CFR Part 2, Appendix C, were
satisfied.
7.
S stem
(RCS)
Leak Durin
Valve
Re ackin
- Unit I (62703)
On March 29,
1990,
mechanics
were repacking valve PSI-AV-458, which
crossties
the shutdown cooling system
and the fuel pool cooling
system, with the valve presumably
closed.
During the work the last
two rings of packing were forcibly ejected
from the valve by
shutdown cooling system pressure.
The mechanics
attempted
but were
unable to reinsert the packing while leakage
was approximately
11
30
gpm.
Pressurizer
level decreased
from 55 to 47 percent since
letdown and charging were secured at the time.
After calling the
control
room, radiological protection,
and their supervisor,
the
mechanics left the area
and were subsequently
found to not be
contaminated.
Several
supervisory individuals responded
to the
event
and reduced the leak rate to 15
gpm by removing the mechanical
stop
on the remote actuator
and shutting the valve more fully.
The
"B" Train shutdown cooling loop was then warmed
up and placed into
service in accordance
with the operating procedure.
The shift
supervisor
allowed the reduced
leak to continue while warming up the
"B'rain based
on the fact that pressurizer
level
was being
controlled with charging
and letdown returned service,
and the water
leaking from the valve was draining to floor drains without
significantly worsening the existing radiological conditions.
After approximately
1>> hours
the "B" shutdown cooling loop was
placed in service
and the "A'oop was isolated
and drained.
Since
the "B" shutdown cooling loop was available,
the decision to attempt
repacking valve SI-UV-458 with 150 psig system pressure
against it
appeared
to the
NRC inspector to represent
poor maintenance
planning.
The licensee
responded
by restricting the repacking of
valves which could not be isolated
and drained, to valves which are
only exposed to static
head pressure.
The licensee further committed to evaluate
and promulgate the
criteria for and authority required for repacking unisolated
valves
at pressures
above
a static
head.
The inspector
concluded that the
work schedulinq for this activity was poor in that it failed to
consider minimizing core cooling vulnerabilities, worker safety,
the
potential for spreading radiological contamination,
and redundant
train availability.
This event
was briefly discussed
in a meeting with APS and Region
V
management
personnel
on March 30, 1990, which followed an
enforcement
conference
of the
same
day related to radiation
protection concerns.
No violations of NRC requirements
or deviations
were identified.
8.
Poor Freeze
Seal
Work Im lementation - Unit 1 (62703)
The inspector
reviewed the work package for placing
a freeze
seal
on
the piping between
the spray chemical addition tank and valve
SIA-UV-603 and observed
the freeze
seal while it was in progress.
Procedure
33MT-9ZZ02, "Freeze Sealing",
contains
Appendix E, Freeze
Seal
Safety Evaluation sheet,
which had been prepared
by the planner
on June
10,
1988.
The inspector
questioned
three aspects
of the
planning documented
in Appendix E.
Section l)C) of Appendix
E
identifies the
maximum pressure
expected
across
the freeze
seal
as
20 psig and Section 1)D) lists relief SI-PSV-250
as the upstream
relief.
On March 19,
1990, while the freeze
seal
was in place, this
valve was tested
and found to relieve at 30 psig rather than
15 psig
as intended.
The Spray Chemical Addition Tank was at atmospheric
12
pressure
at the time so SI-PSV-250
had
no actual
impact on the
freeze seal.
Section l)C) lists measures
to ensure
the
maximum differential
pressure
across
the freeze
seal
is not exceeded.
One of these
measures
listed is to "Monitor SI-PI-336 periodically."
SI-PI-336
is a remote reading pressure
transmitter
which is only indicated in
the Control
Room.
The inspector questioned
the Control
Room
operators
whether they were, or had been,
asked to monitor SI-PI-336
periodically during the freeze sealing operation
and they responded
that they were not specifically monitoring SI-PI-336 and
had
received
no requests
to do so.
Section 4)0) identifies radiological considerations
in the job
planning process.
The planner indicated "Non-Radioactive"
even
though the Spray Chemical Addition system connects directly to the
suction of the containment
spray
pumps which are part of the
shutdown cooling loop and connected to the refueling water tank,
both of which -are contaminated
systems
and sometimes
require hot
particle controls.
The inspector
determined that the spray chemical
addition tank is sampled for hydrazine but not activity.
However,
the inspector
noted that the sample
and container
are handled
as
potentially contaminated:
When the inspector
reviewed the package after the freeze
seal
was in
place,
the Mork Group Supervisor
(WGS) review signature
was noted to
not be present
on the
WCSE checklist.
When the inspector questioned
the
WGS, the
MGS reviewed
and signed the checklist and explained
that this checklist
was
no longer used
and that
he did not notice
the blank when
he reviewed
and signed the
WO.
The
MO did not contain contingency plans in the event the freeze
plugs were to thaw or otherwise fail to seal.
The inspector
acknowledged that the
MCSE checklist contains
some of the
information which would be included in contingency planning.
The
inspector
concluded that there
was not complete contingency planning
for this job.
The licensee
was planning
on revising 33MT-9ZZ02, "Freeze Sealing,"
when this job was being performed.
After the inspector questioned
these
various issues
with the
WGS, the
MGS and his Supervisor
met
with the inspector to discuss
a draft of the proposed
changes
to the
freeze sealing procedure.
Several of these
issues
were already
addressed
in the draft primarily by the addition of a Freeze
Seal
Coordinator
who will be responsible
for coordinating the diverse,
inter-department
responsibilities
among the various work groups.
The draft also included
an expansion of the
WCSE check list to
include detailed contingency plans for the failure of the seal
and
a
loss of ventilation in the freeze
seal
area.
The inspector
concluded that the planned
changes
appear to be improvements
which
address
these
concerns.
The inspector
had
no further questions.
No violations of NRC requirements
or deviations
were identified.
Containment
Pre-Closeout
Ins ection - Unit 1 (71707).
During licensee
preparation
for closeout of the Unit 1 containment,
the inspector
noted several
deficiencies
not previously identified
or included
on the "Containment Closeout
Punch List."
The
deficiencies
included improper valve locking devices,
an unconnected
grounding cable for a safety-related
conduit,
and several
examples
of deteriorated
"Seal-Tite" sheathing
on Anaconda flexible conduit
for safety-related
and non-safety-re1ateg
components
(none with
Environmental gua1ification),
several
burned-out lights,
and
apparent boric acid accumulation
on various piping.
The licensee
promptly addressed
the deficiencies
and repaired the conduit.
No violations of NRC requirements
or deviations
were identified.
Eatin
and'Smokin
Debris Inside The Radiolo ical Control Area
n1
On three separate
occassions,
the inspector
observed
indications of
possible eating or smoking on the Auxiliary Building roof,
a
radiological control area.
Smoking or eating in thss
area is
prohibited by the licensee's
procedures.
The licensee
issued
a
memo
to all plant radiation workers after the first observation
indicating that eating
and smoking materials
could
no longer be
brought into the
RCA.
Approximately two weeks later, the inspector
noticed additional debris
on the Auxiliary Building roof.
The
licensee
made
a more thorough inspection
and found more debris.
Two
days later, the inspector noticed another indication of eating
on
the Auxiliary Building roof.
A prior example of food debris
and
cigarettes
in the
RCA was reported in inspection report
50-530/88-39.
The licensee
responded that the
new policy required
time to be fully effective.
Mhile it was too soon to evaluate
the
effectiveness
of the
new policy, the licensee
committed to make more
frequent observations
for at lease
one month in the areas
where
evidence of eating
or smoking in the
RCA have
been
noted in the
past.
No violations of NRC requirements
or deviations
were identified.
Deener izin
Four Channels of Lo arithmic Power Excore Neutron
ux
ns
rumen
a son -
ns
On March 21,
1990, with Unit 1 in Mode 5 while preparing to repair
the cabinet slides
and
make other hardware
and circuit repairs to
the
Log Power
Excore Neutron Flux Instrumentation
and Plant
Protection
System
(PPS) cabinets
and drawers, all four channels
of
Log Power Excore Neutron Flux Instrumentation
were deenergized.
The
cabinet slide work was to be done by a vendor who had
a work policy
of not working on energized
equipment.
The Technical Specification
Limiting Condition for Operation
(LCO) 3.3-1 and associated
Table
3.3-1 require
two channels
of Log Power Instrumentation
in Mode
5
and specifies
actions with only one channel
The
NRC
inspector
concluded that a violation of TSLCO 3.3-1 occurred in that
TSLCO 3.3-1 required at lease
one channel
of log power available
when in Mode
5 and did not cover the situation of all Log Power
channels
After discussions
with the
NRC, the licensee
re-energized
two channels
of Log Power Instrumentation.
The
inspector further concluded that the safety significance of
down-powering the
Log Power
channels
was reduced
since both startup
~ channels
and associated
Boron Dilution Alarm Systems
were operable
at the time.
The inspector considered that the work planning
was
poor in that while the contractor would not work in energized
cabinets,
the
APS staff would not have
been prevented
from doing so.
This event
was briefly discussed
in a meeting with APS and Region
V
management
personnel
and on March 30,
1990 which followed an
enforcement
conference of the
same
day related to radiation
protection concerns.
This is an apparent violation of NRC requirements,
Enforcement
Item
(528/90-12-01).
Fires
on the Instrumentation
and Control
I8C) Sho
Roof
nl
On April 3, 1990,
and April 9, 1990,
two fires occurred
on the Unit
1 Instrumentation
and Control (I8C) shop roof as
a result of welding
slag which fell from construction activities adjacent to the roof.
An Operations
Support Building is being constructed
and tied to the
existing Control Building above the
I8C shop roof.
The fires were
localized
and declared
out by appropriate
individuals less
than
10
minutes after discovery.
No Notice of an Unusual
Event was
declared.
The onsite Fire Department did respond in both cases.
The inspector questioned
the licensee
and the contractor supervisor
for the construction activities regarding fire watches.
For the
first fire, a fire watch was not present
nor required since the
contract stated that fire watches
would only be required
when hot
work is occur ring on existing structures.
The Hot/Work Permit did
not require
a fire watch because it incorrectly stated that
no
flammable material
was within 35 feet of the hot work.
After the
first fire, the Manager of Nuclear Construction directed the
contractor to utilize fire watches for all future welding, cutting,
burning or grinding activities being performed over,
on, or in
existing structures
regardless
of whether weld/burn permits required
them or not.
A fire watch was present
during the welding activities which caused
the second fire.
The fire watch procedure
requires
a fire watch to
be posted during the
30 minutes after hot work activities
have
ceased.
During the welding activity on April 19, 1990, the Fire
Marshall issued
a Stop Work Order because
the fire extinguisher
available to the fire watch was marked
as not having
a current
inspection.
Approximately twenty-five minutes after the welding
ceased,
the fire watch asked
a roofer to watch the area for five
more minutes
so the fire watch could get
a properly marked/current
fire extinguisher to address
the Fire Marshall's
concern.
It was
during this five minute period the fire was discovered
as smoldering
felt wrapping.
A fire blanket
had been placed over the combustible
material,
but had been
removed during another activity without
objection from the fire watch.
The Plant Hanager
stopped the
construction activities until an incident investigation report was
complete
and satisfactory to the plant manager.
Corrective Actions
included retraining fire watches
in the Hot Work Permit procedure,
requiring currently marked fire extinguishers
and requiring fire
blankets.
No violations of NRC requirements
or deviations
were identified.
Snubber Visual Ins ections
Performed Without An Available
roce ure -
n>
During the
NRC inspector's
review of snubber
inspections
being
performed,. technicians
performing the inspections,
(procedure
were unable to identify or locate the work order or
procedure
by which they were working.
The technicians
had
an
unofficial checklist with them identifying the snubbers
to be
inspected
and the characteristics
to be inspected.
However, the
checklist- did not specify the inspection
method or acceptance
criteria.
The technician
searched
through documents
stored at the
hot tool issue counter
and in the snubber trailer, but still was
unable to locate either the procedure
or the work order.
The
technician did find some completed
data sheets
from the appropriate
procedure
from other snubber inspections,
from which the inspector
confirmed that the informal checklist correctly identified all
required. inspection points.
Later in the day, the technician
s
supervisor
produced the approved surveillance test procedure,
confirming that the work was authorized.
The inspector
concluded
that use of an informal check list appeared
to be inconsistent with
licensee policies,
however
no technical
concerns
were identified in
this case.
The licensee
acknowledged that workers not having ready
access
to work documents,
especially surveillance test procedures,
during the performance of work is inconsistent with licensee
policies.
No violations of NRC requirements
or deviations
were identified.
Zone III Controls - Unit 2 (93702
On approximately April 5, 1990, the licensee
discovered
approximately
44 small
beads,
later confirmed to be lead, inside the
cleanliness
Zone III controlled area
around the reactor cavity.
At
the time, the unit was defueled
and the cavity was drained.
The
beads
were found primarily along the rails adjacent to the cavity,
though
some were found in other areas,
including on the rails,
and
on the refueling machine.
An inspection of the edges of the cavity
was performed but no beads
were identified within the cavity or in
The licensee
performed
an engineering
evaluation of the beads
(EER 90-ZC-051)
and determined that
a small
number of lead beads
which are
used for shielding or weighing down
purposes
would not present
a safety concern
and would not interfere
with reactor operations
or water chemistry.
The licensee's
16
investigation of this event is still in progress
and will be
evaluated
by the inspector.
The inspector
toured the Zone III controlled area
and found another
bead,
some pieces of degraded rail,
some lock wire and other small
debris along the rails.
Additionally, the inspector
found that -the
Zone III controls in place at this time did not address
work over
the reactor cavity and transfer canal.
Carpenters
and other
craftsmen
were working on grating over the transfer canal without
Zone III controls.
The grating was not covered with herculite in
all areas until after a screw dropped into the transfer
canal
and
a
nail fell onto the Zone III Attendant's
desk.
The licensee
subsequently
covered the grating and the end of the transfer canal
with herculite
and
gC suspended
the carpenters'ork
until
conditions were reviewed.
In this case,
a licensee
equality Control
Inspector
had noted the Zone III control problem and was attempting
to pick up loose material
on the grating when the screw dropped into
the transfer
canal.
The failure to adequately establish
Zone III
requirements
is a violation of 30AC-9MPOl, "Foreign Material
Exclusion and Zone III Controls," which appears
to be
a violation of
NRC requirements.
The violation is not being cited because
the
criteria specified in Section
IJ'.G. of the Enforcement Policy were
satisfied.
Plant Housekee
in
- Units j, and
3 (71707)
During the inspecti.on period the inspectors
reviewed the material
condition of plant equipment
and general
housekeeping
related to the
prevention of accumulation of combustible materials.
while Unit j.
has
been in an extended
outage,
equipment material condition was
reviewed in connection with the approach to restart of Unit l.
During the inspection period the inspectors
had the following
concerns
about storage of combustible materials
and plant
housekeeping.
In Unit 3 on March 5, 1990, while the Unit was at
full power operations,
the inspectors
observed
several
wooden
pallets,
2 wooden packing cases
about 2'x2'x3',
one apparently full
55 gallon drum of a chemical with a flammability tag of No.
3
(severe
hazard)
and several
other drums also labelled flammability
No.
3.
These materials
were stacked
next to the Unit 3 Condensate
Storage
Tank (CST),
a safety-related
tank.
Licensee
procedure
paragraph
3.2. 10 states
that combustible material shall
not be stored next to outdoor safety-related
water storage
tanks.
This is an apparent violation of NRC requirements.
(Enforcement
Item 50-530/90-12-02).
In Unit 1 at the 140 foot eleVation of the Main Steam Support
Structure
(MSSS) areai
the inspectors
observed six cardboard
boxes,
approximately
2 x2'x3
apparently filled with'nsulating materials
on the "A" Train side
and similar boxes
on the "B" Train side.
Licensee
procedure
"Control of Combustible/Flammable
Materials
and Liquids", Paragraph
3. 1 requires that combustibles
stored in a nuclear safety-related
area,
have
a storage
permit and
that
a copy be posted.
There
was
no permit posted
or issued for the
I
II
materials
stored
on the 140 foot elevation of the
MSSS area.
The
inspector
noted that none of the safety-related
equipment
was
required to be operable for the existing plant mode;
however,
licensee
management
agreed that it was not appropriate
housekeeping
to store
such material in that location and had it promptly removed.
Other Unit 1 poor housekeeping
conditions were noted by the
inspectors
on March
5 and 7,
1990 in the
77 foot and 87 foot West
and East mechanical
piping penetration
areas.
Three full and
over-flowing bags of contaminated
Radiation Protection
(RP) clothing
were noted,
various small
RP items were lying on, the floor both
inside
and outside of the posted
contaminated
zones,
and various,
tools, piping caps
and plant components
were left unattended,
loose
and not in use.
The inspectors
noted
a crescent
wrench which was
laying across
a contaminated
area
boundary.
In addition, the
inspectors
also noted
a number of valves with boric acid leaks.
These
included CHV-198, CHV-930, CHV-087, SIA-HY-691, SJA-UV-637,
SIB-UV-646 and CHB-HV-255.
Other observations
in this area
included
a number of emergency lighting boxes apparently with corrosion
products,
and
a section of metal flashing between buildings that was
severely corroded.
In the Unit 1 condensate
transfer
pump
room on March 7, 1990, the
inspectors
observed
several
plant equipment discrepancies:
(1)
CTB
V-031,
a test connection,
was
uncapped
and leaking; (2) under the
"B" condensate
transfer
pump, the drip pump drains were clogged
and
the pan
had approximately 1/2 inch of standing water and in the
water of the pan contained biological growth and corrosion products;
and (3) under the "A" condensate
transfer
pump was corrosion
and
standing water.
The condensate
transfer
pumps are listed in the
licensee's
FSAR as being required for safe
shutdown in the event of
a plant fire.
During the inspection the inspectors
expressed
concern to licensee
management
about these
housekeeping
and materiaI condition
observations,
particularly related to needed
improvements prior to
the restart of Unit 1.
Subsequently,
the inspector
observed
improvements
in the specific areas
noted
and otherwise in general
as
outage
work was completed.
One violation of NRC requirements
was -identified.
Stuck Shut Essential
Ventilation Exhaust
Dam er for "D" Batter
On April 4, 1990, the "D" Battery Essential
Exhaust
Fan
was
observed
by the inspector to be apparently stuck in the closed
position.
The inspector
had observed that the lever on the normal
exhaust
fan dampers
in all four Unit 3 battery
rooms
was
perpendicular to the flowpaths,
even though the normal exhaust
fans
were running.
By standard
convention this would indicate the
were shut.
Plant personn'el
could not readily confirm that
the dampers
were open
and therefore
the ventilation lineup was
altered to observe
the action of both the normal
and essential
18
exhaust
fans
and dampers.
The normal
were confirmed to have
been
open, with some air flow.
The "D" essential
was found
to be stuck in the closed position when the "D" essential
exhaust
fan was started.
The damper technical
manual
states
that the damper
arm is to be positioned
as
needed to act as
a counter balance for
the damper.
= The arm may be installed in any position relative to
the actual
damper position to achieve this purpose.
Previously the
Unit 2 "8" essential
exhaust
fan damper
was observed
also to be
stuck
(See Inspection
Report 89-43).
The licensee
committed to
evaluating the preventive
maintenance
program in this area.
The "D"
essential
was
r epai red and declared
on April 6,
1990.
No violations of NRC requirements
or deviations
were identified.
Auxiliar
S ra
Pressurizer
S ra
Nozzle
Usa
e Factor - Unit 3
Determination of the Pressurizer
Spray Nozzle Usage Factor
has not
been performed
on Unit 3 since the March 3, 1989, plant trip.
Procedure
"Reactor Coolant System Transient
and
Operational
Cycles," is intended to be performed monthly in order to
assure
that the requirements
of Technical Specification (T/S) 5.7
are satisfied.
The T/S does not specify
how often the analyses
to
demonstrate
compliance with T/S 5.7 is to be performed.
The licensee
did not perform the surveillance
because
the requisite
data
was not recorded
when the auxiliary spray system
was
used in
March 1989,
so
an engineering evaluation
had to be performed to
determine
the correct data to use.
The engineering evaluation,
89-RC-245,
determined
the needed
data but also stated that the'/S
prescribed
formula for calculating auxi 1iary spray differential
temperature
should not be used.
This disposition clouded the issue
and resulted in an additional delay.
The licensee
has since
determined that this aspect of the disposition of EER 89-RC-245
was
inappropriate
and committed to perform the surveillance in
accordance
with the T/S methodology.
Since T/S Table 5.7-2 does not
extend to the range of differential temperatures
observed
during the
March, 1989, Unit 3 event,
and also in a Unit 1 event in 1986, the
licensee
performed
an evaluation
(EER 86-RC-159) to determine
an
appropriate
usage factor for the temperatures
observed.
The
surveillance
procedure,
includes this information.
The
NRC inspector
concluded that the licensee
had not exceeded
the
auxiliary spray pressurizer
nozzle cumulative
usage factor.
The
licensee
committed to reevaluate
a change to the T/S to include this
information.
No violations of NRC requirements
or deviations
were identified.
Desi
n Characteristics
of Auxiliar
(AFM) Motor
era
e
a
e
a ves -
ns
s
an
The inspector
reviewed the licensee's
documentation
related to the
condition in which AFM flow isolation valves
(AF-034 through 037)
l
19
were found not to go fully closed using the motor operator acting
against
design differential pressure.
This condition was
documented
in EER 87-AF-42 on the basis of actual test results in which the
valve remained
10 percent
open when it torqued out in the closing
direction.
The inspector
posed
several
questions with respect to
the associated
Justification for Continued Operation
(JCO).
These
questions
were not an immediate operability issue in that the
control valves wil1 provide
some pressure
reduction
so that the
isolation valves would not be required to close against full design
basis
event pressure
and flow.
In addition, the licensee's
response
to IE Bulletin (IEB) 85-03,
"Motor Operated
Valve
Common
Mode
Failures
Due to Improper Switch Settings"
appeared
to document that
these
valves would fully close
under design differential pressure.
The licensee
determined that the
JCO
may need to be clarified and
commited to review their IEB 85-03 response.
These
questions
and
several
other s will be followed up as
an Unresolved
Item
(528/90-12-02)
in a future inspect~on report.
Reactor Tri
Due to Dro
ed Control Element Assembl
CEA)
nl
During monthly
CEA surveillance testing in Unit 3,
CEA No.
80
dropped to the bottom of the core
and approximately
42 seconds
later
the Plant Protection
System
(PPS)
caused
on Departure
from Nucleate Boiling Rate
(DNBR)/Linear Power Density
(LPD).
Operators
stabilized
the plant in Mode 3 and declared
the
trip uncomplicated.
However, operators
took manual control of the
main feedwater
system
when it became
apparent that No.
1 Steam
Generator
(SG) level
was approaching
the Main Steam Isolation Signal
(MSIS) setpoint of 91 percent.
By taking manual control of
downcomer control valves, tripping both Main Feed
Pumps,
and closing
the economizer block valves,
the
No.
1
SG level
was stabilized at
approximately
84 percent.
,Operators
maintained auxiliary feedwater
flow to the
SG after securing
main feedwater.
The licensee
determined
the cause of the trip to be
a defective
circuit board in the
CEA No.
80 control cabinet.
The circuit board
was replaced
and retested satisfactorily.
The licensee
determined
that the component failure was
random.
The licensee further
determined that the initial unexpected
response
of one main
downcomer control valve could be explained
by the
"sticking" of a transfer switch relay in the Reactor Trip Override
(RTO) portion of the Feedwater
Control
System
(FWCS).
Although they
were unable to duplicate this "sticking", Combustion Engineering
concurred with their conclusion
and the relay was replaced.
Finally, the licensee attributed the moderate
overfeeding of No.
j.
SG to leakage
by the economizer flow control valves.
They
determined that the I/P converter
had drifted slightly out of
calibration causing the valve to seat slightly (1/4 inch) above the
shut seat with a zero
demand signal.
These controls were
re-calibrated
and returned to service.
The licensee
addressed
these
and several
other ancillary issues
in Incident Investigation Report
'I
~
20
20.
21.
22.
23.
2-3-90-001.
The inspector
reviewed this report
and concluded
the
licensee's
actions
appeared
appropriate.
No violations of NRC requirements
or deviations
were identified.
Confirmator
Action Letter Followu
- Unit 1 (92703)
The inspectors
reviewed
a selected
portion of the completed restart
action items commited to by the licensee for the restart of Unit 1.
This review was still in progress
at the end of the inspection
report period and will be documented further in a subsequent
inspection report.
No violation of NRC requirements
or deviations
were identified.
Non-licensed
0 erator Trainin
- Units
1
2
and
3 (41500)
The licensee's
general
employee training was reviewed.
No
violations of NRC requirements
or deviations
were identified.
Review of Licensee
Event
Re orts - Units 1
2 and
3 (90712
and 92700)
The following LERs were reviewed by the Resident
Inspectors:
89-16-LO (Closed) - Unit 1
.
This
LER addresses
an Engineer ed Safety Features
actuation
due to
failed components
in a Balance of Plant Essential
Safety Features
Actuation System Module.
This event
was addressed
in Inspection
Report 528/89-36.
The corrective actions
are complete.
This
LER is
closed.
88-20-Ll (Closed) - Unit 1.
This
LER addresses
Control Element Assembly (CEA) coil grounding
causing slipped
CEAs.
This issue
has
been recently addressed
in
Inspection
Reports
529/89-43,
529/89-49
and 529/89-50.,
The
corrective actions
are either completed or in progress
with
compensatory
measures
in place.
This
LER is closed.
No violations of NRC requirements
or deviations
were identified.
Review of Periodic
and
S ecial
Re orts - Units 1
2 and
3
Periodic
and special
reports
submitted
by the licensee
pursuant to
Technical Specifications
(T/S) 6.9. 1 and 6.9.2 were reviewed
by the
inspector.
This review included the following considerations:
the report
contained the information required to be reported
by NRC
requirements;
test results
and/or supporting information were
consistent with design predictions
and performance specifications;
and the validity of the reported information.
Within the scope of
the above,
the following reports
'were reviewed by the inspector.
Unit j.
o
Monthly Operating
Report for February
and March 1990.
Unit 2
o
Monthly Operating Report for February
and March 1990.
Unit 3
o
Monthly Operating Report for February
and March 1990.
No violations of NRC requirements
or deviations
were identified.
24.
Exit Meetin
(30703)
The inspector
met with licensee
management
representatives
periodically during the inspection
and held an exit meeting
on
April 19,
1990.
h